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

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

    190 Niagara Mohawk Power Corporation PP-190 Niagara Mohawk Power Corporation Presidential permit authorizing Niagara Mohawk Power Corporation to construct, operate, and maintain electric transmission facilitates at the U.S-Canadian border. PP-190 Niagara Mohawk Power Corporation (21.94 KB) More Documents & Publications PP-230-2 International Transmission Company Proposed Open Access Requirement for International Electric Transmission Facilities and Delegation to the Federal Energy Regulatory

  2. Review of integrated resource bidding at Niagara Mohawk

    SciTech Connect (OSTI)

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

    1992-05-01

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

  3. Review of integrated resource bidding at Niagara Mohawk

    SciTech Connect (OSTI)

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

    1992-05-01

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

  4. NIAGARA FALLS STORAGE SITE

    Office of Legacy Management (LM)

    :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

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

    Open Energy Info (EERE)

    Niagara Falls, New York: Energy Resources (Redirected from Niagara Falls, NY) Jump to: navigation, search Equivalent URI DBpedia Coordinates 43.0944999, -79.0567111 Show Map...

  6. American Ref-Fuel of Niagara Biomass Facility | Open Energy Informatio...

    Open Energy Info (EERE)

    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...

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

    Office of Legacy Management (LM)

    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

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

    Open Energy Info (EERE)

    New York's 28th congressional district.12 Registered Energy Companies in Niagara Falls, New York Canrom Photovoltaics Inc References US Census Bureau Incorporated...

  9. EIS-0153: Niagara Import Point Project

    Broader source: Energy.gov [DOE]

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

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

    Open Energy Info (EERE)

    York Niagara, New York North Tonawanda, New York Olcott, New York Pendleton, New York Porter, New York Ransomville, New York Rapids, New York Royalton, New York Somerset, New York...

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

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

    Community Solar Initiative | Department of Energy St. Regis Mohawk Tribe Paves the Way to a Sustainable Future; Kicks Off Community Solar Initiative St. Regis Mohawk Tribe Paves the Way to a Sustainable Future; Kicks Off Community Solar Initiative June 12, 2015 - 1:51pm Addthis Six photovoltaic arrays generate 32 kilowatts of energy to power 20 units at the Akwesasne Housing Authority’s (AHA) Sunrise Acres housing complex on the St. Regis Mohawk Reservation. Pictured from left to right

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

    Office of Legacy Management (LM)

    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

  13. Niagara Falls Storage Site, Lewiston, New York: geologic report

    SciTech Connect (OSTI)

    Not Available

    1984-06-01

    This report is one of a series of engineering and environmental reports planned for the US Department of Energy's properties at Niagara Falls, New York. It describes the essential geologic features of the Niagara Falls Storage Site. It is not intended to be a definitive statement of the engineering methods and designs required to obtain desired performance features for any permanent waste disposal at the site. Results are presented of a geological investigation that consisted of two phases. Phase 1 occurred during July 1982 and included geologic mapping, geophysical surveys, and a limited drilling program in the vicinity of the R-10 Dike, planned for interim storage of radioactive materials. Phase 2, initiated in December 1982, included excavation of test pits, geophysical surveys, drilling, observation well installation, and field permeability testing in the South Dike Area, the Northern Disposal Area, and the K-65 Tower Area.

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

    SciTech Connect (OSTI)

    Not Available

    1984-07-01

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

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

    Office of Legacy Management (LM)

    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

  16. ADDENDUM TO ACTION DESCRIPTION MEMORANDUM NIAGARA FALLS STORAGE SITE

    Office of Legacy Management (LM)

    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

  17. New runners to boost peak output at Niagara Falls

    SciTech Connect (OSTI)

    Reason, J.

    1990-01-01

    Retrofitted Francis turbines will improve the value of power generated from Niagara Falls by increasing the peak output of the hydroturbine units at the Robert Moses hydroelectric plant. The computer-designed runners are expected to add 330 MW to the peak capacity of the 28-yr-old plant and significantly increase the efficiency at high flow rates. Next year, the first new runner will be retrofit to the highly instrumented Unit 4. If the retrofit unit meets it increased-performance expectations, the other 12 units will be upgraded between 1993 and 1998. The work is part of an overall expansion of the Niagara Power Project designed to made better use of the power value of Niagara river water, within the constraints of a treaty with Canada and the scenic value of the falls. These constraints, together with varying flows and heads, introduced enormous complexities into the selection and design of the new runners. The alterations being made to Unit 4, in addition to replacing the turbine runner, include modifying the draft tube-liners, increasing the wicket-gate stroke, replacing the turbine discharge ring (to accommodate longer blades), making various electrical modifications to the generator, and replacing the transformer. But the key to the retrofit is the computer-designed runner. Charles Grose, senior project manager, New York Power Authority, White Plains, NY, emphasizes that such computer design techniques were not available a few years ago; neither were the computer-controlled machining techniques necessary to manufacture the new runners. Other aspects of the upgrading that were analyzed include runner stability, resonance, shaft torsional stress, and runaway speed.

  18. Environmental surveillance plan for the Department of Energy's Niagara Falls Storage Site (NFSS), Lewiston, New York

    SciTech Connect (OSTI)

    Englert, J.P.; Hinnefeld, S.L.

    1981-09-09

    The Niagara Falls Storage Site (NFSS) is a United States Department of Energy owned facility used for the storage of low-level radioactive residues. The site occupies 190 acres of the former Lake Ontario Ordnance Works and is located in the Niagara County town of Lewiston, in western New York State. The city of Niagara Falls is approximately eight (8) miles south of the NFSS. The purpose of this report is to describe environmental monitoring programs presently operated by NLO, and to suggest programs and revisions which should be implemented as a result of NLO's remedial actions at the NFSS.

  19. Keosauqua Municipal Light & Pwr | Open Energy Information

    Open Energy Info (EERE)

    Keosauqua Municipal Light & Pwr Jump to: navigation, search Name: Keosauqua Municipal Light & Pwr Place: Iowa Phone Number: 319-293-3406 Website: villagesofvanburen.comdirecto...

  20. Measurement of polycyclic aromatic hydrocarbons in the air along the niagara river

    SciTech Connect (OSTI)

    Hoff, R.M.; Chan, K.W.

    1987-06-01

    Two week-long studies in 1982-1983 have measure ambient concentrations of polycyclic aromatic hydrocarbons (PAH) and phthalate esters in air in both the particulate and gas phase along the US-Canadian border and the Niagara River. Concentrations of the PAH species monitored varied from 3 pg m/sup -3/ to 40 ng m/sup -3/. PAH's with three rings or less were found in significant proportions in the gas phase while larger molecules are almost solely in the particulate phase. Particulate components of the PAH loadings appear to originate locally with Buffalo, NY, Niagara Falls, NY, and Niagara Falls, Ontario, as probably sources. Gas-phase PAH components have a more regional character indicating regional or long-range transport. Levels of benzo(a)pyrene are consistent with previous particulate measurements made along the river since 1981.

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

    SciTech Connect (OSTI)

    Not Available

    1992-09-01

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

  2. Preliminary study on direct recycling of spent PWR fuel in PWR...

    Office of Scientific and Technical Information (OSTI)

    Preliminary study on direct recycling of spent PWR fuel in PWR system Citation Details ... conference on advances in nuclear science and engineering, Bali (Indonesia), 14-17 ...

  3. Niagara Air Quality Survey Report, 1987: Occidental Chemical Corporation, Niagara Falls, New York, USA, non-aqueous phase liquid (NAPL) incineration test. Report no. ARB-166-87-AR/SP

    SciTech Connect (OSTI)

    Bell, R.W.; DeBrou, G.

    1988-01-01

    An ambient air quality survey was conducted in the Niagara Falls area of Ontario from October 8-12, 1987 to provide on-site real-time screening for selected polychlorinated biphenyl congeners and other chlorinated organics at times when the Occidental Chemical Corporation was conducting tests at its liquid hazardous waste incineration facility in Niagara Falls, N.Y. During the incineration tests, the winds were such that the gaseous emissions from the Occidental facility were carried into the U.S. Since the monitoring units were restricted to the Canadian side of the Niagara River, only upwind air quality parameters could be measured.

  4. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect (OSTI)

    J.M. Acaglione

    2003-09-17

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

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

    Broader source: Energy.gov [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)

  6. Superfund Record of Decision (EPA Region 2): Love Canal (93rd Street School), Niagara County, City of Niagara Falls, NY. (Third remedial action), (amendment), May 1991. Final report

    SciTech Connect (OSTI)

    Not Available

    1991-05-15

    The Love Canal (93rd Street) site is an inactive hazardous waste site located in Niagara Falls, New York. The 19-acre 93rd Street School site, one of several operable units for the Love Canal Superfund site, is the focus of the Record of Decision (ROD). The fill material is reported to contain fly ash and BHC (a pesticide) waste. The ROD amends the 1988 ROD, and addresses final remediation of onsite contaminated soil through excavation and offsite disposal. The primary contaminants of concern affecting the soil are VOCs including toluene and xylenes; other organics including PAHs and pesticides; and metals including arsenic, chromium, and lead.

  7. Niagara Falls Storage Site, environmental monitoring report for 1979 and 1980

    SciTech Connect (OSTI)

    Weidner, R.B.; Boback, M.W.

    1981-10-05

    The Niagara Falls Storage Site is a 190-acre facility located in Niagara County, New York. It is owned by the US Department of Energy (DOE) and is used for the storage of radioactive residues. This site is managed by NLO, Inc., contract operator of the DOE Feed Materials Production Center near Cincinnati, Ohio. During 1979 and 1980, water and air samples were collected at and near the storage site to provide information about radionuclides in the offsite environment. Results show that uranium and radium concentrations in ground and surface water were within DOE Guide values for uncontrolled areas. Radon-222 in air at the site west boundary exceeded the DOE Guide but offsite monitoring in the general area showed radon-222 concentrations well within the Guide.

  8. Installation restoration program. Phase I. Records search. Niagara Falls Air Force Reserve facility, New York

    SciTech Connect (OSTI)

    Not Available

    1983-12-01

    The Department of Defense (DOD) has developed a program to identify and evaluate past hazardous material disposal sites on DOD property, to control the migration of hazardous contaminants, and to control hazards to health or welfare that may result from these past disposal operations. This program is called the Installation Restoration Program (IRP). The IRP has four phases consisting of Phase I, Initial Assessment/Records Search; Phase II, Confirmation and Quantification; Phase III, Technology Base Development; and Phase IV, Operations/Remedial Measures. Niagara Falls AFRF is located in Niagara County, New York, approximately six miles northeast of the City of Niagara Falls and approximately fifteen miles north of Buffalo. The installation is currently comprised of 985 acres with a base population of approximately 2,560. The following areas were determined to have a sufficient potential to create environmental contamination and follow-on investigation is warranted: Bldg. 600 JP-4 Pipeline Leak; POL JP-4 Tank C; Landfill; BX MOGAS Tank Leak; NYANG Hazardous Waste Drum Storage; POL JP-4 Tank A; JP-4 Tank Truck Spill; Bldg. 202 Drum Storage Yard; Fire Training Facility No. 1, 2 and 3; Bldg. 850 Drum Storage Yard; and AFRES Hazardous Waste Drum Storage.

  9. VERA Core Simulator Methodology for PWR Cycle Depletion (Conference...

    Office of Scientific and Technical Information (OSTI)

    VERA Core Simulator Methodology for PWR Cycle Depletion Citation Details In-Document Search Title: VERA Core Simulator Methodology for PWR Cycle Depletion Authors: Kochunas, ...

  10. Superfund record of decision amendment (EPA Region 2): Hooker (102nd Street Landfill), Niagara Falls, NY, June 9, 1995

    SciTech Connect (OSTI)

    1995-08-01

    This decision document presents the selected modification to the original remedial action (PB91-921417) for the 102nd Street Landfill Site (the `Site`), located in Niagara Falls, New York. The modification to the selected remedy addresses the river sediments within the shallow embayment of the Niagara River adjacent to the Site. The major components of the modification to the selected remedy include: dredging the Niagara River sediments to the `clean line` with respect to Site-related contamination. These sediments, after dewatering, will NOT be incinerated, but will be consolidated on the landfill. Any NAPL found within these sediments will be extracted, and will be incinerated at an off-site facility.

  11. Jet engine test stand and soil stockpile. 107th fighter-interceptor group Niagara Falls Air Force Reserve Station, New York Air National Guard, Niagara Falls, New York. Final site assessment addendum report, 9-12 February 1993

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    THis report outlines additional site assessment activities which were conducted at the Jet Engine Test Stand (JETS), Building No. 852 located at the 197th Fighter-Interceptor Group, Niagara Falls Air National Guard Station (NFANGS), Air Force Reserve Facility (AFRF) approximately 6 miles northeast of Niagara Falls, New York (Figure 1.1). The additional site assessment activities were performed in response to requests, dated February 9 and 12, 1993, by the New York State Department of Environmental Conservation (NYSDEC) to further investigate contaminated soil and groundwater conditions at the JETS and at an existing soil stockpile (Appendix A).

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

    Office of Legacy Management (LM)

    GROUND LEVEL INVESTIGATION OF ANOMALOUS RADIATION LEVELS IN NIAGARA FALLS, NEW YORK W. D. Cottrell, D. J. Christian, and F. F. Haywood ,d ;v ~ !;);;J;$ '9;) -i, - 'L." ; i--j -7,) ;3 i, Work performed by Health and Safety Research Division Oak Ridge National Laboratory Oak Ridge, Tennessee 37630 O&J. 2,7 +, / 7&y' March 1979 \ operated by UNION CARBIDE CORPORATIOII for the DEPARTMENT OF ENERGY as part of the Formerly Utilized Sites- Remedial Action Program

  13. Niagara Falls Storage Site annual site environmental monitoring report. Calendar year 1985

    SciTech Connect (OSTI)

    Not Available

    1986-04-01

    During 1985, an environmental monitoring program was continued at the Niagara Falls Storage Site (NFSS), a United States Department of Energy (DOE) surplus facility located in Niagara County, New York, presently used for the interim storage of low-level radioactive residues and contaminated soils and rubble. The monitoring program is being conducted by Bechtel National, Inc. Monitoring results show that the NFSS is in compliance with DOE concentration guides and radiation protection standards. Derived Concentration Guides (DCGs) represent the concentrations of radionuclides in air or water that would limit the radiation dose to 100 mrem/yr. The applicable limits have been revised since the 1984 environmental monitoring report was published. The limits applied in 1984 were based on a radiation protection standard of 500 mrem/yr; the limits applied for the 1985 are based on a standard of 100 mrem/yr. To determine whether the site is in compliance with DOE standards, environmental measurements are expressed as percentages of the applicable DCG, while the calculated doses to the public are expressed as percentages of the applicable radiation protection standard. The monitoring program measured radon gas concentrations in air; uranium and radium concentrations in surface water, groundwater, and sediments; and external gamma dose rates. Environmental samples collected were analyzed to determine compliance with applicable standards. Potential radiation doses to the public were also calculated.

  14. Niagara Falls Storage Site environmental monitoring report, Lewiston, New York, calendar year 1984

    SciTech Connect (OSTI)

    Not Available

    1985-07-01

    During 1984, an environmental monitoring program was continued at the Niagara Falls Storage Site, a United States Department of Energy (DOE) surplus facility located in Niagara County, New York, presently used for the storage of radioactive residues, contaminated soils and rubble. The monitoring program measured radon gas concentrations in air; uranium and radium concentrations in surface water, groundwater, and sediments; and external gamma exposure rates. Environmental samples collected were analyzed to determine compliance with applicable standards. Radiation doses to the public were also calculated. During 1984, annual average radon concentrations at the site boundary and exclusion area locations of the site were below the DOE Concentration Guide (CG) for uncontrolled areas. Annual average uranium and radium-226 concentrations in groundwater and surface water were below the DOE CG for release to uncontrolled areas. Sediment samples generally showed average concentrations of uranium and radium-226 lower than those measured in the past years. External gamma exposure rates were below the DOE Radiation Protection Standards. All radiation doses to the public were within DOE standards.

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

    SciTech Connect (OSTI)

    Not Available

    1990-05-01

    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.

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

    Broader source: Energy.gov [DOE]

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

  17. Health assessment for S-Area Landfill/Hooker, Niagara Falls, New York, Region 2. CERCLIS No. NYD000000001. Preliminary report

    SciTech Connect (OSTI)

    Not Available

    1989-06-01

    The S-Area Landfill is a National Priorities List site located in Niagara Falls, New York, that was operated from 1947 to 1975 by the Occidental Chemical Corporation (OCC). From 1947 to 1975, OCC is reported to have disposed of an estimated 19,000 tons of chlorobenzenes and approximately 17,000 tons of hexachlorocyclopentadiene. Other wastes disposed of at the site include organic phosphates, hexachlorobutadiene, trichlorophenols, and chlorinated toluenes. Environmental contamination from the S-Area exists on-site and off-site in soils and ground water. Further off-site contamination potentially exists in the Niagara River. The S-Area presents a potential public health threat to the consumers of the City of Niagara Falls drinking water and an incremental increase in contamination to fish in the Niagara River.

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

    Office of Legacy Management (LM)

    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

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

    SciTech Connect (OSTI)

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

    1986-11-01

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

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

    Open Energy Info (EERE)

    Valley El Pwr Assn Jump to: navigation, search Name: Pearl River Valley El Pwr Assn Place: Mississippi Phone Number: Columbia: 601-736-2666 -- Hattiesburg: 601-264-2458 -- Purvis:...

  1. Northeast Missouri El Pwr Coop | Open Energy Information

    Open Energy Info (EERE)

    Pwr Coop Jump to: navigation, search Name: Northeast Missouri El Pwr Coop Place: Missouri Phone Number: 573-769-2107 Website: www.northeast-power.coop Outage Hotline: 573-769-2107...

  2. Sam Rayburn Municipal Pwr Agny | Open Energy Information

    Open Energy Info (EERE)

    Municipal Pwr Agny Jump to: navigation, search Name: Sam Rayburn Municipal Pwr Agny Place: Texas Phone Number: 936-336-3684 or 936-336-5666 Website: www.cityofliberty.orgGOVERNME...

  3. Red River Valley Coop Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    Red River Valley Coop Pwr Assn Jump to: navigation, search Name: Red River Valley Coop Pwr Assn Place: Minnesota Website: www.rrvcoop.com Facebook: https:www.facebook.comRRVCPA...

  4. Polk County Rural Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Polk County Rural Pub Pwr Dist Jump to: navigation, search Name: Polk County Rural Pub Pwr Dist Place: Nebraska Phone Number: (888) 242-5265 Website: www.pcrppd.com Outage...

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

    Open Energy Info (EERE)

    E Pwr Coop Inc Jump to: navigation, search Name: Central Montana E Pwr Coop Inc Place: Montana Phone Number: 406-268-1211 Website: www.cmepc.org Outage Hotline: 406-268-1211...

  6. Superfund Record of Decision (EPA Region 2): Forest Glen Subdivision, Niagara Falls, NY. (First remedial action), December 1989

    SciTech Connect (OSTI)

    Not Available

    1989-12-29

    The Forest Glen Subdivision site consists of 21 acres of developed residential properties and undeveloped land in Niagara Fall, Niagara County, New York. Land in the area surrounding the Forest Glen subdivision is used for residential and industrial purposes, including a mobile home park, small shopping mall, and the CECOS Landfill. Chemical companies reportedly disposed of wastes onsite from the early 1950s to the early 1970s. Sampling by EPA's Field Investigation Team revealed the presence of high concentrations of unknown and tentatively identified compounds (TICs) in August 1987, and further soil sampling was conducted to identify the TICs. EPA has executed interim measures to stabilize site conditions including collecting, staging, and securing drums in areas north and east of the subdivision and temporarily covering visibily contaminated soil with concrete. The remedial activity is the first of two planned operable units and addresses resident relocation only. A subsequent operable unit will address the remediation of site contamination once the relocation is complete.

  7. Ocean disposal option for bulk wastes containing naturally occurring radionuclides: an assessment case history. [From Niagara Falls storage site

    SciTech Connect (OSTI)

    Stull, E.A.; Merry-Libby, P.

    1985-01-01

    There are 180,000 m/sup 3/ of slightly contaminated radioactive wastes (36 pCi/g radium-226) currently stored at the US Department of Energy's Niagara Falls Storage Site (NFSS), near Lewiston, New York. These wastes resulted from the cleanup of soils that were contaminated above the guidelines for unrestricted use of property. An alternative to long-term management of these wastes on land is dispersal in the ocean. A scenario for ocean disposal is present

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

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

  9. Life state response to environmental crisis: the case of the Love Canal, Niagara Falls, New York

    SciTech Connect (OSTI)

    Masters, S.K.

    1986-01-01

    This thesis explored the differences between two life stages - young and old - in perceiving and responding to man-made environmental disaster, as well as the support resources utilized to cope with disaster - personal, familial/friendship, and organizational. Because of the characteristics of man-made environmental disaster, and because of the different conditions of life and constructions of reality of older and younger families, it was expected that definitions of the situation would vary by life stage and locus of control - authoritative and personal. The research took place in the Love Canal neighborhood of Niagara Falls, New York. Fifty-eight families were interviewed in the fall of 1978, and thirty-nine of these families were reinterviewed in the spring of 1979. Interviews were tape recorded, transcribed, and coded. The data were presented in contingency tables and interview excerpts. The interview schedules elicited information of perception of impact, responses to impact, and the utilization of support resources. In an authoritative locus of control situation, the major findings were that both older and younger families perceived impact, that older families were slightly less disrupted, that younger families relied on organizational and familial/friendship support resources, and that older families relied on familial/friendship support resources.

  10. Environmental monitoring plan for the Niagara Falls Storage Site and the Interim Waste Containment Facility

    SciTech Connect (OSTI)

    Not Available

    1986-04-01

    As part of the US Department of Energy's (DOE) Surplus Facility Management Program (SFMP), the Niagara Falls Storage Site (NFSS) is undergoing remedial action. Vicinity properties adjacent to and near the site are being cleaned up as part of DOE's Formerly Utilized Sites Remedial Action Program (FUSRAP). These programs are a DOE effort to clean up low-level radioactive waste resulting from the early days of the nation's atomic energy program. Radioactively contaminated waste from these remedial action activities are being stored at the NFSS in an interim waste containment facility (IWCF). When the remedial actions and IWCF are completed in 1986, activities at the site will be limited to waste management. The monitoring program was prepared in accordance with DOE Order 5484.1 and is designed to determine the contribution of radioactivity from the site to the environs and to demonstrate compliance with applicable criteria. Major elements of this program will also supplement other monitoring requirements including the performance monitoring system for the IWCF and the closure/post-closure plan. Emphasis will be directed toward the sampling and analysis of groundwater, surface water, air and sediment for parameters which are known to be present in the material stored at the site. The monitoring program will employ a phased approach whereby the first 5 years of data will be evaluated, and the program will be reviewed and modified as necessary. 17 refs., 10 figs., 3 tabs.

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

    SciTech Connect (OSTI)

    Not Available

    1992-09-01

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

  12. Geochemical information for sites contaminated with low-level radioactive wastes: I. Niagara Falls Storage Site

    SciTech Connect (OSTI)

    Seeley, F.G.; Kelmers, A.D.

    1984-11-01

    The Niagara Falls Storage Site (NFSS) became radioactively contaminated as a result of wastes that were being stored from operations carried out to recover uranium from pitchblende ore in the 1940s and 1950s. The US Department of Energy (DOE) is considering various remedial action options for the NFSS. This report describes the results of geochemical investigations performed to help provide a quantitative evaluation of the effects of various options. NFSS soil and groundwater samples were characterized; and uranium and radium sorption ratios, as well as apparent concentration limit values, were measured in site soil/groundwater systems by employing batch contact methodology. The results suggest that any uranium which is in solution in the groundwater at the NFSS may be poorly retarded due to the low uranium sorption ratio values and high solubility measured. Further, appreciable concentrations of uranium in groundwater could be attained from soluble wastes. Release of uranium via groundwater migration could be a significant release pathway. Solubilized radium would be expected to be effectively retarded by soil at the NFSS as a result of the very high radium sorption ratios observed. The addition of iron oxyhydroxide to NFSS soils resulted in much higher uranium sorption ratios. Additional field testing of this potential remedial action additive could be desirable. 10 references.

  13. Superfund Record of Decision (EPA Region 2): Hooker Chemical S-Area, Niagara Falls, NY. (First remedial action), September 1990. Final report

    SciTech Connect (OSTI)

    Not Available

    1990-09-21

    The Hooker Chemical S-Area site is a former landfill area located in Niagara Falls, New York. The site lies adjacent to the Niagara River. Approximately 63,000 tons of chemical processing wastes were disposed of at the landfill. Ground water beneath the site also has been contaminated from aqueous phase and non-aqueous phase liquid chemicals. Chemicals have migrated toward the Niagara Falls Drinking Water Treatment Plant (DWTP) which lies to the east of the site, contaminating the Bedrock intake structures. The Record of Decision (ROD) addresses the landfill, a contaminated ground water plume, bedrock contamination, and the DWTP. The primary contaminants of concern affecting the soil, sediment, and ground water are VOCs including PCE; and other organics including chlorinated organics and pesticides. The selected remedial action for the site is included.

  14. Superfund Record of Decision (EPA Region 2): Hooker-102nd Street Landfill, Niagara Falls, NY. (First remedial action), September 1990. Final report

    SciTech Connect (OSTI)

    Not Available

    1990-09-26

    The 22-acre Hooker-102nd Street site is a former industrial landfill in the city of Niagara Falls, Niagara County, New York. The site is adjacent to, and partially within the Niagara River's 100-year floodplain. These studies and the Remedial Investigation (RI) initiated in 1984, identified contamination in ground water, onsite and offsite soil, rivershore sediment, and within a storm sewer. Additionally, the presence of a leachate plume of non-aqueous phase liquids (NAPLs) was discovered emanating from the landfill area. The Record of Decision (ROD) is the final remedy which addresses all of the contaminated media. The primary contaminants of concern affecting the soil, sediment, and ground water are VOCs including benzene, TCE, and toluene; other organics including PCBs and phenols; and metals including arsenic.

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

    Office of Legacy Management (LM)

    OR/20722-133 POST-REMEDIAL ACTION REPORT FOR THE NIAGARA FALLS STORAGE SITE VICINITY PROPERTIES - 1985 AND 1986 JANUARY 1989 Prepared for UNITED STATES DEPARTMENT OF ENERGY OAK RIDGE OPERATIONS OFFICE Under Contract No. DE-AC05-810R20722 M. E. Kaye and A. M. Feldman Bechtel National, Inc. Oak Ridge, Tennessee Bechtel J o b No. 14501 TABLE OF CONTENTS Abbreviations 1.0 Introduction 2.0 Remedial Action Guidelines 3.0 Remedial Action 4.0 Post-Remedial Action Sampling 5.0 Post-Remedial Action Status

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

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1984-01-01

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

  18. Superfund record of decision (EPA Region 2): Love Canal, Niagara Falls, New York, October 1987. Second remedial action

    SciTech Connect (OSTI)

    Not Available

    1987-10-26

    The Love Canal site is located in the southeast corner of the city of Niagara Falls and is approximately one-quarter mile north of the Niagara River. The canal was one of two initial excavations designed to provide inexpensive hydroelectric power for industrial development around the turn of the 20th century. Hooker Chemicals and Plastics Corporation (Hooker), now Occidental Chemical Corporation, disposed of over 21,000 tons of chemical wastes, including dioxin-tainted trichlorophenols, into Love Canal between 1942 and 1953. In the mid to late 1970s, continued periods of high precipitation contributed to water accumulation in the disposal area causing chemically-contaminated leachate to be carried to the surface and into contact with residential-basement foundations. Also, dioxin and other contaminants migrated from Love Canal to the sewers which have outfalls to nearby creeks. The remedial program at Love Canal has been extensive and has occurred in two phases. Approximately 30,400 cu yd - 40,900 cu yd of creek and sewer sediments are contaminated with 2,3,7,8-tetrachlorodibenzo-p-dioxin, commonly referred to as dioxin.

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

    SciTech Connect (OSTI)

    Not Available

    1987-06-01

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

  20. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect (OSTI)

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  1. South Mississippi El Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    South Mississippi El Pwr Assn Place: Mississippi Phone Number: 601.268.2083 Website: www.smepa.coop Outage Hotline: 601.268.2083 References: EIA Form EIA-861 Final Data File for...

  2. East Mississippi Elec Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    search Name: East Mississippi Elec Pwr Assn Place: Mississippi Phone Number: Meridian Office: 601-581-8600 -- Quitman Office: 601-776-6271 -- DeKalb Office: 601-743-2641 --...

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

    Open Energy Info (EERE)

    Valley Rrl Pwr Line, Inc Place: Colorado Website: www.gvp.org Twitter: @GVRuralPower Outage Hotline: 970-242-0040 Outage Map: www.gvp.orgcontentoutage-map References: EIA Form...

  4. Impact of High Burnup on PWR Spent Fuel Characteristics (Journal...

    Office of Scientific and Technical Information (OSTI)

    Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined ...

  5. Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys

    Broader source: Energy.gov [DOE]

    The purpose of this research effort is to determine the effects of canister/cask vacuum drying and storage on radial hydride precipitation in high‐burnup (HBU) pressurized water reactor (PWR)...

  6. PWR representative behavior during a LOCA

    SciTech Connect (OSTI)

    Allison, C.M.

    1981-01-01

    To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the public literature. Therefore, the objective of this paper is (a) to present calculations of LOCA induced behavior for Pressurized Water Reactor (PWR) core representative fuel rods, and (b) to discuss the variability in those calculations given the variability in fuel rod condition at the initiation of the LOCA. This analysis was limited to the study of changes in fuel rod behavior due to different power operating histories. The other two important parameters which affect that behavior, initial fuel rod design and LOCA coolant conditions were held invarient for all of the representative rods analyzed.

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

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

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

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

    SciTech Connect (OSTI)

    Not Available

    1989-04-01

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

  9. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    the CONFU assembly exhibits negative reactivity feedback coefficients comparable in ... NUCLEAR FUELS; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; REPROCESSING; SIMULATION; ...

  10. Swing-Down of 21-PWR Waste Package

    SciTech Connect (OSTI)

    A.K. Scheider

    2001-05-04

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design.

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

    SciTech Connect (OSTI)

    Faidy, C.

    1997-04-01

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

  12. 1,"Robert Moses Niagara","Hydroelectric","New York Power Authority",2438.8

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

    York" ,"Plant","Primary energy source","Operating company","Net summer capacity (MW)" 1,"Robert Moses Niagara","Hydroelectric","New York Power Authority",2438.8 2,"Ravenswood","Natural gas","TC Ravenswood LLC",2216.5 3,"Nine Mile Point Nuclear Station","Nuclear","Nine Mile Point Nuclear Sta LLC",1937 4,"Oswego Harbor

  13. Report on the performance monitoring system for the interim waste containment at the Niagara Falls Storage Site, Lewiston, New York

    SciTech Connect (OSTI)

    Not Available

    1985-10-01

    The Niagara Falls Storage Site (NFSS) is an interim storage site for low-level radioactive waste, established by the US Department of Energy (DOE) at Lewiston, New York. The waste containment structure for encapsulating low-level radioactive waste at the NFSS has been designed to minimize infiltration of rainfall, prevent pollution of groundwater, preclude formation of leachate, and prevent radon emanation. Accurately determining the performance of the main engineered elements of the containment structure will be important in establishing confidence in the ability of the structure to retain the wastes. For this purpose, a waste containment performance monitoring system has been developed to verify that these elements are functioning as intended. The key objective of the performance monitoring system is the early detection of trends that could be indicative of weaknesses developing in the containment structure so that corrective action can be taken before the integrity of the structure is compromised. Consequently, subsurface as well as surface monitoring techniques will be used. After evaluating several types of subsurface instrumentation, it was determined that vibrating wire pressure transducers, in combination with surface monitoring techniques, would satisfactorily monitor the parameters of concern, such as water accumulation inside the containment facility, waste settlement, and shrinkage of the clay cover. Surface monitoring will consist of topographic surveys based on predetermined gridlines, walkover surveys, and aerial photography to detect vegetative stress or other changes not evident at ground level. This report details the objectives of the performance monitoring system, identifies the elements of the containment design whose performance will be monitored, describes the monitoring system recommended, and outlines the costs associated with the monitoring system. 5 refs., 4 figs., 3 tabs.

  14. Targeted Health Assessment for Wastes Contained at the Niagara Falls Storage Site to Guide Planning for Remedial Action Alternatives - 13428

    SciTech Connect (OSTI)

    Busse, John; Keil, Karen; Staten, Jane; Miller, Neil; Barker, Michelle; MacDonell, Margaret; Peterson, John; Chang, Young-Soo; Durham, Lisa

    2013-07-01

    The U.S. Army Corps of Engineers (USACE) is evaluating potential remedial alternatives at the 191-acre Niagara Falls Storage Site (NFSS) in Lewiston, New York, under the Formerly Utilized Sites Remedial Action Program (FUSRAP). The Manhattan Engineer District (MED) and Atomic Energy Commission (AEC) brought radioactive wastes to the site during the 1940's and 1950's, and the U.S. Department of Energy (US DOE) consolidated these wastes into a 10-acre interim waste containment structure (IWCS) in the southwest portion of the site during the 1980's. The USACE is evaluating remedial alternatives for radioactive waste contained within the IWCS at the NFSS under the Feasibility Study phase of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process. A preliminary evaluation of the IWCS has been conducted to assess potential airborne releases associated with uncovered wastes, particularly during waste excavation, as well as direct exposures to uncovered wastes. Key technical issues for this assessment include: (1) limitations in waste characterization data; (2) representative receptors and exposure routes; (3) estimates of contaminant emissions at an early stage of the evaluation process; (4) consideration of candidate meteorological data and air dispersion modeling approaches; and (5) estimates of health effects from potential exposures to both radionuclides and chemicals that account for recent updates of exposure and toxicity factors. Results of this preliminary health risk assessment indicate if the wastes were uncovered and someone stayed at the IWCS for a number of days to weeks, substantial doses and serious health effects could be incurred. Current controls prevent such exposures, and the controls that would be applied to protect onsite workers during remedial action at the IWCS would also effectively protect the public nearby. This evaluation provides framing context for the upcoming development and detailed evaluation of

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

    SciTech Connect (OSTI)

    Not Available

    1986-05-01

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

  16. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

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

    2004-05-15

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

  17. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect (OSTI)

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

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

    SciTech Connect (OSTI)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-06

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

  19. Chemical behavior of fission products in the ORNL fission product release program. Supplement. [PWR; BWR

    SciTech Connect (OSTI)

    Collins, J.L.; Osborne, M.F.; Lorenz, R.A.

    1983-01-01

    Tests data are presented for BWR and PWR rods in test HI-4 and test HI-5. Operating conditions fission product release data are included.

  20. Health assessment for Hooker Chemical (102nd Street Landfill), Niagara Falls, New York, Region 2. CERCLIS No. NYD980506810. Preliminary report

    SciTech Connect (OSTI)

    Not Available

    1989-06-01

    The 102nd Street Landfill is two sites that comprise 22 acres. Occidental Chemical Corporation (OCC) and its predecessor, the Oldbury Electrochemical Company, deposited approximately 23,500 tons of mixed organic solvents, organic and inorganic phosphates, and related chemicals. Included in the site are approximately 300 tons of hexachlorocyclohexane process cake, including lindane. In addition, brine sludge, fly ash, electrochemical cell parts and related equipment in unknown quantities were dumped at the site. On-site contamination of the 102nd Street Landfill includes soils contaminated with non-aqueous phase liquids on both portions of the Landfill. Off-site contamination, based on current studies, results from contaminated ground-water leaching into the Niagara River which causes contamination of the river water, sediments, and aquatic organisms, including fish. The 102nd Street Landfill continues to represent a potential public health threat.

  1. Health assessment for Hyde Park Landfill National Priorities List (NPL) site, Niagara Falls, New York, Region 2. CERCLIS No. NYD000831644. Final report

    SciTech Connect (OSTI)

    Not Available

    1989-02-07

    The Hyde Park Landfill National Priorities List Site was used by Hooker Chemical and Plastic Corporation, now Occidental Chemical Corporation, to dispose of approximately 80,000 tons of waste from 1953 to 1975. Significant amounts of 2,3,7,8-tetrachlorodibenzo-p-dioxin is believed to be in the landfill. Site-related contaminants have been detected in the overburden and bedrock aquifers. Analyses of samples taken from ground water seeps at the Niagara Gorge Face also show site-related contaminants. Leachate from the landfill appears to have entered Bloody Run Creek. Sediment sample analyses from the creek show site-related contaminants. The 1985 U.S. Environmental Protection Agency Enforcement Decision Document outlines remedial activities to be conducted at the site. The site without remediation is of potential public health concern because of the risk to human health resulting from possible exposure to hazardous substances at concentrations that may result in adverse health effects.

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

    SciTech Connect (OSTI)

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

    1992-12-01

    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.

  3. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    Energy Science and Technology Software Center (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

  4. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in

  5. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17

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

  6. Comparison of PWR-IMF and FR fuel cycles

    SciTech Connect (OSTI)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj |; Necas, Vladimir

    2007-07-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  7. CASL - PWR Reactor Vessel Multi-Physics CFD Model

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

    PWR Reactor Vessel Multi-Physics CFD Model Jin Yan*1, Yiban Xu1, Andrew Petrarca1, Zeses Karoutas1, Emre Tatli1, Emilio Baglietto2, Jess Gehin3 1Westinghouse Electric Company LLC 2Massachusetts Institute of Technology 3Oak Ridge National Lab *Correspondence to: yan3j@westinghouse.com A complete 3D SolidWorks CAD model of Watts Bar Unit 1 was constructed based on drawings. A single fuel assembly CAD model including all geometrical details was created based on the Westinghouse V5H 17x17 fuel

  8. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    V. Delabrosse

    2003-02-27

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  9. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  10. Waterside corrosion of Zircaloy fuel rods. Final report. [PWR

    SciTech Connect (OSTI)

    Garzarolli, F.; Jung, W.; Schoenfeld, H.; Garde, A.M.; Parry, G.W.; Smerd, P.G.

    1982-12-01

    There is an economic incentive to extend average fuel-rod-discharge burnup to about 50 GWd/t. For these higher burnups it is necessary to know if increased waterside corrosion of the cladding will influence fuel-rod performance. For this reason, EPRI sponsored a joint program with C-E and KWU with the objective of investigating PWR waterside corrosion. This final report presents and discusses the results of various subtasks that comprised this project. In the review of corrosion data and models in the literature it was concluded that the PWR environment enhances the corrosion rate by about three times that expected from ex-reactor tests. A large number of fuel rods were characterized in both spent-fuel-pool and hot-cell campaigns. Chemical, physical and microstructural attributes of irradiated and unirradiated oxide films were measured. These included determinations of chemical composition, crystal structure, microstructure, density, specific heat, thermal conductivity, and post-irradiation autoclave corrosion behavior. Procedures used to calculate the fuel-rod surface temperature were reviewed. A model has been developed to predict in-reactor corrosion behavior.

  11. Westinghouse VANTAGE+ fuel assembly to meet future PWR operating requirements

    SciTech Connect (OSTI)

    Doshi, P.K.; Chapin, D.L.; Scherpereel, L.R.

    1988-01-01

    Many utilities operating pressurized water reactors (PWRs) are implementing longer reload cycles. Westinghouse is addressing this trend with fuel products that increase fuel utilization through higher discharge burnups. Higher burnup helps to offset added enriched uranium costs necessary to enable the higher energy output of longer cycles. Current fuel products have burnup capabilities in the area of 40,000 MWd/tonne U or more. There are three main phenomena that must be addressed to achieve even higher burnup levels: accelerated cladding, waterside corrosion, and hydriding; increased fission gas production; and fuel rod growth. Long cycle lengths also require efficient burnable absorbers to control the excess reactivity associated with increased fuel enrichment while maintaining a low residual absorber penalty at the end of cycle. Westinghouse VANTAGE + PWR fuel incorporates features intended to enhance fuel performance at very high burnups, including advances in the three basic elements of the fuel assembly: fuel cladding, fuel rod, and fuel assembly skeleton. ZIRLO {sup TM} cladding, an advanced Zircaloy cladding that contains niobium, offers a significant improvement in corrosion resistance relative to Zircaloy-4. Another important Westinghouse PWR fuel feature that facilitates long cycles is the zirconium diboride integral fuel burnable absorber (ZrB{sub 2}IFBA).

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

    SciTech Connect (OSTI)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

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

  13. PWR loss of feedwater ATWS: analysis and sensitivity study

    SciTech Connect (OSTI)

    Shier, W.G.; Lu, M.S.; Levine, M.M.; Diamond, D.J.

    1983-01-01

    The incident at the Salem Nuclear plant has presented a renewed interest in the analysis of the consequences of anticipated transients without scram (ATWS). This paper presents the results of an analysis of a complete loss of feedwater ATWS for a typical 4-loop PWR. The loss of feedwater transient was selected since previous analyses have shown that this transient produces one of the more limiting overpressure conditions in the primary system. These results provide a detailed analysis of this transient using current analytical techniques and show the sensitivity to several important parameters and plant modeling techniques. The RELAP5/MOD1 computer code has been used for this analysis. The code version is designated as Cycle 13 with additional modifications provided by both INEL and BNL.

  14. PWR and BWR spent fuel assembly gamma spectra measurements

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

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  15. Study of a Station Blackout Event in the PWR Plant

    SciTech Connect (OSTI)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao [Institute of Nuclear Energy Research P.O. Box 3-3, Longtan, 32500, Taiwan (China)

    2002-07-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

  16. Fort Drum integrated resource assessment

    SciTech Connect (OSTI)

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

    1992-12-01

    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.

  17. Selection of clc, cba, and fcb chlorobenzoate-catabolic genotypes from groundwater and surface waters adjacent to the Hyde Park, Niagara Falls, chemical landfill

    SciTech Connect (OSTI)

    Peel, M.C.; Wyndham, R.C.

    1999-04-01

    The frequency of isolation of three nonhomologous chlorobenzoate catabolic genotypes (clc, cba, and fbc) was determined for 464 isolates from freshwater sediments and groundwater in the vicinity of the Hyde Park industrial landfill site in the Niagara watershed. Samples were collected from both contaminated and noncontaminated sites during spring, summer, and fall and enriched at 4, 22, or 32 C with micromolar to millimolar concentrations of chlorobenzoates and 3-chlorobiphenyl. Hybridization at moderate stringency to restriction-digested genomic DNA with DNA probes revealed the chlorocatechol 1,2-dioxygenase operon (clcABD), the 3-chlorobenzoate 3,4-(4,5)-dioxygenase operon (cbaABC), and the 4-chlorobenzoate dehalogenase (fcbB) gene in isolates enriched from all contaminated sites in the vicinity of the industrial landfill. Nevertheless, the known genes were found in less than 10% of the isolates from the contaminated sites, indicating a high level of genetic diversity in the microbial community. The known genotypes were not enriched from the noncontaminated control sites nearby. The clc, cba, and fcb isolates were distributed across five phenotypically distinct groups based on Biolog carbon source utilization, with the breadth of the host range decreasing in the order clc > cba > fcb. Restriction fragment length polymorphism (RFLP) patterns showed that the cba genes were conserved in all isolates whereas the clc and fcb genes exhibited variation in RFLP patterns.

  18. Performance monitoring report for the Niagara Falls Storage Site Waste Containment Structure, Lewiston, New York: Calendar year 1987 and January--June of 1988

    SciTech Connect (OSTI)

    Blanke, J.A.; Johnson, R.T.; Stanley, W.F.

    1989-01-01

    A performance monitoring program has been developed for the Niagara Falls Storage Site (NFSS) Waste Containment Structure (WCS). The WCS contains soils contaminated with residual radioactive materials, rubble, and radioactive residues removed from various areas of the NFSS and vicinity properties during remedial action conducted by the Department of Energy (DOE) from 1982 through 1986. The NFSS is a part of the DOE Surplus Facilities Management Program (SFMP). The purpose of the performance monitoring program is to verify that the WCS main engineering elements are functioning to minimize infiltration of rainfall; prevent pollution of groundwater; preclude formation of leachate; and prevent radon emanation. This report presents the findings of performance monitoring conducted at the WCS during calendar year 1987, and January through June of 1988. the data received during the initial performance monitoring period in 1986 (Ref. 3) established a baseline for interpretation contained in this report. The period covered by this report has been expanded to include 6 months in 1988 because the impact of the winter is most evident in the spring growing season. 5 refs., 12 figs., 8 tabs.

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

    SciTech Connect (OSTI)

    Not Available

    1993-05-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1988-04-01

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

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

    Office of Scientific and Technical Information (OSTI)

    The optimization of 350 MWt small long life PWR result small excess reactivity and reduced ... on advances in nuclear science and engineering, Denpasar, Bali (Indonesia), 16-19 Sep ...

  2. Design study of long-life PWR using thorium cycle (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    life PWR core because it gives reactivity swing less than 1%Deltakk and longer ... long time operation with reduced excess reactivity as low as 0.53%Deltakk and reduced ...

  3. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  4. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  5. Assessment of alternatives for long-term management of uranium ore residues and contaminated soils located at DOE's Niagara Falls Storage Site

    SciTech Connect (OSTI)

    Merry-Libby, P.

    1984-11-05

    About 11,000 m/sup 3/ of uranium ore residues and 180,000 m/sup 3/ of wastes (mostly slightly contaminated soils) are consolidated within a diked containment area at the Niagara Falls Storage Site (NFSS) located about 30 km north of Buffalo, NY. The residues account for less than 6% of the total volume of contaminated materials but almost 99% of the radioactivty. The average /sup 226/Ra concentration in the residues is 67,000 pCi/g. Several alternatives for long-term management of the wastes and residues are being considered, including: improvement of the containment at NFSS, modification of the form of the residues, management of the residues separately from the wastes, management of the wastes and residues at another humid site (Oak Ridge, TN) or arid site (Hanford, WA), and dispersal of the wastes in the ocean. Potential radiological risks are expected to be smaller than the nonradiological risks of occupational and transportation-related injuries and deaths. Dispersal of the slightly contaminated wastes in the ocean is not expected to result in any significant impacts on the ocean environment or pose any significant radiological risk to humans. It will be necessary to take perpetual care of the near-surface burial sites because the residues and wastes will remain hazardous for thousands of years. If controls cease, the radioactive materials will eventually be dispersed in the environment. Predicted loss of the earthen covers over the buried materials ranges from several hundred to more than two million years, depending primarily on the use of the land surface. Groundwater will eventually be contaminated in all alternatives; however, the groundwater pathway is relatively insignificant with respect to radiological risks to the general population. A person intruding into the residues would incur an extremely high radiation dose.

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

    SciTech Connect (OSTI)

    Clayton, Christopher; Kothari, Vijendra; Starr, Ken; Widdop, Michael; Gillespie, Joey

    2012-02-26

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

  7. Analysis of Potential Hydrogen Risk in the PWR Containment

    SciTech Connect (OSTI)

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

    2006-07-01

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

  8. Assessment of PWR waterside corrosion models and data. Final report

    SciTech Connect (OSTI)

    Cox, B.

    1985-10-01

    The published data on waterside corrosion of PWR fuel cladding and unfuelled components have been reviewed, and the models used to assess the data have been studied. All corrosion models use too simplified a view of the corrosion process to obtain other than a general trend for the actual oxidation data. The in-reactor post-transition oxidation of the Zircaloys appears to be heavily dependent on water chemistry variations both between reactors, and along the length of an individual fuel rod. Crud deposition may be one primary cause of this, perhaps by allowing the independent development of the water chemistry within the crud layer, as much as by its effect on cladding surface temperatures. However, the effect of the thickening of the oxide film, which permits the development of an independent water chemistry inside the oxide, leading to an accelerating oxidation rate at large oxide thicknesses, seems to be the most important factor. It is concluded that a spectrum of results ranging from essentially no in-reactor enhancement of the oxidation rate to a sizeable enhancement (>10) may be seen depending upon the thickness of the oxide films, the water chemistry of the reactor, and crud deposition. A post-irradiation test that may help to distinguish between the factors involved has been suggested. 105 refs., 38 figs.

  9. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect (OSTI)

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

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

    SciTech Connect (OSTI)

    Putney, J.M.; Preece, R.J.

    1993-06-01

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

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

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

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

  12. In-core and ex-core calculations of the VENUS simulated PWR benchmark experiment

    SciTech Connect (OSTI)

    Williams, M.L.; Chowdhury, P.; Landesman, M.; Kam, F.B.K.

    1985-01-01

    The VENUS PWR engineering mockup experiment was established to simulate a beginning-of-life, generic PWR configuration at the zero-power VENUS critical facility located at CEN/SCK, Mol, Belgium. The experimental measurement program consists of (1) gamma scans to determine the core power distribution, (2) in-core and ex-core foil activations, (3) neutron spectrometer measurements, and (4) gamma heating measurements with TLD's. Analysis of the VENUS benchmark has been performed with two-dimensional discrete ordinates transport theory, using the DOT-IV code.

  13. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  14. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

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

    SciTech Connect (OSTI)

    Fei, T.; Hoffman, E.A.; Kim, T.K.; Taiwo, T.A.

    2013-07-01

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

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

    SciTech Connect (OSTI)

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

    1993-02-01

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

  17. Proceedings: 1983 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect (OSTI)

    1986-03-01

    Participants in this international workshop discussed research investigating mechanisms and propagation rates of intergranular corrosion in PWR steam generators. Laboratory test results, which have been consistent with power plant experience, permitted preliminary definition of corrosion rates in alloy 600 tubing.

  18. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

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

    SciTech Connect (OSTI)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-30

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

  20. Optimization of small long-life PWR based on thorium fuel

    SciTech Connect (OSTI)

    Subkhi, Moh Nurul; Suud, Zaki Waris, Abdul; Permana, Sidik

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

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

    SciTech Connect (OSTI)

    Preece, R.J.; Putney, J.M.

    1993-07-01

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

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

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01

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

  3. PCI-related cladding failures during off-normal events - draft. [PWR; BWR

    SciTech Connect (OSTI)

    Van Houten, R.; Tokar, M.; MacDonald, P.E.

    1984-05-01

    Pellet-cladding interaction (PCI) has long been identified as a fuel rod failure mechanism during power increases in both pressurized and boiling water reactors, and commercial guidelines have practically eliminated such failures during standard operations. A question remains regarding the possible formation of through-wall cladding cracks during several types of postulated off-normal reactor events involving power increases. This report includes preliminary findings for reactor events of the type addressed by Chapter 15 of the NRC Standard Review Plan. Specifically, the BWR turbine trip without bypass, PWR control rod withdrawal error, subcritical PWR control rod withdrawal error, BWR control blade withdrawal error, and the PWR steamline break are analyzed on the joint bases of peak rod power, power increase, ramp rate, and duration at elevated power. These Chapter 15 events are compared to numerous test reactor results and to other relevant investigations, and tentative conclusions on transient severity and data base adequacy are presented. Progress in developing computer codes for predicting PCI-induced fuel rod failures is also discussed. 49 references.

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

    SciTech Connect (OSTI)

    Clayton, Christopher; Kothari, Vijendra; Starr, Ken; Widdop, Michael; Gillespie, Joey

    2012-07-01

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

  5. Griffiss AFB integrated resource assessment

    SciTech Connect (OSTI)

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

    1993-02-01

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

  6. Analysis of a double-ended cold-leg break simulation: THTF Test 3. 05. 5B. [PWR

    SciTech Connect (OSTI)

    Craddick, W.G.; Pevey, R.E.

    1982-09-01

    On July 3, 1980, an experiment was performed in the Oak Ridge National Laboratory Thermal-Hydraulic Test Facility that simulated a double-ended cold-leg break pressurized-water reactor (PWR) accident. Analysis of the experiment revealed that nuclear fuel rods exposed to the same hydrodynamic environment as that which existed in the experiment would have departed from nucleate boiling both earlier and later than the fuel rod simulator (FRS), depending on the size of the gap between the nuclear fuel pellets and cladding and on the initial power of the nuclear fuel rod. Comparison of the results of the current experiment, which used an FRS bundle with geometry similar to 17 x 17 PWR fuel assemblies, to the results of earlier experiments, which used an FRS bundle with geometry similar to 15 x 15 PWR fuel assemblies, revealed no differences that can be attributed to the difference in geometries.

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

    SciTech Connect (OSTI)

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

    2000-10-03

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

  8. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  9. VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion

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

    CASL-U-2015-0302-000 VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion L3:VMA.AMA.P11.06 David Salazar, Westinghouse Fausto Franceschini, Westinghouse September 30, 2015 L3:VMA.AMA.P11.06 Official Use Only ii Protected under CASL Master NDA CASL-U-2015-0302-000 REVISION LOG Revision Date Affected Pages Revision Description 0 09/30/2015 All Initial issuance Document pages that are: Export Controlled ____________No______________________________________ IP/Proprietary/NDA

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

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.

    1997-04-01

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

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

    SciTech Connect (OSTI)

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

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

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

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

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

  13. Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5

    Energy Science and Technology Software Center (OSTI)

    1999-06-02

    CONTEMPT4/MOD6 describes the response of multicompartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user-supplied descriptions of compartments,more » inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. CONTEMPT4/MOD6 also provides analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation to accommodate degraded core type accidents.« less

  14. Phenomenon analysis of stress corrosion cracking in the vessel head penetrations of French PWR`s

    SciTech Connect (OSTI)

    Pichon, C.; Buisine, D.; Faidy, C.; Gelpi, A.; Vaindirlis, M.

    1995-12-31

    During a hydrotest in 1991, a leak was detected on,a reactor vessel head (RVH) penetration of a French PWR. This leak was due to a phenomenon of Primary Water Stress Corrosion Cracking (PWSCC) affecting these penetrations in Alloy 600. The destructive and non-destructive examinations undertaken during the following months highlighted the generic nature of the degradations. In order to well understand this phenomenon and implement the most suitable maintenance policy, a large scale scientific program was decided and performed jointly by Electricite de France and FRAMATOME. The paper will present all the results obtained in this program concerning the parameters governing the PWSCC. In particular the following fields will be developed: (1) the material, its microstructure in line with the manufacturing and its susceptibility to PWSCC; (2) the stresses and their evaluations by measurements, mock up corrosion tests and Finite Element Analysis (FEA); (3) the effect of surface finish on crack initiation; and (4) the crack growth rate. This phenomenon analysis will be useful for evaluating the risk of PWSCC on other Alloy 600 areas in PWR`s primary system.

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

    SciTech Connect (OSTI)

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

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

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

    SciTech Connect (OSTI)

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

    2009-05-01

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

  17. Study of a transient identification system using a neural network for a PWR plant

    SciTech Connect (OSTI)

    Ishihara, Yoshinao; Kasai, Masao; Kambara, Masayuki [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Mitsuda, Hiromichi; Kurata, Toshikazu; Shirosaki, Hidekazu [Inst. of Nuclear Safety System, Inc., Kyoto (Japan)

    1996-08-01

    This paper presents the procedure and results of a system for identifying PWR plant abnormal events, which uses neural network techniques. The neural network recognizes the abnormal event from the patterns of the transient changes of analog data from plant parameters when they deport from their normal state. For the identification of abnormal events in this study, events that cause a reactor to scram during power operation were selected as the design base events. The test data were prepared by simulating the transients on a compact PWR simulator. The simulation data were analyzed to determine how the plant parameters respond after the occurrence of a transient. A method of converting the pattern of the transient changes into characteristic parameters by fitting the data to pre-determined functions was developed. These characteristic parameters were used as the input data to the neural network. The neural network learning procedure used a generalized delta rule, namely a back-propagation algorithm. The neural network can identify the type of an abnormal event from a limited set of events by using these characteristic parameters obtained from the pattern of the changes in the analog data. From the results of this application of a neural network, it was concluded that it would be possible to use the method to identify abnormal events in a nuclear power plant.

  18. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect (OSTI)

    Wagner, J.C.

    2002-10-23

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

  19. Development of a model for predicting intergranular stress corrosion cracking of Alloy 600 tubes in PWR primary water. Final report

    SciTech Connect (OSTI)

    Garud, Y.S.

    1985-01-01

    A preliminary mathematical model developed in this study may make it possible to predict stress corrosion cracking on the primary side of PWR steam generator tubing. The study outlines a comprehensive testing program that will provide the operational and experimental data to further develop and verify the model.

  20. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  1. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

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

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

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

  3. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect (OSTI)

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  4. Decay Heat of Major Radionuclides for PWR Spent Fuels to 10,000 Years

    SciTech Connect (OSTI)

    J.S. Tang

    2001-12-20

    The objective of this calculation is to determine decay heat of a pressurized-water reactor (PWR) spent nuclear fuel (SNF) assembly with four different initial-enrichment and burnup characteristics. The major contributing radionuclides to the decay heat are also identified and graphically presented. The scope of this calculation is limited to the time period of the first 10,000 years after discharge from reactors. The results of this calculation will be used to evaluate the effects of the projected commercial spent nuclear fuel (CSNF) inventory on the repository design based on revised nuclear energy forecasts. This calculation was performed in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (BSC (Bechtel SAIC Company) 2001). AP-3.12Q, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the repository design activity.

  5. On the explanation and calculation of anomalous reflood hydrodynamics in large PWR cores

    SciTech Connect (OSTI)

    Rodriguez, S.E.

    1985-01-01

    Reflood hydrodynamics from large-scale (1:20) test facilities in Japan have yielded apparently anomalous behavior relative to FLECHT tests. Namely, even at reflooding rates below one inch per second, very large liquid volume fractions (10-15%) exist above the quench fronts shortly after flood begins; thus cladding temperature excursions are terminated early in the reflood phase. This paper discusses an explanation for this behavior: liquid films on the core's unheated rods. The experimental findings are shown to be correctly simulated with a new four-field (vapor, films, droplets) version of the best-estimate TRAC-PF1 computer code, TRAC-FF. These experimental and analytical findings have important implications for PWR large-break LOCA licensing.

  6. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect (OSTI)

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  7. Fuel-rod response during the large-break LOCA Test LOC-6. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% /sup 235/U). Each rod was surrounded by an individual flow shroud.

  8. LOCA rupture strains and coolability of full-length PWR fuel bundles

    SciTech Connect (OSTI)

    Mohr, C.L.; Hesson, G.M.

    1983-03-01

    The LOCA Simulation Program tests sponsored by the United States Nuclear Regulatory Commission are the first full-length nuclear-heated experiments designed to investigate the deformation and rupture characteristics as well as the coolability of nuclear-heated fuel under accident conditions. The results of the seven tests preformed in the program using 32-rod full-length PWR fuel bundles have shown that for a wide range of flow blockage condtions no significant reduction in coolability of the fuel bundle could be found. These results have been confirmed by data from out-of-pile electrically-heated experiments. Although there is a difference between nuclear and electrically-heated test data, the conclusion is still the same. Coolability of a deformed bundle during reflood is dominated by the dispersion of droplets in the deformed zone which provides adequate cooling and which is not reduced by the deformation of the fuel rod cladding.

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

    SciTech Connect (OSTI)

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

    1987-05-21

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

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

    SciTech Connect (OSTI)

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

    1985-02-01

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

  11. Probability and consequences of a rapid boron dilution sequence in a PWR

    SciTech Connect (OSTI)

    Diamond, D.J.; Kohut, P.; Nourbakhsh, H.; Valtonen, K.; Secker, P.

    1995-11-01

    The reactor restart scenario is one of several beyond-design-basis events in a pressurized water reactor (PWR) which can lead to rapid boron dilution in the core. This in turn can lead to a power excursion and the potential for fuel damage. A probabilistic analysis had been done for this event for a European PWR. The estimated core damage frequency was found to be high partially because of a high frequency for a LOOP and assumptions regarding operator actions. As a result, a program of analysis and experiment was initiated and corrective actions were taken. A system was installed so that the suction of the charging pumps would switch to the highly borated refueling water storage tank (RWST) when there was a trip of the RCPs. This was felt to reduce the estimated core damage frequency to an acceptable level. In the US, this original study prompted the Nuclear Regulatory Commission to issue an information notice to follow work being done in this area and to initiate studies such as the work at BNL reported herein. In order to see if the core damage frequency might be as high in US plants, a probabilistic assessment of this scenario was done for three plants. Two important conservative assumptions in this analysis were that (1) the mixing of the injectant was insignificant and (2) fuel damage occurs when the slug passes through the core. In order to study the first assumption, analysis was carried out for two of the plants using a mixing model. The second assumption was studied by calculating the neutronic response of the core to a slug of deborated water for one of the plants. All three types of analyses are summarized below. More information is available in the original report.

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

    SciTech Connect (OSTI)

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

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

  13. The CASTOR-V/21 PWR spent-fuel storage cask: Testing and analyses: Interim report

    SciTech Connect (OSTI)

    Dziadosz, D.; Moore, E.V.; Creer, J.M.; McCann, R.A.; McKinnon, M.A.; Tanner, J.E.; Gilbert, E.R.; Goodman, R.L.; Schoonen, D.H.; Jensen, M.

    1986-11-01

    A performance test of a Gesellschaft fuer Nuklear Service CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask was performed. The test was the first of a series of cask performance tests planned under a cooperative agreement between Virginia Power and the US Department of Energy. The performance test consisted of loading the CASTOR-V/21 cask with 21 PWR spent fuel assemblies from Virginia Power's Surry reactor. Cask surface and fuel assembly guide tube temperatures, and cask surface gamma and neutron dose rates were measured. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Limited spent fuel integrity data were also obtained. Results of the performance test indicate the CASTOR-V/21 cask exhibited exceptionally good heat transfer performance which exceeded design expectations. Peak cladding temperatures with helium and nitrogen backfills in a vertical cast orientation and with helium in a horizontal orientation were less than the allowable of 380/sup 0/C with a total cask heat load of 28 kW. Significant convection heat transfer was present in vertical nitrogen and helium test runs as indicated by peak temperatures occurring in the upper regions of the fuel assemblies. Pretest temperature predictions of the HYDRA heat transfer computer program were in good agreement with test data, and post-test predictions agreed exceptionally well (25/sup 0/C) with data. Cask surface gamma and neutron dose rates were measured to be less than the design goal of 200 mrem/h. Localized peaks as high as 163 mrem/h were measured on the side of the cask, but peak dose rates of <75 mrem/h can easily be achieved with minor refinements to the gamma shielding design. From both heat transfer and shielding perspectives, the CASTOR-V/21 cask can, with minor refinements, be effectively implemented at reactor sites and central storage facilities for safe storage of spent fuel.

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

    SciTech Connect (OSTI)

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

    1992-12-01

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

  15. Radionuclide release from PWR fuels in a reference tuff repository groundwater subsquently changed to Radionuclide release from PWR fuels in J-13 well water

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-04-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: (1) fuel rod sections split open to expose bare fuel particles; (2) rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; (3) rod sections with water-tight end fittings and two 200-{mu}m diameter holes through the cladding; and (4) undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested in deionized water. Selected initial results are also given for Turkey Point fuel specimens tested in J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water.

  16. Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Varacalle, D.J. Jr.; Koizumi, Y.; Giri, A.H.; Koske, J.E.; Sanchez-Pope, A.E.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAP5/MOD1 computer code showed good agreement with the experimental data.

  17. Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Grush, W.H.; Woerth, S.C.; Koizumi, Y.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data.

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

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29

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

  19. Examination of spent PWR fuel rods after 15 years in dry storage.

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-02-11

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas

  20. Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-07-01

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 deg. C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited pre-storage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission

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

    SciTech Connect (OSTI)

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

    2009-01-01

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

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

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E.

    1997-04-01

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

  3. TREAT source-term experiment STEP-1 simulating a PWR LOCA

    SciTech Connect (OSTI)

    Simms, R.; Baker, L. Jr.; Blomquist, C.A.; Ritzman, R.L.

    1986-01-01

    In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.

  4. CEPAN method of analyzing creep collapse of oval cladding. Volume 5. Evaluation of interpellet gap formation and clad collapse in modern PWR fuel rods

    SciTech Connect (OSTI)

    Adams, W.M.; Fisher, H.D.; Litke, H.J.; Mordarski, W.J.

    1985-04-01

    This report presents the results from a review of interpellet-gap formation, ovality, creepdown and clad collapse data in modern PWR fuel rods. Conclusions are reached regarding the propensity of modern PWR fuel to form such gaps and to undergo clad collapse. CEPAN, a creep-collapse predictor approved by the NRC in 1976, has been reformulated to include the creep analysis of cladding with finite interpellet gaps. The basis for this reformulation is discussed in detail. The model previously used in the calculation of the augmentation factor, a peak linear heat rate penalty due to the presence of interpellet gaps within the fuel rod, has been modified to incorporate gap-formation statistics from modern fuel. Finnally, the benefits of the limited gap formation and the CEPAN reformulation for the licensing of modern PWR fuel rods are evaluated.

  5. On-line PWR RHR pump performance testing following motor and impeller replacement

    SciTech Connect (OSTI)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  6. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect (OSTI)

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

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

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

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

  8. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect (OSTI)

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  9. Radionuclide release from PWR fuels in a reference tuff repository groundwater

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-03-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; rod sections with water-tight end fittings and two 200-{mu}m-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables.

  10. Containment pressurization and burning of combustible gases in a large, dry PWR containment during a station blackout sequence

    SciTech Connect (OSTI)

    Lee, M.; Fan, C.T. (National Tsing-Hua Univ., Dept. of Nuclear Engineering, Hsinchu (TW))

    1992-07-01

    In this paper, responses of a large, dry pressurized water reactor (PWR) containment in a station blackout sequence are analyzed with the CONTAIN, MARCH3, and MAAP codes. Results show that the predicted containment responses in a station blackout sequence of these three codes are substantially different. Among these predictions, the MAAP code predicts the highest containment pressure because of the large amount of water made available to quench the debris upon vessel failure. The gradual water boiloff by debris pressurizes the containment. The combustible gas burning models in these codes are briefly described and compared.

  11. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    SciTech Connect (OSTI)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made.

  12. Fort Drum integrated resource assessment

    SciTech Connect (OSTI)

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

    1992-12-01

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

  13. O:ELECTRICPP-190.PDF

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

    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

  14. LOFTRAN/RETRAN comparison calculations for a postulated loss-of-feedwater ATWS in the Sizewell 'B' PWR

    SciTech Connect (OSTI)

    Papez, K.L.; Risher, D.H.

    1983-05-01

    The loss-of-main-feedwater transient without reactor trip (scram) has received particular attention in pressurized water reactor (PWR) anticipated transient without scram (ATWS) analysis primarily due to the potential for reactor coolant system over pressurization. To assist in the licensing of the U.K. PWR, Sizewell 'B', comparative calculations of a loss-of-feedwater ATWS have been performed using the Westinghouse-developed LOFTRAN loop analysis code and the Electric Power Research Institute/ Energy Incorporated-developed RETRAN-01 code. The calculations were performed with and without the emergency boration system (EBS), which is included in the Sizewell reference design. Initial results showed good agreement between the codes for the major features of the transient, but also a time shift in the transient profiles at the time of the pressurizer pressure peak. This was found to be due to differences in the steam generator modeling, which resulted in a difference in the onset of the very rapid degradation in heat transfer as the steam generators approach dryout. When the same model was used in both codes, very good agreement was obtained. Remaining differences in the results are attributed primarily to differences in the boron injection models, which resulted in an over-prediction of the core boron concentration in the RETRAN calculation. The results with an EBS indicate that the peak pressurizer pressure is relatively insensitive to variations in modeling.

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

    SciTech Connect (OSTI)

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

    1983-01-01

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

  16. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5. [PWR

    SciTech Connect (OSTI)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)) commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected.

  17. Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants. Case study: PWR during routine operations

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Haggard, D.L.

    1986-01-01

    The purpose of this report is to present data in support of evaluating the impact of fuel cladding failure events on occupational radiation exposure. To determine quantitatively whether fuel cladding failure contributes significantly to occupational radiation exposure, radiation exposure measurements were taken at comparable locations in two mirror-image pressurized-water reactors (PWRs) and their common auxiliary building. One reactor, Unit B, was experiencing degraded fuel characterized as 0.125% fuel pin-hole leakers and was operating at approximately 55% of the reactor's licensed maximum core power, while the other reactor, Unit A, was operating under normal conditions with less than 0.01% fuel pin-hole leakers at 100% of the reactor's licensed maximum core power. Measurements consisted of gamma spectral analyses, radiation exposure rates and airborne radionuclide concentrations. In addition, data from primary coolant sample results for the previous 20 months on both reactor coolant systems were analyzed. The results of the measurements and coolant sample analyses suggest that a 3560-megawatt-thermal (1100 MWe) PWR operating at full power with 0.125% failed fuel can experience an increase of 540% in radiation exposure rates as compared to a PWR operating with normal fuel. In specific plant areas, the degraded fuel may elevate radiation exposure rates even more.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    Results of testing employing surrogate instrumented rods (non-high-burnup, 17 x 17 PWR fuel assembly) to capture the response to the loadings experienced during normal conditions of transport indicate that strain- or stress-based failure of fuel rods seems unlikely; performance of high-burnup fuels continues to be assessed.

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

    SciTech Connect (OSTI)

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    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

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

    SciTech Connect (OSTI)

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

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

  1. Impact Analysis of a Dipper-Type and Multi Spring-Type Fuel Rod Support Grid Assemblies in PWR

    SciTech Connect (OSTI)

    Song, K.N.; Yoon, K.H.; Park, K.J.; Park, G.J.; Kang, B.S.

    2002-07-01

    A spacer grid is one of the main structural components in a fuel assembly of a Pressurized light Water Reactor (PWR). It supports fuel rods, guides cooling water, and maintains geometry from external impact loads. A simulation is performed for the strength of a spacer grid under impact load. The critical impact load that leads to plastic deformation is identified by a free-fall test. A finite element model is established for the nonlinear simulation of the test. The simulation model is tuned based on the free-fall test. The model considers the aspects of welding and the contacts between components. Nonlinear finite element analysis is carried out by a software system called LS/DYNA3D. The results are discussed from a design viewpoint. (authors)

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

    SciTech Connect (OSTI)

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

    1981-11-01

    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).

  3. The influence of dissolved hydrogen on primary water stress corrosion cracking of Alloy 600 at PWR steam generator operating temperatures

    SciTech Connect (OSTI)

    Jacko, R.J.; Economy, G.; Pement, F.W.

    1992-12-31

    PWR primary coolant chemistry uses an intentional dissolved hydrogen concentration of 20 to 50 ml (STP)/kg of water to effect a net suppression of oxygen-producing radiolysis, to minimize corrosion in primary loop materials and to maintain a low redox potential. Speculation has attended a possible influence of dissolved hydrogen on the kinetics of initiation of Primary Water Stress Corrosion Cracking (PWSCC) behavior of Alloy 600 steam generator tubing. Three series of experiments are presented for conditions in which the level of dissolved hydrogen was intentionally varied over the hydrogen and temperature ranges of interest for steam generator operation. No significant effect of dissolved hydrogen was found on PWSCC of Alloy 600.

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

    SciTech Connect (OSTI)

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

    1980-01-01

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

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

    Open Energy Info (EERE)

    Organization Natural Currents Energy Services Project Technology *MHK TechnologiesRED HAWK Project Licensing FERC License Docket Number P-14408 Environmental Monitoring and...

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

    Open Energy Info (EERE)

    & Drain Dist Place: Arizona Phone Number: (928) 785-3351 Website: www.wmidd.orgpower.html Outage Hotline: (928) 785-3351 References: EIA Form EIA-861 Final Data File for 2010 -...

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

    SciTech Connect (OSTI)

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

    1993-09-01

    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.

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

    SciTech Connect (OSTI)

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter; Bertholdt, Horst-Otto; Adams, Andreas; Impertro, Michael; Roesch, Josef

    2013-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

    Hermann, O.W.

    2000-02-01

    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.

  11. Tests with Inconel 600 to obtain quantitative stress-corrosion cracking data for evaluating service performance. [PWR

    SciTech Connect (OSTI)

    Bandy, R.; van Rooyen, D.

    1982-09-01

    Inconel 600 tubes in pressurized water reactor (PWR) steam generators form a pressure boundary between radioactive primary water and secondary water which is converted to steam and used for generating electricity. Under operating conditions the performance of alloy 600 has been good, but with some occasional small leaks resulting from stress corrosion cracking (SCC), related to the presence of unusually high residual or operating stresses. The suspected high stresses can result from either the deformation of tubes during manufacture, or distortion during abnormal conditions such as denting. The present experimental program addresses two specific conditions, i.e., (1) where deformation occurs but is no longer active, such as when denting is stopped and (2) where plastic deformation of the metal continues, as would occur during denting. Laboratory media consist of pure water as well as solutions to simulate environments that would apply in service; tubing from actual production is used in carrying out these tests. The environments include both normal and off chemistries for primary and secondary water. The results reported here were obtained in several different tests. The main ones are (1) split tube reverse U-bends, (2) constant extension rate tests (CERT), and (3) constant load. The temperature range covered is 290 to 365/sup 0/C.

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

    SciTech Connect (OSTI)

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

    1985-11-01

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

  13. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    SciTech Connect (OSTI)

    Clerc, T.; Hebert, A.; Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B.

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  14. Effects of Zircaloy oxidation and steam dissociation on PWR core heat-up under conditions simulating uncovered fuel rods

    SciTech Connect (OSTI)

    Viskanta, R.; Mohanty, A.K.

    1986-04-01

    The studies described in this report identify the regimes of slow transients in a partially uncovered core of a PWR. The threshold height and onset time for oxidation of the cladding of a fuel rod have been evaluated. The effects of oxidation in increasing the decay heat load, component temperature, reduction of cladding thickness and generation of hydrogen have been estimated. The condition for steam starvation has been determined. At high uncovered core heights, typically say 2.8 m for a geometry simulating the TMI-2 type of reactor, the solid and coolant temperatures can reach the limits of steam dissociation. The effects of radiation heat exchange between cladding and coolant, Zircaloy oxidation, steam dissociation, gap conductance between fuel and cladding and system pressure on the heatup of fuel rods have been investigated. The time for uncovering a certain core height is taken as the independent parameter. It is seen that if the uncovering process is allowed to continue beyond 9 minutes corresponding to an uncovered height of 1.9 m, onset of cladding oxidation can be a reality. These values provide a guideline for the response time of the emergency core cooling systems. 10 refs., 22 figs.

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

    SciTech Connect (OSTI)

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

    2012-05-01

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

  16. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect (OSTI)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  17. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 2: primary coolant loop model. Final report

    SciTech Connect (OSTI)

    Eberhardt, A.C.

    1981-09-01

    This report describes the Zion Station reactor coolant loop model developed by Sargent and Lundy Engineers for Lawrence Livermore National Laboratory as part of its Load Combination Program. This model was developed for use in performing seismic time history analyses of an actual pressurized water reactor (PWR) system. It includes all major items affecting the seismic response of a 4-loop Westinghouse nuclear steam supply system: the components, supports, and interconnecting piping. The model was further expanded to permit static analysis of dead weight, thermal, and internal pressure load conditions.

  18. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5: probabilistic fracture mechanics analysis. Final report

    SciTech Connect (OSTI)

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

    1981-08-01

    The purpose of the portion of the Load Combination Program covered in this volume was to estimate the probability of a seismic induced loss-of-coolant accident (LOCA) in the primary piping of a commercial pressurized water reactor (PWR). Such results are useful in rationally assessing the need to design reactor primary piping systems for the simultaneous occurrence of these two potentially high stress events. The primary piping system at Zion I was selected for analysis. Attention was focussed on the girth butt welds in the hot leg, cold leg and cross-over leg, which are centrifugally cast austenitic stainless steel lines with nominal outside diameters of 32 - 37 inches.

  19. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 2. Primary Coolant Loop Model. Load Combination Program, Project I final report

    SciTech Connect (OSTI)

    Eberhardt, A.C.

    1981-06-01

    This report describes the Zion Station reactor coolant loop model developed by Sargent and Lundy Engineers for Lawrence Livermore National Laboratory as part of its Load Combination Program. This model was developed for use in performing seismic time history analyses of an actual pressurized water reactor (PWR) system. It includes all major items affecting the seismic response of a 4-loop Westinghouse nuclear steam supply system: the components, supports, and interconnecting piping. The model was further expanded to permit static analysis of dead weight, thermal, and internal pressure load conditions. 7 refs., 42 figs., 9 tabs.

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

    SciTech Connect (OSTI)

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

    1980-01-01

    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.

  1. 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 (OSTI)

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

    1982-01-01

    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.

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

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-03-14

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

  3. BIOMASS REBURNING - MODELING/ENGINEERING STUDIES

    SciTech Connect (OSTI)

    Vladimir Zamansky; Chris Lindsey; Vitali Lissianski

    2000-01-28

    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.

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

    SciTech Connect (OSTI)

    Comnes, G.A.; Stoft, S.; Greene, N.; Hill, L.J.

    1995-11-01

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

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

    SciTech Connect (OSTI)

    1980-05-01

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

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

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

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

    SciTech Connect (OSTI)

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

    1982-09-01

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

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

    SciTech Connect (OSTI)

    Bierschbach, M.C.; Mencinsky, G.J.

    1993-10-01

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

  9. Corrosion and hydriding performance evaluation of three zircaloy-2 clad fuel assemblies after continuous exposure in PWR cores 1 and 2, at Shippingport, PA

    SciTech Connect (OSTI)

    Hillner, E.

    1980-01-01

    Three original Zircaloy-2 clad blanket fuel bundles from the pressurized water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure during Cores 1 and 2. Detailed visual examination of these components after approx. 6300 calendar days of operation (51,140 EFPH) revealed only the anticipated uniform light gray (post-transition) corrosion products with no evidence of unexpected corrosion deterioration, fuel rod warpage, or other damage. All corrosion films were found to be tightly adherent to the underlying cladding. An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced appreciably greater end-of-life rod average oxide film thickness when compared with corresponding values produced from a set of empirical equations generated from the out-of-pile (autoclave) testing of Zircaloy coupons.

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

    SciTech Connect (OSTI)

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

    2008-05-05

    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.

  11. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect (OSTI)

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  12. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 1. Summary, Load Combination Program. Project I final report

    SciTech Connect (OSTI)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1981-06-01

    This report summarizes work performed to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading and to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR, is the demonstration plant used in this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated by a deterministic fracture mechanics model with stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without earthquake, is very small (on the order of 10/sup -12/). The probability of a leak was found to be several orders of magnitude greater than that of a large LOCA, complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported.

  13. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

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

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

    640,119 434,526 324,474 278,422 233,453 200,394 1982-2014 Import Price 4.67 5.43 4.96 3.83 5.59 8.60 1989-2014 Export Volume 0 0 38,783 68,843 184,071 201,691 1982-2014 Export Price -- -- 4.69 3.61 4.29 5.56 199

    223,532 202,549 188,208 183,548 185,119 196,365 1990-2016 Base Gas 114,992 114,956 114,913 114,853 114,603 114,779 1990-2016 Working Gas 108,540 87,594 73,296 68,695 70,516 81,586 1990-2016 Net Withdrawals 2,020 21,931 14,573 4,660 -1,571 -11,246 1990-2016 Injections 4,390 351 2,066

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

    Office of Legacy Management (LM)

    of Engineers but will eventually transfer to the U.S. Department of Energy Office of Legacy Management. Assessment of Historical Knolls Atomic Power Laboratory Waste Storage...

  16. Niagara Falls, NY Natural Gas Exports to Canada

    Gasoline and Diesel Fuel Update (EIA)

    3a . January Monthly Peak Hour Demand, Actual by North American Electric Reliability Corporation Region, 2005 through 2009 (Megawatts) Texas Power Grid Western Power Grid Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) Peak Hour Demand (MW) 2005 613,416 41,247 32,236 47,041 154,200 166,190 29,072 40,966 102,464 2006 563,711 34,464 37,056 43,661 149,252 134,239 26,864

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

    Gasoline and Diesel Fuel Update (EIA)

    88,983 32,770 3,159 1,650 2,957 2,539 1996-2015 Pipeline Prices 5.43 4.68 3.22 4.04 5.08 3.2

  18. DOE - Office of Legacy Management -- Niagara Falls Vicinity Properties...

    Office of Legacy Management (LM)

    This site is currently managed by the U.S. Army Corps of Engineers but will eventually transfer to the U.S. Department of Energy Office of Legacy Management. Assessment of ...

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

    Office of Legacy Management (LM)

    This site is currently managed by the U.S. Army Corps of Engineers but will eventually transfer to the U.S. Department of Energy Office of Legacy Management. Assessment of ...

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

    Open Energy Info (EERE)

    Overseeing Organization ECOsponsible Project Licensing FERC License Docket Number P-13840 Environmental Monitoring and Mitigation Efforts See Tethys << Return to the MHK database...

  1. MHK Projects/Niagara Community | Open Energy Information

    Open Energy Info (EERE)

    Overseeing Organization ECOsponsible Project Licensing FERC License Docket Number P-13839 Environmental Monitoring and Mitigation Efforts See Tethys << Return to the MHK database...

  2. Evaluation of Final Radiological Conditions at Areas of the Niagara...

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

    Paper presented at the Waste Management 2012 Conference. February 26 through March 1, 2012, Phoenix, Arizona. Christopher Clayton, Vijendra Kothari, and Ken Starr, U.S. Department ...

  3. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    SciTech Connect (OSTI)

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  4. 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

    SciTech Connect (OSTI)

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

    2006-07-01

    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

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

    SciTech Connect (OSTI)

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

    2006-10-20

    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

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

    SciTech Connect (OSTI)

    Rebak, R B; Hua, F H

    2004-07-12

    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.

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

    SciTech Connect (OSTI)

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

    1981-06-01

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

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

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01

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

  9. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (1.4 kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.

    1981-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.4 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a stainless steel canister representative of actual fuel canisters, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel near-surface drywell tests being conducted at E-MAD, the spent fuel deep geologic storage test being conducted in Climax granite on the Nevada Test Site, and for five constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

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

    Energy Savers [EERE]

    June 12, 2015 - 1:51pm Addthis Six photovoltaic arrays generate 32 kilowatts of energy to ... Indigenous Collaboration, Inc. Six photovoltaic arrays generate 32 kilowatts of energy ...

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

    SciTech Connect (OSTI)

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

    1982-09-01

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

  12. Review of consolidated Edison`s integrated resource bidding program

    SciTech Connect (OSTI)

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

    1993-07-01

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

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

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

    Thousand Cubic Feet) 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 4.09 4.15 5.51 3.0

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

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

    Thousand Cubic Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2011 5.04 5.23 4.96 4.84 4.94 5.04 4.94 4.56 4.39 3.96 4.54 4.07 2012 3.71 3.32 2.93 2.33 4.18 4.09 2013 4.13 3.91 4.34 4.47 4.42 4.24 4.03 3.76 3.85 3.87 4.08 4.67 2014 6.87 10.13 9.18 5.08 4.70 4.57 4.07 3.69 3.69 3.92 5.09 4.80 2015 4.29 5.02 4.48 2.93 2.70 2.63 2.55 2.64 2.67 2.59 2.72 2.42 2016 2.97 2.69 2.33 2.11 1.91 2.2

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

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

    Feet) 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 594 39 2,215 3 0 0 0 0 0 2010's 0 6,535 23,386 158,102 162,321 172,59

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

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

    Feet) 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 12,716 11,162 12,737 13,738 13,789 13,174 13,904 13,939 13,022 13,267 12,940 13,715 2014 14,211 13,161 14,549 13,012 13,641 13,234 13,634 13,645 13,274 13,071 13,092 13,797 2015 13,470 10,576 13,577 13,247 14,219 13,657 14,297 14,009 14,476 14,578 17,873 18,617 2016 18,408 16,645 18,791 18,312 19,004 18,80

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

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

    Thousand Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1990's 2.90 2.87 2.62 2.58 2000's 4.10 4.94 3.55 5.71 6.41 9.06 7.43 7.36 9.58 4.63 2010's 5.43 4.68 3.22 4.04 5.08 3.20

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

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

    Thousand Cubic Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2011 5.02 4.80 4.46 4.75 4.72 5.03 4.80 4.49 4.24 3.98 4.05 3.87 2012 3.46 3.13 2.81 2.38 2.97 3.26 2.43 3.22 3.66 4.18 3.85 2013 3.89 3.75 4.21 4.46 4.38 3.86 3.57 3.33 4.12 3.91 3.24 4.26 2014 4.28 5.86 5.38 21.67 22.33 21.67 22.33 22.33 3.24 2.68 2015 3.45 4.36 3.65 2.08 3.14 2.85 2.73 2.31 1.97 2016 2.78 2.63 1.92 2.21 2.06 2.45

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

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

    Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1990's 291,193 288,865 303,599 372,515 2000's 421,016 308,102 367,448 369,052 363,350 390,272 354,703 356,529 298,911 188,525 2010's 88,983 32,770 3,159 1,650 2,957 2,539

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

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

    Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2011 9,497 6,894 4,421 2,459 935 1,060 1,452 979 644 1,443 1,404 1,583 2012 1,411 720 681 93 7 56 31 6 12 35 105 2013 299 291 49 103 12 0 0 6 281 97 41 471 2014 1,092 908 893 0 0 0 0 0 57 6 2015 168 796 162 23 6 356 351 333 346 2016 618 462 347 277 298 358

  1. U.S. Army Corps of Engineers Buffalo District Office 1776 Niagara...

    Office of Legacy Management (LM)

    ... In 1960, the property was transferred to the Ashland Oil Company and has been used as part of this company's oil refinery activities since that time. In 1974, Ashland Oil Company ...

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

    Gasoline and Diesel Fuel Update (EIA)

    Thousand Cubic Feet) 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 4.09 4.15 5.51 3.0 Thousand Cubic Feet)

    Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2011 5.04 5.23 4.96 4.84 4.94 5.04 4.94 4.56 4.39 3.96 4.54 4.07 2012 3.71 3.32 2.93 2.33 4.18 4.09 2013 4.13 3.91 4.34 4.47 4.42 4.24 4.03 3.76 3.85 3.87 4.08 4.67 2014 6.87 10.13 9.18 5.08 4.70 4.57 4.07 3.69 3.69 3.92 5.09

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

    Gasoline and Diesel Fuel Update (EIA)

    Thousand Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1990's 2.90 2.87 2.62 2.58 2000's 4.10 4.94 3.55 5.71 6.41 9.06 7.43 7.36 9.58 4.63 2010's 5.43 4.68 3.22 4.04 5.08 3.20 Thousand Cubic Feet)

    Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2011 5.02 4.80 4.46 4.75 4.72 5.03 4.80 4.49 4.24 3.98 4.05 3.87 2012 3.46 3.13 2.81 2.38 2.97 3.26 2.43 3.22 3.66 4.18 3.85 2013 3.89 3.75 4.21 4.46 4.38 3.86 3.57 3.33 4.12 3.91 3.24 4.26 2014

  4. Consolidation and disposal of PWR fuel inserts

    SciTech Connect (OSTI)

    Wakeman, B.H. (Virginia Electric and Power Co., Glen Allen, VA (United States))

    1992-08-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry Installation will accommodate 84 such casks with a total storage capacity of 811 MTU of spent pressurized water reactor fuel assemblies. Virginia Power provided three storage casks for testing at the Idaho National Engineerinq Laboratory's Test Area North and the testing results have been published by the Electric Power Research Institute. Sixty-nine spent fuel assemblies were transported in truck casks from the Surry Power Station to Test Area North for testing in the three casks. Because of restrictions imposed by the cask testing equipment at Test Area North, the irradiated insert components stored in these fuel assemblies at Surry were removed prior to transport of the fuel assemblies. Retaining these insert components proved to be a problem because of a shortage of spent fuel assemblies in the spent fuel storage pool that did not already contain insert components. In 1987 Virginia Power contracted with Chem-Nuclear Systems, Inc. to process and dispose of 136 irradiated insert components consisting of 125 burnable poison rod assemblies, 10 thimble plugging devices and 1 part-length rod cluster control assembly. This work was completed in August and September 1987, culminating in the disposal at the Barnwell, SC low-level radioactive waste facility of two CNS 3-55 liners containing the consolidated insert components.

  5. Stress corrosion cracking of Alloy 600. [PWR

    SciTech Connect (OSTI)

    Serra, E.

    1981-11-01

    The stress corrosion cracking of Alloy 600 tubing has affected the performance of several pressurized water reactor steam generators. The purpose of this report is to summarize the research which has followed that reviewed by D. van Rooyen in 1975. Although several papers and reports have been published there still is not a general model that can explain the stress corrosion cracking behavior of Alloy 600 in deaerated or aerated high-temperature pure water or in the environments that might exist in the primary and secondary coolant of a steam generator. Such a model, if it exists, must cover the complex interaction of the environmental, metallurgical, and mechanical variables which control the susceptibility of Alloy 600 to stress corrosion cracking. Each of these classes of variables is discussed in the text.

  6. Best-Estimate Analysis PWR LOCA.

    Energy Science and Technology Software Center (OSTI)

    2005-11-11

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

  7. LOFT lead rod test results evaluation. [PWR

    SciTech Connect (OSTI)

    Driskell, W.B.; Tolman, E.L.

    1980-07-30

    The purpose for evaluating the LOFT Lead Rod Test (simulations of large break, loss-of-coolant accidents) data was to determine; (a) if the centerline thermocouple and fuel rod elongation sensor data show indications of the collapsed fuel rod cladding, (b) the capability of the FRAP-T5 computer code to accurately predict cladding collapse, and (c) if cladding surface thermocouples enhance fuel rod cooling. With consideration to unresolved questions on data integrity, it was concluded that: the fuel rod centerline thermocouple and elongation sensor data do show indications of the fuel rod cladding collapse; the FRAP-T5 code conservatively predicts cladding collapse; and there is an indication that cladding surface thermocouples are enhancing fuel rod cooling.

  8. Fuel performance during severe accidents. [PWR

    SciTech Connect (OSTI)

    Buescher, B.J.; Gruen, G.E.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. This program is underway in the Power Burst Facility at the Idaho National Engineering Laboratory. In preparation for the first test, predictions have been performed using the TRAC-BD1 computer. This paper presents the calculated results showing a slow heatup to 2400 K over 5 hours, and the analysis includes accelerated oxidation of the zirconium cladding at temperatures above 1850 K.

  9. Cracked-fuel mechanics. [PWR; BWR

    SciTech Connect (OSTI)

    Williford, R.E.; Lanning, D.D.

    1982-01-01

    This paper presents a modelling concept and a set of measurable parameters that have been shown to improve the prediction of the mechanical behavior of cracked fuel/cladding systems without added computational expense. The transition from classical annular gap/cylindrical pellet models to modified bulk properties and further to local behavior for cracked fuel systems is discussed. The results of laboratory experiments to verify these modelling parameters are shown. Data are also presented from laboratory experiments on unirradiated and irradiated rods which show that fuel rod mechanical response depends on fuel fragment size. The impact of these data on cracked fuel behavior and failure modelling is also discussed.

  10. PWR fuel performance program. Final report

    SciTech Connect (OSTI)

    Kunishi, H.; Miller, R.S.; Roberts, E.

    1985-09-01

    Tests of 15 x 15 and 17 x 17 fuel assemblies, irradiated at burnup levels well beyond the standard 33 GWd/MtU, showed that no inherent material limitations stand in the way of such assemblies achieving average burnups of approximately 55 GWd/MtU. The study also showed that cladding waterside corrosion can be minimized by controlling the lithium-to-boron ratio in the coolant to reduce crud deposition.

  11. Superfund record of decision (EPA Region 2): Forest Glen subdivision, Operable Unit 2, Niagara Falls, NY, March 31, 1998

    SciTech Connect (OSTI)

    1998-09-01

    This operable unit represents the second of three operable units planned for the site. It addresses the principal threats posed by the site through controlling the source of contamination. The major components of the selected remedy include: excavation of contaminated soils from the southern portion of the site, and contaminated sediment from East Gill Creek, and consolidation of these materials in the northern portion of the site followed by grading in preparation for placement of the cap; construction of an 8.5-acre cap over the consolidated soils in the northern portion of the site in conformance with the major elements described in 6 New York Code of Rules and Regulations Part 360 for solid waste landfill caps; removal and off-site disposal of the vacant trailers and two permanent homes to facilitate the excavation of soils; and capping the Wooded Wetland with six inches of clean sediment.

  12. The role of IRP in the natural gas industry: A case study

    SciTech Connect (OSTI)

    Wright, J.A.; Brockman, L.; Herman, P.

    1994-09-29

    The natural gas industry has changed radically over the last decade. The Federal Energy Regulatory Commission`s Order 636 completed plans to unbundle interstate pipeline services and create open access for distribution companies and their customers. There has also been increasing competition for local distribution companies (LDCs) from fuel oil, electricity and unregulated energy service companies. Meanwhile, the Energy Policy Act of 1992 includes provisions that encourage energy efficiency and promote reliance on competitive forces. In response to these changes, coupled with growing environmental concerns and the need for increased energy efficiency, a number of state public utility commissions and LDCs took an interest in integrated resource planning (IRP) for gas utilities. Gas IRP was in its formative stages and a variety of regulatory approaches were being considered when this project began. In response, this project originated with the total project scope being to define, implement and institutionalize an IRP process for the Gas Customer Service Business Unit of Niagara Mohawk Power Corporation (NMGas).

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

    SciTech Connect (OSTI)

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

    1994-12-01

    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.

  14. Common trenching reduces damage to buried utilities

    SciTech Connect (OSTI)

    Alfiere, E.P.

    1982-09-01

    Since 1972 Niagara Mohawk Power Co. has established a utility corridor, installing 503 miles of buried gas mains and electric cables in a common trench. Their guidelines for common trenching included (1) the developer's responsibility for providing a subdivision map showing the location of each sidewalk, lot, and roadway, (2) an easement strip paralleling the front lot (street) line that is to be cleared and graded by the developer before construction is started, (3) an electric planning department to prepare detailed construction drawings, coordinate plans with other utilities, determine the responsibility for trenching and backfilling, and determine that all the necessary easements have been secured, and (4) construction specifications varying the width and depth of the trench with the number and type of utilties occupying the joint trench. Advantages of the common trench program comprise reduced exposure to digups, communication and concern for each utility's facility, water and sewer construction installed before the common trench, and cost sharing that would reduce each facility's construction and restoration costs.

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

    DOE Patents [OSTI]

    Bollinger, L.R.

    1983-11-03

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

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

    DOE Patents [OSTI]

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

    1975-12-19

    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.

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

    DOE Patents [OSTI]

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

    1980-06-06

    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.

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

    Open Energy Info (EERE)

    9,647.445 93,322.028 60,225 3,117.42 30,825.248 8,207 692.763 8,259.846 11 13,457.628 132,407.122 68,443 2008-06 9,059.584 86,892.462 60,106 3,046.146 30,089.083 8,193 709.428...

  19. Property:EnvReviewPwrPlantSiting | Open Energy Information

    Open Energy Info (EERE)

    Showing 15 pages using this property. R RAPIDGeothermalEnvironmentAlaska + None RAPIDGeothermalEnvironmentCalifornia + The California Energy Commission or a designated...

  20. Vermont Public Pwr Supply Auth | Open Energy Information

    Open Energy Info (EERE)

    Website: www.vppsa.com Facebook: https:www.facebook.compagesVermont-Public-Power-Supply-Authority Outage Hotline: 802-244-7678 Outage Map: vtoutages.com...

  1. East River Elec Pwr Coop, Inc | Open Energy Information

    Open Energy Info (EERE)

    Facebook: https:www.facebook.compagesEd-the-Energy-Expert431620883566287?refts&frefts Outage Hotline: (605) 256-8057 or (605) 256-8056 or (605) 256-8059...

  2. Arizona Electric Pwr Coop Inc | Open Energy Information

    Open Energy Info (EERE)

    Facebook: https:www.facebook.compagesArizonas-GT-Cooperatives347352335037?refts Outage Hotline: (520) 586-3631 References: EIA Form EIA-861 Final Data File for 2010...

  3. Virgin Islands Wtr&Pwr Auth | Open Energy Information

    Open Energy Info (EERE)

    form View source History View New Pages Recent Changes All Special Pages Semantic SearchQuerying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit with...

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

    Open Energy Info (EERE)

    Yes Activity Buying Transmission Yes Activity Distribution Yes Activity Wholesale Marketing Yes Activity Bundled Services Yes This article is a stub. You can help OpenEI by...

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

    SciTech Connect (OSTI)

    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.

    1982-07-01

    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.

  6. Rushmore Electric Pwr Coop Inc | Open Energy Information

    Open Energy Info (EERE)

    MRO NERC MRO Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility...

  7. Michigan South Central Pwr Agy | Open Energy Information

    Open Energy Info (EERE)

    Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility...

  8. South Mississippi El Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility...

  9. Vermont Yankee Nucl Pwr Corp | Open Energy Information

    Open Energy Info (EERE)

    Utility Location Yes Ownership I NERC Location NPCC NERC NPCC Yes ISO NE Yes Activity Wholesale Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility...

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

    SciTech Connect (OSTI)

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

    1980-04-01

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

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

    SciTech Connect (OSTI)

    Bailey, W.J.; Tokar, M.

    1982-12-01

    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.

  12. Northwest Rural Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 13805 Utility Location Yes Ownership P NERC Location WECC NERC MRO Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  13. South Central Public Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 17548 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes RTO SPP Yes Activity Distribution Yes Activity Retail Marketing Yes This article is a stub....

  14. Seward County Rrl Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 16954 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  15. North Central Public Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 13698 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  16. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

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

    1999-10-01

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

  17. Demonstration of a noise-surveillance system at a PWR

    SciTech Connect (OSTI)

    Smith, C.M.

    1982-01-01

    The automated surveillance system has monitored the Sequoyah Nuclear Plant during its first fuel cycle. The system was able to acceptably adapt to different plant operating conditions. While evaluations are still ongoing, results indicate that the system was able to adapt to signals with different statistical character and that the discriminants are useful in detecting spectral changes. The system monitored long-term noise behavior, detected spectra that differ from what is considered normal, and provided concise storage of spectra together with the plant operating condition associated with the stored spectra.

  18. Prediction of quench and rewet temperatures. [PWR; BWR

    SciTech Connect (OSTI)

    Gunnerson, F.S.

    1980-01-01

    Many postulated nuclear reactor accidents result in high-temperature dryout or film boiling within the nuclear core. In order to mitigate potential fuel rod damage or rod failure, safe or lower fuel rod temperatures must be reestablished by promoting coolant/cladding contact. This process is commonly referred to as quenching or rewetting, and often, these terms are not differentiated. All theoretical predictions of the cooling process by various models based on single or multidimensional analytical and numerical studies require a knowledge of either the quenching or the rewetting temperature. The purpose of this paper is to define quench and rewet temperatures and present a method whereby each may be estimated.

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

    SciTech Connect (OSTI)

    Christensen, R.

    1980-03-01

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

  20. PBF LOCA test LOC-6 fuel-behavior report. [PWR

    SciTech Connect (OSTI)

    Broughton, T.M.; Vinjamuri, K.; Hagrman, D.L.; Golden, D.W.; MacDonald, P.E.

    1983-04-01

    This report presents the results of Loss-of-Coolant (LOC) Test LOC-6, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., for the US Nuclear Regulatory Commission. Postirradiation examination results are included, together with the results of thermal-hydraulic and fuel behavior calculations using the RELAP4 and FRAP-T6/BALON-2 computer codes. Two of the four light water reactor type fuel rods ballooned and ruptured during the test. Peak cladding temperatures at the rupture locations were high in the alpha phase (1066 and 1098/sup 0/K). The effects of initial rod internal prepressurization and prior irradiation were investigated during the experiment. The effect of rod prepressurization was found to be significant, and, for burnups of about 17,000 MWd/t, prior irradiation increased cladding circumferential strains at failure.

  1. Core melt/coolant interactions: modelling. [PWR; BWR

    SciTech Connect (OSTI)

    Berman, M.; McGlaun, J.M.; Corradini, M.L.

    1983-01-01

    If there is not adequate cooling water in the core of a light-water reactor (LWR), the fission product decay heat would eventually cause the reactor fuel and cladding to melt. This could lead to slumping of the molten core materials into the lower plenum of the reactor vessel, possibly followed by failure of the vessel wall and pouring of the molten materials into the reactor cavity. When the molten core materials enter either region, there is a strong possibility of molten core contacting water. This paper focuses on analysis of recent FITS experiments, mechanistic and probabilistic model development, and the application of these models to reactor considerations.

  2. Severe fuel-damage scoping test performance. [PWR

    SciTech Connect (OSTI)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle.

  3. Relocation and freezing of liquefied fuel-rod material. [PWR

    SciTech Connect (OSTI)

    Moore, R.L.; Broughton, J.M.

    1982-01-01

    Severe degraded core cooling accidents, such as occurred at TMI-2 can potentially reach temperatures in excess of cladding melting. When the molten cladding is in contact with UO/sub 2/ fuel, the UO/sub 2/ will be dissolved contributing significantly to the total amount of liquefied material flowing down the rod and eventually freezing in a lower, cooler region of the core. The primary objectives of this paper are to evaluate the relocation and freezing characteristics of liquefied fuel rod material over a wide range of system conditions, physical characteristics of the fuel rod and liquefied material, and material thermo-physical properties to determine the relative influence of the controlling parameters. First the analytical model used in the analysis is briefly reviewed. The results of the analyses are then presented and discussed, and this is followed by the conclusions.

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

    SciTech Connect (OSTI)

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

    1982-03-01

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

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

    SciTech Connect (OSTI)

    Siefken, L.J.

    1983-01-01

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

  6. Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR

    SciTech Connect (OSTI)

    Chung, H.M.

    1983-06-01

    Zircaloy cladding from high-burnup (> 20 MWd/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.

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

    SciTech Connect (OSTI)

    Olsen, C.S.

    1982-10-01

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

  8. Real Time Pricing and the Real Live Firm

    SciTech Connect (OSTI)

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

    2004-05-26

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

  9. Superfund explanation of significant difference for the record of decision (EPA region 2): Love Canal, Niagara Falls, NY, September 5, 1996

    SciTech Connect (OSTI)

    1997-11-01

    The United States Environmental Protection Agency (EPA) and the New York State Department of Environmental Conservation (NYSDEC) announce this Explanation of Significant Differences (ESD) to explain modifications to the selected remedy for the final destruction and disposal of Love Canal dioxin-contaminated sewer and creek sediments. These modifications are embodied in proposed changes to a partial consent decree between the United States and the State of New York and the Occidental Chemical Corporation (OCC) in the United States District Court for the Western District of New York.

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

    Broader source: Energy.gov [DOE]

    Paper presented at the Waste Management 2012 Conference.February 26 through March 1, 2012, Phoenix, Arizona.

  11. EIS-0109: Final Environmental Impact Statement

    Broader source: Energy.gov [DOE]

    Long-Term Management of the Existing Radioactive Wastes and Residues at the Niagara Falls Storage Site

  12. Marinette County, Wisconsin: Energy Resources | Open Energy Informatio...

    Open Energy Info (EERE)

    Wisconsin Beecher, Wisconsin Coleman, Wisconsin Crivitz, Wisconsin Dunbar, Wisconsin Goodman, Wisconsin Marinette, Wisconsin Middle Inlet, Wisconsin Niagara, Wisconsin Pembine,...

  13. EIS-0358: Notice of Intent to Prepare an Environmental Impact...

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

    Wellton-Mohawk Generating Facility, Yuma County, Arizona, EIS-0358 May 19, 2003 (68 FR 27056) More Documents & Publications CX-004898: Categorical Exclusion Determination...

  14. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    SciTech Connect (OSTI)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  15. Effect of rod bow to partial closure on critical heat flux in PWR fuel assembly

    SciTech Connect (OSTI)

    Macbuff, R.B.; Fighetti, C.F.

    1983-07-01

    The effects of partial closure due to bowed rods on critical heat flux (CHF) in a pressurized water reactor rod bundle were evaluated by conducting tests in an electrically heated test section. The test section consisted of a 5x5 square rod array with 24 heated rods of 9.14 mm (0.360 in) diameter, each with a heated length of 3.66 m (12 ft) with Exxon Nuclear Company spacer grids on a 0.521 m (20.5 in) pitch. The central rod was a 12.2 mm (0.480 in) diameter unheated guide tube. The gap between two centrally located high powered rods was reduced 71% from nominal. The bow was approximately mid-span between two grids in the region in which CHF was observed in unbowed test sections. The results of these tests indicate a reduction in CHF of approximately 5% at 2400 psia and no reduction at 2000 psia when test repeatability is taken into consideration. The reduction in CHF at 2400 psia is substantially smaller than that inferred by linear interpolation of previously reported test results at 50 and 100% reduction of spacing. The local nature of CHF reduction due to rod bow was confirmed.

  16. Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys

    Broader source: Energy.gov [DOE]

    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.

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

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27

    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.

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

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20

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

  19. Surveillance of PLUS7{sup TM} fuel for PWR nuclear power plant...

    Office of Scientific and Technical Information (OSTI)

    1999. The irradiation tests for the in-reactor verification using four lead test ... After in-reactor verifications during two cycles, this fuel was commercially supplied to ...

  20. Optimization of small long-life PWR based on thorium fuel (Journal...

    Office of Scientific and Technical Information (OSTI)

    in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. ...

  1. Surveillance of PLUS7{sup TM} fuel for PWR nuclear power plant

    SciTech Connect (OSTI)

    Jang, Y. K.; Kim, J. I.; Shin, J. C.; Chung, J. G.; Chung, S. K.; Kim, M. S.; Lee, T. H.; Yoon, Y. B.; Kim, T. W.

    2012-07-01

    The surveillance program on the advanced nuclear fuel of PLUS{sup TM} developed for Optimized Power Reactors of 1000 MWe (OPR1000s) and Advanced Power Reactors of 1400 MWe (APR1400s) in Korea was completed in the early of 2011. This fuel had been jointly developed through the extensive out-of-pile tests with Westinghouse for three years since 1999. The irradiation tests for the in-reactor verification using four lead test assemblies (LTAs) had been started in Ulchin unit 3 in 2002. During the overhaul period after each irradiation test, the eight (8) burnup-dependent parameters were measured without disassembling using the precise measurement systems in pool-side. After three cycle irradiations, one test assembly was disassembled and the rod-wise inspection on twenty rods was performed. During this stage, five (5) parameters were measured and evaluated. Among these twenty rods, ten rods including skeleton were sent to hot-cell test facility for further detailed examination and are currently being examined. After in-reactor verifications during two cycles, this fuel was commercially supplied to eight (8) OPR1000s sequentially. Currently all eight (8) OPR1000s were replaced with this fuel. In addition, this fuel is going to be supplied to four (4) APR1400s being constructed in Braka, UAE as well as four(4) OPR1000s and four(4) APR1400s being constructed in Korea. In the meanwhile, the surveillance program for the commercially supplied fuel has been launched to confirm growth, creep, corrosion and deformation, etc. obtained during LTA irradiation. Four (4) limiting fuel assemblies, that is, two (2) assemblies to be discharged after 2 cycle irradiations and the other two (2) after 3 cycle irradiations were selected for this surveillance program. Irradiation data of commercially supplied fuels are compared and confirmed to LTA irradiation performance behaviors on this paper. Among the eight (8) burnup-dependent parameters, the interesting ones were irradiation-induced growths of grid width, rod and assembly lengths while cladding oxide thickness, rod creep and swelling followed the LTA results, and assembly bow/twist and rod bow were well below the design limits. The rod growths were on the lower bound of database and contributed to assembly growth. It was evaluated that these results were attributed to PLUS{sup TM} design characteristics which adopted Inconel top and bottom grids and conformal-shaped mid grid spring. (authors)

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

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

    DOCUMENT AVAILABILITY Online Access: U.S. Department of Energy (DOE) reports produced after 1991 and a growing number of pre-1991 documents are available free via DOE's SciTech ...

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

    SciTech Connect (OSTI)

    Matzie, R A

    1980-04-01

    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.

  4. Post-Service Examination of PWR Baffle Bolts, Part I. Examination and Test Plan

    SciTech Connect (OSTI)

    Leonard, Keith J.; Sokolov, Mikhail A.; Gussev, Maxim N.

    2015-04-30

    In support of extended service and current operations of the US nuclear reactor plants, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating with Ginna Nuclear Power Plant, The Westinghouse Electric Company, LLC, and ATI Consulting, the selective procurement of baffle bolts that were withdrawn from service in 2011 and currently stored on site at Ginna. The goal of this program is to perform detailed microstructural and mechanical property characterization of baffle former bolts following in-service exposures. This report outlines the selection criteria of the bolts and the techniques to be used in this study. The bolts available are the original alloy 347 steel fasteners used in holding the baffle plates to the baffle former structures within the lower portion of the pressurized water reactor vessel. Of the eleven possible bolts made available for this work, none were identified to have specific damage. The bolts, however, did show varying levels of breakaway torque required in their removal. The bolts available for this study varied in peak fluence (highest dose within the head of the bolt) between 9.9 and 27.8x1021 n/cm2 (E>1MeV). As no evidence for crack initiation was determined for the available bolts from preliminary visual examination, two bolts with the higher fluence values were selected for further post-irradiation examination. The two bolts showed different breakaway torque levels necessary in their removal. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long-term irradiation behavior and compared to results obtained in high flux experimental reactor conditions. The two bolts selected for the ORNL study will be shipped to Westinghouse with bolts of interest to their collaborative efforts with the Electric Power Research Institute. Westinghouse will section the ORNL bolts into samples specified in this report and return them to ORNL. Samples will include bend bars for fracture toughness and crack propagation studies along with thin sections from which specimens for bend testing, subscale tensile and microstructural analysis can be obtained. Additional material from the high stress concentration region at the transition between the bolt head and shank will also be preserved to allow for further investigation of possible crack initiation sites.

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

    SciTech Connect (OSTI)

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

    1980-01-01

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

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

    Office of Scientific and Technical Information (OSTI)

    ... are 77 and 5 vo Hp respectively, whereas at 100 psig and 300F the upper and lower ... mixture at one atm is composed of steam, the limits coincide at about 10 hydrogen. ...

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

    SciTech Connect (OSTI)

    Kim, Woong Sik; Jo, Jong Chull; Yune, Young Gill; Kim, Hho Jung

    2004-07-01

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

  8. Stress-corrosion cracking of Inconel alloy 600 in high-temperature water: an update. [PWR

    SciTech Connect (OSTI)

    Bandy, R.; van Rooyen, D.

    1983-01-01

    Inconel 600 has been tested in high-temperature aqueous media (without oxygen) in several tests. Data are presented to relate failure times to periods of crack initiation and propagation. Quantitative relationships have been developed from tests in which variations were made in temperature, applied load, strain rate, water chemistry, and the condition of the test alloy.

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

    SciTech Connect (OSTI)

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    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.

  10. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes. [PWR; BWR

    SciTech Connect (OSTI)

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping.

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

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

    Slide 7 Southwestern Power Administration early afternoon operating day prior. Customer Electrical Demand Loads Transmission System ConditionsOutages Electrical System Operations ...

  12. Specimen Machining for the Study of the Effect of Swelling on CGR in PWR Environment.

    SciTech Connect (OSTI)

    Teysseyre, Sebastien Paul

    2015-06-01

    This report describes the preparation of ten specimens to be used for the study of the effect of swelling on the propagation of irradiation assisted stress corrosion cracking cracks. Four compact tension specimens, four microscopy plates and two tensile specimens were machined from a AISI 304 material that was irradiated up to 33 dpa. The specimens had been machined such as to represent the behavior of materials with 3.7%swelling and <2% swelling.

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

    SciTech Connect (OSTI)

    Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

    2012-06-19

    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.

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

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    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.

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

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    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.

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

    SciTech Connect (OSTI)

    Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa; Kosaka, Yuji; Arakawa, Yasushi

    2007-07-01

    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)

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

    SciTech Connect (OSTI)

    Anikovsky, V.V.; Karzov, G.P.; Timofeev, B.T.

    1997-04-01

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

  18. Material control and accountancy at EDF PWR plants; GCN: Gestion du Combustible Nucleaire

    SciTech Connect (OSTI)

    de Cormis, F. )

    1991-01-01

    The paper describes the comprehensive system which is developed and implemented at Electricite de France to provide a single reliable nuclear material control and accounting system for all nuclear plants. This software aims at several objectives among which are: the control and the accountancy of nuclear material at the plant, the optimization of the consistency of data by minimizing the possibility of transcription errors, the fulfillment of the statutory requirements by automatic transfer of reports to national and international safeguards authorities, the servicing of other EDF users of nuclear material data for technical or commercial purposes.

  19. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    SciTech Connect (OSTI)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  20. Development of a new flux map processing code for moveable detector system in PWR

    SciTech Connect (OSTI)

    Li, W.; Lu, H.; Li, J.; Dang, Z.; Zhang, X.

    2013-07-01

    This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

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

    SciTech Connect (OSTI)

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

    1984-05-01

    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.

  2. Source terms: an investigation of uncertainties, magnitudes, and recommendations for research. [PWR; BWR

    SciTech Connect (OSTI)

    Levine, S.; Kaiser, G. D.; Arcieri, W. C.; Firstenberg, H.; Fulford, P. J.; Lam, P. S.; Ritzman, R. L.; Schmidt, E. R.

    1982-03-01

    The purpose of this document is to assess the state of knowledge and expert opinions that exist about fission product source terms from potential nuclear power plant accidents. This is so that recommendations can be made for research and analyses which have the potential to reduce the uncertainties in these estimated source terms and to derive improved methods for predicting their magnitudes. The main reasons for writing this report are to indicate the major uncertainties involved in defining realistic source terms that could arise from severe reactor accidents, to determine which factors would have the most significant impact on public risks and emergency planning, and to suggest research and analyses that could result in the reduction of these uncertainties. Source terms used in the conventional consequence calculations in the licensing process are not explicitly addressed.

  3. SCDAP severe core-damage studies: BWR ATWS and PWR station blackout

    SciTech Connect (OSTI)

    Laats, E.T.; Chambers, R.; Driskell, W.E.

    1983-01-01

    The Severe Accident Sequence Analysis (SASA) Program, sponsored by the US Nuclear Regulatory Commission (NRC), is addressing a number of accident scenarios that potentially pose a health hazard to the public. Two of the scenarios being analyzed in detail at the Idaho National Engineering Laboratory (INEL) are the station blackout at the Bellefonte nuclear plant and the anticipated transient without scram (ATWS) at the Browns Ferry-1 plant. The INEL analyses of the station blackout and ATWS have been divided into four parts, which represent the sequence being followed in this study. First, the evaluation of long term irradiation effects prior to the station blackout or ATWS was conducted using the FRAPCON-2 fuel rod behavior code; second, the reactor primary and secondary coolant system behavior is being analyzed with the RELAP5 code; third, the degradation of the core is being analyzed with the SCDAP code; and finally, the containment building response is being analyzed with the CONTEMPT code. This paper addresses only the SCDAP/MODO degraded core analyses for both the station blackout and ATWS scenarios.

  4. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    SciTech Connect (OSTI)

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

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

    SciTech Connect (OSTI)

    DeHart, M.D.

    1999-08-01

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

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

    Office of Scientific and Technical Information (OSTI)

    CONTRACT A T - I M - G E N - H BETTIS PLANT PITTSBURGH, PENNSYLVANIA Operated for the ... e t y of the P W R Reactor Plant ( S h i p p i n g p o r t Atomic Power S t a t i o n ) . ...

  7. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    SciTech Connect (OSTI)

    Radulescu, Georgeta

    2010-06-01

    Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensing Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with predicted spent fuel compositions (i.e., determine the penalty in reactivity due to isotopic composition bias and uncertainty) for use in disposal criticality analysis employing burnup credit. The method used in this calculation to propagate the isotopic bias and bias-uncertainty values to k{sub eff} is the Monte Carlo uncertainty sampling method. The development of this report is consistent with 'Test Plan for: Isotopic Validation for Postclosure Criticality of Commercial Spent Nuclear Fuel'. This calculation report has been developed in support of burnup credit activities for the proposed repository at Yucca Mountain, Nevada, and provides a methodology that can be applied to other criticality safety applications employing burnup credit.

  8. Isotopic Generation and Confirmation of the PWR Application Model 

    SciTech Connect (OSTI)

    L.B. Wimmer

    2003-11-10

    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.

  9. Analytical model for transient gas flow in nuclear fuel rods. [PWR; BWR

    SciTech Connect (OSTI)

    Rowe, D.S.; Oehlberg, R.N.

    1981-08-01

    An analytical model for calculating gas flow and pressure inside a nuclear fuel rod is presented. Such a model is required to calculate the pressure loading of cladding during ballooning that could occur for postulated reactor accidents. The mathematical model uses a porous media (permeability) concept to define the resistance to gas flow along the fuel rod. 7 refs.

  10. Variations in Zircaloy-4 cladding deformation in replicate LOCA simulation tests. [PWR

    SciTech Connect (OSTI)

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

    1982-09-01

    Five single-rod, heated-shroud replicate burst tests were conducted to study statistical variations in Zircaloy cladding deformation under simulated loss-of-coolant accident conditions. The test conditions used (low steam coolant flow and a heating rate of approx. 10 K/s to tube failure at approx. 775/sup 0/C) were conductive to large deformation and matched those used in two of the Multirod Burst Test Program bundle tests so that the results could be used to aid in interpretation of differences observed for individual rods in bundle tests. The largest variation observed was in burst strain, which ranged from 50 to 96%. Burst temperature ranged from 767 to 779/sup 0/C, burst pressure from 9405 to 9870 kPa, average strain over the heated length from 18 to 23%, and tube volume increase from 39 to 51%. As expected, cladding deformation was influenced by small temperature gradients: the more uniform the temperature, the greater (and more uniform) the deformation.

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

    SciTech Connect (OSTI)

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

    1984-07-01

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

  12. Multirod burst test program. Progress report, July-December 1979. [BWR; PWR

    SciTech Connect (OSTI)

    Chapman, R.H.

    1980-08-01

    A series of scoping tests designed to explore the effect of shroud heating on Zircaloy cladding deformation was conducted in the single-rod test facility, which was recently modified to permit independent heating of the shroud under specified conditions. To facilitate comparison of the test results, the series included tests under conditions used previously. Significantly greater deformation was observed in heated shroud tests than would be expected from unheated stroud tests. Fabrication of fuel pin simulators for the B-5 (8 x 8) bundle test continued with approx.90% of the required number being completed. Five fuel pin simulators, identical to the simulators used in the Japanese Atomic Energy Research Institute multirod bundle burst tests, were delivered by the Japanese manufacturer. The surface temperature distribution of the simulators was characterized for several heating rates by infrared scanning and was compared to similar characterizations of Oak Ridge National Laboratory simulators. Plans are under way for conducting burst tests on the Japanese simulators in the single-rod test facility. 14 refs., 116 figs., 3 tabs.

  13. Experiment prediction for LOFT nuclear experiments L5-1/L8-2. [PWR

    SciTech Connect (OSTI)

    Chen, T.H.; Modro, S.M.

    1982-01-01

    The LOFT Experiments L5-1 and L8-2 simulated intermediate break loss-of-coolant accidents with core uncovery. This paper compares the predictions with the measured data for these experiments. The RELAP5 code was used to perform best estimate double-blind and single-blind predictions. The double-blind calculations are performed prior to the experiment and use specified nominal initial and boundary conditions. The single-blind calculations are performed after the experiment and use measured initial and boundary conditions while maintaining all other parameters constant, including the code version. Comparisons of calculated results with experimental results are discussed; the possible causes of discrepancies are explored and explained. RELAP5 calculated system pressure, mass inventory, and fuel cladding temperature agree reasonably well with the experiment results, and only slight changes are noted between the double-blind and single-blind predictions.

  14. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    SciTech Connect (OSTI)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs.

  15. Thermal-hydraulics of the PFB/LOFT lead rod loss-of-coolant experiments. [PWR

    SciTech Connect (OSTI)

    Varacalle, D.J. Jr.; Garner, R.W.; MacDonald, P.E.; Cox, W.R.

    1980-01-01

    Results of the four PBF/LOFT Lead Rod sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods subjected to a series of nuclear blowdown tests, and to determine if subjecting deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature versus system pressure response with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements. Tests LLR-3, LLR-5, LLR-4, and LLR-4A were performed at system conditions of 595/sup 0/K coolant inlet temperature, 15.5 MPa system pressure, and 41, 46, 57 and 56 kW/m test rod peak linear powers, respectively, at initiation of blowdown. Cladding temperatures during the tests ranged from 870 to 1260/sup 0/K.

  16. FRAP-T5 predictions during reactor shutdown events. [PWR; BWR

    SciTech Connect (OSTI)

    Peeler, G.B.; Laats, E.T.

    1980-01-01

    The Transient Fuel Rod Analysis Program, FRAP-T5, was recently assessed by EG and G Idaho, Inc. As part of this assessment, the measured and FRAP-T5 predicted fuel centerline temperature response during reactor shutdown events were compared. For these events either forced convection or nucleate boiling boundary conditions existed, resulting in a negligible effect on fuel behavior from cladding temperature and deformation uncertainties. This allowed the assessment of internal heat transfer to be emphasized.

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

    SciTech Connect (OSTI)

    Edler, S.K.

    1985-02-01

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

  18. Simulated dry storage test of a spent PWR nuclear fuel assembly in air

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.; Oden, D.R.; Stidham, D.L.; Garnier, J.E.; Weeks, D.L.; Dobbins, J.C.

    1985-02-01

    The purpose of the dry storage test was to investigate the behavior of Zircaloy-clad spent fuel in air between 200 and 275/sup 0/C. Atmospheric air was used for the cover gas because of the interest in establishing regimes where air inleakage into an initially inert system would not cause potential fuel degradation. Samples of the cover gas atmosphere were extracted monthly to determine fission gas concentrations as a function of time. The oxygen concentration was monitored to detect oxygen depletion, which would signal oxidation of the fuel. The gas analyses indicated very low but detectable levels of /sup 85/Kr during the first month of the test. A large increase (five orders of magnitude) in /sup 85/Kr and the appearance of helium in the cover gas indicated that a fuel rod had breached during the second month of the test. Stress rupture calculations showed that the stresses and temperatures were too low to expect breaches to form in defect-free cladding. It is theorized that the breach occurred in a fuel rod weakened by an existing cladding or end cap defect. Calculations based on the rate of /sup 85/Kr release suggest that the diameter of the initial breach was about 25 microns. A post-test fuel examination will be performed to locate and investigate the cause of the cladding breach and to determine if detectable fuel degradation progressed after the breach occurred. The post-test evaluation will define the consequences of a fuel rod breach occurring in an air cover gas at 270/sup 0/C, followed by subsequent exposure to air at a prototypic descending temperature.

  19. Microstructural characteristics of PWR spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1985-11-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  20. Experiment data report for Multirod Burst Test (MRBT) Bundle B-5. [PWR

    SciTech Connect (OSTI)

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

    1984-08-01

    A reference source of MRBT bundle B-5 test data is presented with interpretation limited to that necessary to understand pertinent features of the test. Primary objectives of this 8 x 8 multirod burst test were to investigate the effects of array size and rod-to-rod interactions on cladding deformation in the high-alpha-Zircaloy temperature range under simulated light-water reactor loss-of-coolant accident (LOCA) conditions. B-5 test conditions, nominally the same as used in an earlier 4 x 4 (B-3) test, simulated the adiabatic heatup (reheat) phase of an LOCA and were conducive to large deformation. The fuel pin simulators were electrically heated (average linear power generation of 3.0 kW/m) and were slightly cooled with a very low flow (Re approx. 140) of low-pressure superheated steam. The cladding temperature increased from the initial temperature (335/sup 0/C) to the burst temperature at a rate of 9.8/sup 0/C/s. The simulators burst in a very narrow temperature range, with an average of 768/sup 0/C. Cladding burst strain ranged from 32% to 95%, with an average of 61%. Volumetric expansion over the heated length of the cladding ranged from 35% to 79%, with an average of 52%. The results clearly show deformation was greater in the bundle interior and suggest rod-to-rod mechanical interactions caused axial propagation of the deformation.

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

    SciTech Connect (OSTI)

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

    1980-01-01

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

  2. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  3. Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR

    SciTech Connect (OSTI)

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

    1982-01-01

    Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive to large deformation were selected in order to favor rod-to-rod interactions. The tests showed that rod-to-rod interactions play an important role in the deformation process.

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

    SciTech Connect (OSTI)

    Bailey, R.T.

    1982-04-01

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

  5. Experiment operations plan for the MT-4 experiment in the NRU reactor. [PWR

    SciTech Connect (OSTI)

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700/sup 0/F).

  6. LOCA analyses for nuclear steam supply systems with upper head injection. [PWR

    SciTech Connect (OSTI)

    Byers, R.K.; Bartel, T.J.

    1980-01-01

    The term Upper Head Injection describes a relatively new addition to a nuclear reactor's emergency cooling system. With this feature, water is delivered directly to the top of the reactor vessel during a loss-of-coolant accident, in addition to the later injection of coolant into the primary operating loops. Established computer programs, with various modifications to models for heat transfer and two-phase flow, were used to analyze a transient following a large break in one of the main coolant loops of a reactor equipped with upper head injection. The flow and heat transfer modifications combined to yield fuel cladding temperatures during blowdown which were as much as 440K (800/sup 0/F) lower than were obtained with standard versions of the codes (for best estimate calculations). The calculations also showed the need for more uniformity of applications of heat transfer models in the computer programs employed.

  7. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors. [PWR; BWR

    SciTech Connect (OSTI)

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper.

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

    SciTech Connect (OSTI)

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

    1982-01-01

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

  9. Comparisons of the SCDAP computer code with bundle data under severe accident conditions. [PWR; BWR

    SciTech Connect (OSTI)

    Allison, C.M.; Beers, G.H.

    1983-01-01

    The SCDAP computer code, which is being developed under the sponsorship of the United States Nuclear Regulatory Commission, models the progression of light water reactor core damage including core heatup, core disruption and debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data. Comparisons between SCDAP calculations and the experimental data showed good agreement. Calculated and measured bundle temperatures for SFD-ST were within 200 K for the entire bundle and within 20 K for maximum cladding temperatures. For ESSI-2, calculated and measured maximum cladding temperatures were within 50 K, and the extensive liquefaction and relocation that was calculated was in agreement with experimental results.

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

    SciTech Connect (OSTI)

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

    1982-09-01

    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.

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

    SciTech Connect (OSTI)

    Edler, S.K.

    1984-06-01

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

  12. Advanced digital PWR plant protection system based on optimal estimation theory

    SciTech Connect (OSTI)

    Tylee, J.L.

    1981-04-01

    An advanced plant protection system for the Loss-of-Fluid Test (LOFT) reactor plant is described and evaluated. The system, based on a Kalman filter estimator, is capable of providing on-line estimates of such critical variables as fuel and cladding temperature, departure from nucleate boiling ratio, and maximum linear heat generation rate. The Kalman filter equations are presented, as is a description of the LOFT plant dynamic model inherent in the filter. Simulation results demonstrate the performance of the advanced system.

  13. Long-term, low-temperature oxidation of PWR spent fuel: Interim transition report

    SciTech Connect (OSTI)

    Einziger, R.E.; Buchanan, H.C.

    1988-05-01

    Since some of the fuel rods will be breached and eventually most of the cladding will corrode, exposing fuel, one factor influencing the ability of spent fuel to retain radionuclides is its oxidation state in the expected moist air atmosphere. Oxidation of the fuel could split the cladding, exposing additional fuel and changing the leaching characteristics. Thermodynamically, there is no reason why UO{sub 2} should not oxidize completely to UO{sub 3} at repository temperatures. The underlying uncertainty is the rate of oxidation. Extrapolation of higher temperature data indicates that insufficient oxidation to convert all of the fuel to U{sub 3}O{sub 8} will occur during the first 10,000 years. However, lower oxidation states, such as U{sub 4}O{sub 9} and U{sub 3}O{sub 7}, might form. To date, the tests have run between 3200 and 4100 hours out of a planned 16,000-hour duration. Some preliminary conclusions can be drawn: (1) Moisture content of the air has no significant effect on oxidation rate, (2) the data have an uncertainty of 15 to 20%, which must be accounted for in the interpretation of single sample tests, and (3) below 175{degree}C, the oxidation rate is dependent on the particle size in the sample. The smaller particles oxidize more rapidly. 19 refs., 23 figs., 7 tabs.

  14. TRAC-PF1/MOD1 US/Japanese PWR conservative LOCA prediction

    SciTech Connect (OSTI)

    Gruen, G E; Fisher, J E

    1987-11-01

    This report documents the results of a 200%, double-ended, cold-leg-break, loss-of-coolant-accident (LOCA) calculation using the TRAC-PF1/MOD1 computer code. The reactor system represented a typical United States/Japanese pressurized water reactor with a 15 x 15 fuel bundle arrangement 12-ft long, four loops, and cold-leg Emergency Core Cooling (ECC) Systems. Conservation boundary and initial conditions were used. Reactor power was 102% of the 3250 MWt rated power, decay heat was set to 120% of American Nuclear Society Standard 5.1, highest core lifetime values for power peaking and fuel stored energy were used, and the LOCA occurred simultaneously with a loss of offsite power. Best estimate assumptions were used for the break flow model, fuel rod heat transfer and metal-water reaction correlations, and steady-state fuel temperature profiles. A flow blockage model, having the capability to account for the effects of cladding ballooning or rupturing, was not used. Except for these best estimate assumptions, the boundary and initial conditions were consistent with those used in licensing calculations. Maximum fuel rod temperatures were 1380 K (2020/sup 0/F) and 1040 K (1410/sup 0/F) on the hottest evaluation model rod and hottest best estimate rod, respectively. The high reported values or fuel cladding temperature were a direct consequence of the conservative boundary and initial conditions used for the calculation, primarily the 2% overpower condition, the core decay heat assumption, and the degraded ECCS. The calculation demonstrated successful core reflooding before 1478 K (2200/sup 0/F) cladding temperature was exceeded on any fuel rod. 7 refs., 47 figs., 5 tabs.

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

    SciTech Connect (OSTI)

    Christensen, R.

    1982-04-01

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

  16. Calculations conducted in developing an audit capability for ECCS analysis. [PWR

    SciTech Connect (OSTI)

    Bartel, T.J.; Berman, M.; Byers, R.K.; Cole, R.K. Jr.

    1981-12-01

    This study has demonstrated the capability of combining the results of thermal-hydraulic and fuel rod response computer codes to produce audit-type calculations for a pressurized water reactor equipped with a relatively new form of emergency core cooling systems. Models intended specifically for use with such systems were incorporated into the codes, sample calculations were performed, and very cursory comparisons with vendor-supplied results were made. In calculations of the blowdown phase of a large break loss-of-coolant accident, models for fuel rod surface quenching and for separated two-phase flow were observed to have significant effects on peak cladding temperatures and on system conditions at the beginning of core reflood. Models used for the reflood phase, particularly the model for carryover-rate fraction, were also seen to have important consequences. While the demonstration of audit capability was successful, there remain questions connected with details of coupling between the codes, and with uniformity of models as used in all phases of the calculations.

  17. Multirod burst test program progress report, January-June 1982. Final report. [PWR

    SciTech Connect (OSTI)

    Chapman, R.H.

    1982-12-01

    The B-6 (8 x 8) array was tested, and posttest examination was completed; data reduction and analysis are in progress. Preliminary quick-look results are included in this report. All 64 rods were pressurized and burst. The average burst temperature was 931/sup 0/C, and the bundle average heating rate was 3.5/sup 0/C/s during the time of deformation. Preliminary results indicate burst strains ranged from 22 to 56%, with a bundle average of 30%. Analysis of the B-5 test results continues to provide insight to the complexity of cladding deformation in bundles, particularly for conditions conducive to large deformation and rod-to-rod interactions. Additional analyses, including re-evaluation of burst temperatures, are included in this report. The B-6 test concluded the experimental phase of this research program. Future activities will be concerned with analysis and evaluation of experimental data produced by this and other research programs.

  18. FRAP-T6 calculations of fuel-rod behavior during overpower transients. [PWR; BWR

    SciTech Connect (OSTI)

    Chambers, R.; Resch, S.C.

    1982-01-01

    The performance of the FRAP-T6 computer code in calculating fuel rod failure and fission gas release during overpower transient events was analyzed. Comparisons of the code's calculations with experiment data was used to determine the accuracy of the code in these two performance areas. First, the ability of the code to replicate observed failure trends as functions of power, ramp rate, hold time, burnup, pellet-cladding gap size, cladding thickness, and fuel density was examined. Then, the capability of the code's fission gas release model to duplicate experiment measurements of unfailed rods was tested at various burnups.

  19. Experiment operations plan for the TH-2 experiment in the NRU reactor. [PWR; BWR

    SciTech Connect (OSTI)

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for TH-2--the second experiment in the series of thermal-hydraulic tests conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The major objective of TH-2 was to develop the experiment reflood control parameters and the procedures to be used in subsequent experiments in this program. In this experiment, the data acquisition and control system was used to control the fuel cladding temperature during a simulated LOCA by using variable reflood coolant flow.

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

    SciTech Connect (OSTI)

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

    1987-01-01

    For a severe pressurized water reactor accident that leads to a loss of feedwater to the stream generators, such as might occur in a station blackout, fission product decay heating causes a water boil-off. Without effective decay heat removal, the fuel elements will be uncovered. Eventually, steam will oxidize the overheated cladding. The noble gases and volatile fission products, such as cesium and iodine, 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.

  1. Evaluation of prompt release of fission gas from a breached cladding. [PWR; BWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Kumar, R.M.

    1981-01-01

    It is a concern in the current safety analysis of nuclear reactors to understand the different release mechanisms of fission products to accurately determine the radiological source term for a wide variety of accidents. The mechanism which is least understood and which produces an uncertainty in determining the radiological source term during a reactor accident is the early release of fission gas present in the fuel-cladding gap through a cladding breach. In a loss-of-coolant type accident the fuel rods would be surrounded mainly by steam, therefore, the release of the gap gas can simply be treated as a discharge problem through an orifice. However, during reactor normal operation or in those accidents where the failed fuel rods are surrounded by liquid coolant, the release process of the gap gas would be strongly influenced by the coolant conditions (pressure, temperature and flow rate). The purpose of this work is to describe analytically the prompt escape of volatiles and gaseous fission products, present in the fuel-cladding gap through a cladding breach, where the fuel rod is surrounded by liquid coolant.

  2. Assessment of SPEAR-FCODE-BETA for fuel licensing. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Bradley, E.R.; Lanning, D.D.; Fiero, I.B.; Freeburn, H.R.; Garde, A.M.; Krammen, M.A.; Rotondo, P.L.

    1983-03-01

    The EPRI FCODE-BETA fuel performance code is the mechanistic model portion for the SPEAR-BETA fuel reliability code, which was developed for EPRI by Entropy Limited. FCODE-BETA was assessed by Battelle, Pacific Northwest Laboratories, and Combustion Engineering, Inc., for adequacy in simulating fuel performance in fuel licensing proceedings. The assessment included a detailed review of those thermal and mechanical performance models of greatest importance to fuel licensing, including fission gas release, fuel temperatures, and cladding uniform strain. It was concluded that the code is inadequate for licensing calculations in its present form. Recommendations for modeling and usability improvements are made.

  3. Lifetime of PWR silver-indium-cadmium control rods. Final report

    SciTech Connect (OSTI)

    Sipush, P.J.; Woodcock, J.; Chickering, R.W.

    1986-03-01

    A hot cell examination was performed on selected rodlets of a lead rod cluster control assembly (RCCA) which had experienced eleven cycles of operation in Point Beach Unit 1. The principal purpose of the program was to evaluate the performance of RCCAs. The hot cell examination of the rodlets involved detailed visual inspections, profilometry, metallography, cladding chemistry, dosimetry, scanning electron microscopy, corrosion tests, microhardness tests, absorber density measurements, and cladding tensile tests. Wear scars and a hairline crack in the cladding were evaluated. The results of the examinations and analysis of WEPCO site photographs led to an estimate of the service life for RCCAs which are used in Westinghouse 14 x 14 fuel assemblies. Also, wear scar widths were correlated with wear scar depths. The correlation may be used to estimate wear scar depths based on site photographs of wear scars for 14 x 14 RCCAs. The results of the program may be used as guidelines for RCCAs for 15 x 15 and 17 x 17 Westinghouse fuel designs. 10 refs., 89 figs., 26 tabs.

  4. Accelerated high-temperature tests with spent PWR and BWR fuel rods under dry storage conditions

    SciTech Connect (OSTI)

    Porsch, G.; Fleisch, J.; Heits, B.

    1986-09-01

    Accelerated high-temperature tests on 25 intact pressurized water and boiling water reactor rods were conducted for more than 16 months at 400, 430, and 450/sup 0/C in a helium gas atmosphere. The pretest characterized rods were examined by nondestructive methods after each of the three test cycles. No cladding breaches occurred and the creep deformation remained below 1%, which was in good agreement with model calculations. The test atmospheres were analyzed for /sup 85/Kr and tritium. The /sup 85/Kr concentrations were negligible and the tritium release agreed with the theoretical predictions. It can be concluded that for Zircaloy-clad fuel, cladding temperatures up to 450/sup 0/C are acceptable for dry storage in inert cover gases.

  5. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment. [BWR; PWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO/sub 2/ fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm/sup 3//s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO/sub 2/ fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%.

  6. Determination of post-DNB and post-BT fuel design limits. [PWR; BWR

    SciTech Connect (OSTI)

    Croucher, D.W.; Loyd, R.J.

    1980-01-01

    Categories of light water reactor transients and the departure from nucleate boiling (DNB) and boiling transition (BT) fuel design limits in light water reactors are reviewed. These fuel design limits for reactor licensing may be overly conservative because experiments have shown that fuel rods do not fail and may not experience damage as a result of momentary operation in film boiling or dryout conditions. Damage to the fuel rod is strongly dependent on the peak cladding temperature and the length of time at that temperature durng the transient. Testing of two potential licensing fuel design limits is suggested: (a) fuel rod functional capabilities are retained and fuel system dimensions remain within operational telerances; and (b) cladding deformation is permitted, but no significant oxidation is allowed. Damage mechanisms which may affect post-DNB or post-BT operation of fuel rods are permanent rod bowing and pellet-cladding interaction. The data necessary to support a fuel design limit and a means of obtaining these data are outlined.

  7. EPRI/B and W cooperative program on PWR fuel-rod performance. Final report

    SciTech Connect (OSTI)

    Papazoglou, T.P.; Davis, H.H.

    1983-03-01

    Zircaloy-4 fuel cladding specimens were irradiated in a fueled and non-fueled condition for two and four cyles of irradiation, respectively, in the Oconee 2 reactor. The purpose of this long-term surveillance program was to study the in-reactor performance of four Zircaloy-4 cladding types with distinctly different properties, in combination with two types of UO/sub 2/ fuel pellets. The cladding types included Sandvik Special Metals tubing in the cold-worked/stress relieved and cold-worked/recrystallized conditions, and German VDM cladding with two different anneal temperatures. The fuel pellets included a conventional densifying pellet type, and a special (shorter) stable pellet type intended to reduce pellet-clad mechanical interaction. The irradiation growth and creep under compressive stress of the above cladding types were studied and followed up to fluences of 1.3 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV).

  8. Data summary report for fission product release test HI-1. [PWR; BWR

    SciTech Connect (OSTI)

    Osborne, M.F.; lorenz, R.A.; Travis, J.R.; Webster, C.S.

    1982-12-01

    This first in a series of high-temperature fission product release tests was conducted for 30 min at 1400/sup 0/C, with the release taking place into flowing steam. The fuel specimen was a 20-cm-long section of H.B. Robinson fuel rod, irradiated to 28,000 MWd per metric ton (t). After the test, the Zircaloy cladding of the specimen was almost completely oxidized and was quite fragile. The fission product collection system included a thermal gradient tube (700-150/sup 0/C), filters, heated charcoal, and cooled charcoal. Gamma ray analysis of apparatus components and collectors showed that about 2.83% of the /sup 85/Kr and 1.75% of the /sup 137/Cs were released from the fuel. Activation analysis of leach solutions from these components indicated that 2.04% of the /sup 129/I was released. Other analyses revealed small but significant releases of the radionuclides /sup 125/Sb and /sup 106/Ru, and of the elements Br, Rb, Sr, Zr, Ag, Sn, Te, Ba, and La.

  9. Cladding axial elongation models for FRAP-T6. [PWR; BWR

    SciTech Connect (OSTI)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented.

  10. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  11. Fuel Performance Improvement Program. Semiannual progress report, October 1979-March 1980. [PWR; BWR

    SciTech Connect (OSTI)

    Not Available

    1980-04-01

    Progress on the Fuel Performance Improvement Program's fuel design tests and demonstration irradiations for October 1979 through March 1980 is reported. Included are the results of out-of-reactor experiments with Zircaloy cladding using a device that simulates the interaction between fuel and cladding. Also included are reports on the irradiation of the advanced LWR fuel designs in the Halden Boiling Water Reactor and in Consumers Power Company's Big Rock Point Reactor. The establishment of the technical bases and licensing requirements for the advanced fuel concepts are also described.

  12. Feasibility of on-line fuel-condition monitoring. [PWR; BWR

    SciTech Connect (OSTI)

    Petti, D.A.; Osetek, D.J.; Croucher, D.W.; Hartwell, J.K.

    1982-01-01

    The relationship between fuel rod damage and fission product release is investigated to assess the feasibility of using on-line gamma spectroscopy of reactor coolant to estimate not only numbers of detected fuel rods, but also the type of core damage which may occur during an accident or off-normal transient. Fission product release signatures for various fuel conditions and accident scenarios are compared, and unique indicators of fuel damage, ranging from cladding pinholes to severely damaged fuel rods, are suggested, The configuration of monitoring hardware and data analysis soft ware are described, and the benefits, development needs, and usefulness of the envisaged power plant system are discussed.

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

    SciTech Connect (OSTI)

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

    1982-01-01

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

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

    SciTech Connect (OSTI)

    McCormick, R.D.

    1983-01-01

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

  15. Experiment data report for Multirod Burst Test (MRBT) bundle B-4. [PWR; BWR

    SciTech Connect (OSTI)

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

    1982-12-01

    A compilation of bundle B-4 test data is presented. These data were obtained during the test and from pretest and posttest examination of the test array. They are presented in considerable detail but with minimum interpretation. The B-4 test is the only 6 x 6 array in a series of 4 x 4, 6 x 6, and 8 x 8 bundle tests performed by the Multirod Burst Test Program at Oak Ridge National Laboratory. This research is sponsored by the Nuclear Regulatory Commission and is designed to investigate Zircaloy cladding deformation behavior under simulated light-water-reactor loss-of-coolant accident conditions. A brief description of the experiment and a summary of the test results are included with the detailed results of the B-4 test. Both graphical and tabular formats are used to show temperature and pressure data as functions of test time and strain data for the cladding in each of the fuel rod simulators. Photographic documentation is provided for both the overall bundle, before and after testing, and the 36 tubes as they were removed from the tested bundle for strain measurements.

  16. Effects of Lower Drying-Storage Temperatures on the DBTT of High Burnup PWR Cladding

    Broader source: Energy.gov [DOE]

    The purpose of the research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor cladding alloys during cooling for a range of storage temperatures and hoop stresses.

  17. Generating human reliability estimates using expert judgment. Volume 2. Appendices. [PWR; BWR

    SciTech Connect (OSTI)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results.

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

    SciTech Connect (OSTI)

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

    2005-08-25

    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

  19. The War of the Currents: AC vs. DC Power | Department of Energy

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

    Power Company decided to award Westinghouse -- who had licensed Tesla's polyphase AC induction motor patent -- the contract to generate power from Niagara Falls. Although some...

  20. Separations Process Research Unit (SPRU) Site Cleanup By the...

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

    at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program Enforcement Letter, Safety and Ecology Corporation - NEL-2011-04

  1. untitled

    Office of Legacy Management (LM)

    Niagara Falls Storage Site M:LTS111000110000S0824300S0824300-26.mxd coatesc 09282011 12:51:34 PM 0 1,000 500 FEET...

  2. . The Honora+ W

    Office of Legacy Management (LM)

    ... Los Alamos, New Mexico' Bayo Canyon* Los Alamos, New Mexico University,of California. Berkeley, California Chupadera kesa : Middlesex Municipal Landfill Niagara Falls Storage Site ...

  3. Formerly Utilized Sites Remedial Action Program Fact Sheet

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

    Ammunition Plant Seaway Industrial Park Linde Air Products Tonawanda Landfill Niagara Falls Storage Site Guterl Specialty Steel Colonie Shallow Land Disposal Area Shpack...

  4. Formerly Utilized Sites Remedial Action Program Fact Sheet

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

    ... USACE determined the site requires additional remediation and accepted it as an active site. Since then, DOE has also referred sites in Brooklyn, New York, and Niagara Falls ...

  5. DOE - Office of Legacy Management -- New York

    Office of Legacy Management (LM)

    York New York NY Sites 2013 Buffalo Site New York Site Niagara Falls Storage Site Vicinity Properties Tonawanda North Site, Unit 1 Tonawanda North Site, Unit 2

  6. Hydrologic Tracer Studies Conducted August 20 - 25, 1962 Near...

    Office of Legacy Management (LM)

    ... , Mohawk R i v e r a t K n o l l s Atomic Power a b o r a t o r - y , N. Y . , and Sedan ... t h e Mohawk R i v e r a t Knolls Atomic Power Laboratory, N. Y e , and T r i b u t a r y ...

  7. CX-000219: Categorical Exclusion Determination

    Office of Energy Efficiency and Renewable Energy (EERE)

    United States Army Corps Niagara River, New York Small HydropowerCX(s) Applied: A9, A11Date: 11/30/2009Location(s): Niagara River, New YorkOffice(s): Energy Efficiency and Renewable Energy, Golden Field Office

  8. Axial blanket fuel design and demonstration. First semi-annual progress report, January-September 1980. [PWR

    SciTech Connect (OSTI)

    Not Available

    1980-11-01

    The axial blanket fuel design in this program, which is retrofittable in operating pressurized water reactors, involves replacing the top and bottom of the enriched fuel column with low-enriched (less than or equal to 1.0 wt % /sup 235/U) fertile uranium. This repositioning of the fissile inventory in the fuel rod leads to decreased axial leakage and increased discharge burnups in the enriched fuel. Various axial blanket fuel designs, with blanket thicknesses from 0 to 10 inches and blanket enrichments from 0.2 to 1.0 wt % /sup 235/U, were investigated to determine the relationship between uranium utilization and power peaking. Analyses were preformed to assess the nuclear, mechanical, and thermal-hydraulic effects arising from the use of axial blankets. Four axial blanket lead test assemblies are being fabricated for scheduled irradiation in cycle 5 of Sacramento Municipal Utility District's Rancho Seco pressurized water reactor. Analyses to support licensing cycle 5 are in progress.

  9. A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's.

    Energy Science and Technology Software Center (OSTI)

    1990-12-01

    Version 00 SACHET evaluates the dynamic fission (FP) product inventories in the multiple compartment system of pressurized water reactor plants.

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

    SciTech Connect (OSTI)

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

    2006-02-28

    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.

  11. Determining fissile content in PWR spent fuel assemblies using a passive neutron Albedo reactivity with fission chambers technique

    SciTech Connect (OSTI)

    Conlin, Jeremy Lloyd; Tobin, Stephen J

    2010-01-01

    State regulatory bodies and organizations such as the IAEA that are concerned with preventing the proliferation of nuclear weapons are interested in a means of quantifying the amount of plutonium in a given spent fuel assembly. The complexity of spent nuclear fuel makes the measurement of plutonium content challenging. There are a variety of techniques that can measure various properties of spent nuclear fuel including burnup, and mass of fissile content. No single technique can provide all desired information, necessitating an approach using multiple detector systems and types. This paper presents our analysis of the Passive Neutron Albedo Reactivity Fission Chamber (PNAR-FC) detector system. PNAR-FC is a simplified version of the PNAR technique originally developed in 1997. This earlier research was performed with a high efficiency, {sup 3}He-based system (PNAR-3He) with which multiplicty analysis was performed. With the PNAR technique a portion of the spent fuel assembly is wrapped in a 1 mm thick cadmium liner. Neutron count rates are measured both with and without the cadmium liner present. The ratio of the count rate with the cadmium liner to the count rate without the cadmium liner is calculated and called the cadmium ratio. In the PNAR-3He technique, multiplicity measurements were made and the cadmium ratio was shown to scale with the fissile content of the material being measured. PNAR-FC simplifies the PNAR technique by using only a few fission chambers instead of many {sup 3}He tubes. Using a simplified PNAR-FC technique provides for a cheaper, lighter, and thus more portable detector system than was possible with the PNAR-3He system. The challenge with the PNAR-FC system are two-fold: (1) the change in the cadmium ratio is weaker as a afunction of the changing fissile content relative to multiplicity count rates, and (2) the efficiency for the fission chamber based system are poorer than for the {sup 3}He based detectors. In this paper, we present our research on using the PNAR-FC detector system to quantify the fissile content of a spent nuclear fuel assembly.

  12. STRESS CORROSION CRACK GROWTH RESPONSE FOR ALLOY 152/52 DISSIMILAR METAL WELDS IN PWR PRIMARY WATER

    SciTech Connect (OSTI)

    Toloczko, Mychailo B.; Olszta, Matthew J.; Overman, Nicole R.; Bruemmer, Stephen M.

    2015-08-15

    As part of ongoing research into primary water stress corrosion cracking (PWSCC) susceptibility of alloy 690 and its welds, SCC tests have been conducted on alloy 152/52 dissimilar metal (DM) welds with cracks positioned with the goal to assess weld dilution and fusion line effects on SCC susceptibility. No increased crack growth rate was found when evaluating a 20% Cr dilution zone in alloy 152M joined to carbon steel (CS) that had not undergone a post-weld heat treatment (PWHT). However, high SCC crack growth rates were observed when the crack reached the fusion line of that material where it propagated both on the fusion line and in the heat affected zone (HAZ) of the carbon steel. Crack surface and crack profile examinations of the specimen revealed that cracking in the weld region was transgranular (TG) with weld grain boundaries not aligned with the geometric crack growth plane of the specimen. The application of a typical pressure vessel PWHT on a second set of alloy 152/52 – carbon steel DM weld specimens was found to eliminate the high SCC susceptibility in the fusion line and carbon steel HAZ regions. PWSCC tests were also performed on alloy 152-304SS DM weld specimens. Constant K crack growth rates did not exceed 5x10-9 mm/s in this material with post-test examinations revealing cracking primarily on the fusion line and slightly into the 304SS HAZ.

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

    SciTech Connect (OSTI)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2005-10-15

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

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

    SciTech Connect (OSTI)

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    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.

  15. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    SciTech Connect (OSTI)

    Langenbuch, S.; Velkov, K.; Lizorkin, M.

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  16. Stress corrosion cracking of Alloy 600 and Alloy 690 in all volatile treated water at elevated temperatures. Final report. [PWR

    SciTech Connect (OSTI)

    Theus, G.J.; Emanuelson, R.H.

    1983-05-01

    This report describes a continuing study of stress corrosion cracking (SCC) of Inconel alloys 600 and 690 in all-volatile treated (AVT) water. Specimens of alloys 600 and 690 are being exposed to AVT water at 288/sup 0/, 332/sup 0/, 343/sup 0/, and 360/sup 0/C. Alloy 600 generally resists SCC in high-purity water under normal service conditions but is susceptible under other specific conditions. In general, mill-annealed alloy 600 is more susceptible than stress-relieved material. Susceptibility to SCC increases rapidly with increasing exposure temperature. Very high stresses (near or above yield) are required to induce cracking in AVT or other high-purity waters. Most of the data presented in this report are for alloy 600; alloy 690 has not yet cracked. However, the program is being continued and will subsequently characterize the high-purity water cracking behavior, if any, of alloy 690.

  17. 1984 Workshop on secondary-side stress corrosion cracking and intergranular corrosion of PWR steam generator tubing. Proceedings

    SciTech Connect (OSTI)

    Partridge, M.J.

    1986-03-01

    The contractors' meeting on ''Secondary Water Initiated IGC of Alloy 600'' was organized in an effort to give those working in this area an opportunity to share their results, ideas and plans. Topics discussed included: (1) transport phenomena and crevice chemistry; (2) correlations and models of electrochemical and isothermal tests; (3) chemistry of carbides and surface films; (4) model boiler and heat transfer studies; and (5) summary experience and test results regarding IGA/IGSCC rates.

  18. Initiation and propagation of stress-corrosion cracking of Alloy 600 in high-temperature water. [PWR

    SciTech Connect (OSTI)

    Bandy, R.; van Rooyen, D.

    1983-01-01

    Results of stress-corrosion cracking data are presented for Inconel 600 steam-generator tubing. U-bend, constant-load, and slow extension-rate tests are included. Arrhenius plots are presented for failure times vs inverse temperature for crack initiation and propagation. Effect of applied load is expressed in terms of log-log curves for failure times vs stress, and variations in environment and cold work are included. Microstructure and composition of oxide films on Inconel 600 surfaces were examined after exposure to pure water at 365/sup 0/C, and stripping with the bromine-methanol method. Results are discussed in terms of transient creep, film rupture and a mass-transport-limited anodic process.

  19. Stress corrosion of alloys 600 and 690 in acidic sulfate solutions at elevated temperatures. Final report. [PWR

    SciTech Connect (OSTI)

    Newman, J.F.

    1983-10-01

    EPRI project RP 1171-1 demonstrated that alloy 600 was susceptible to stress corrosion cracking in sulfate environments during constant extension rate testing. The current project has extended that investigation to determine the influence of alloy grain size, sensitization, solution pH and temperature. Stress corrosion in very dilute sulfate solutions and the susceptibility of alloy 690 have also been studied. Data have been obtained chiefly by constant extension rate testing using tensile specimens, and by the constant displacement testing of C-rings.

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

    SciTech Connect (OSTI)

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

    1984-01-01

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

  1. Comparison of COMETHE-IIIJ and FCODE-BETA fission gas-release predictions with measurements. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Lee, S.; Rayes, L.; Rumble, E.; Wheeler, D.; Woods, A.

    1983-03-01

    This report describes a comparison of the Fission Product Gas Release (FGR) predictability of two LWR fuel rod modeling codes: COMETHE-IIIJ and FCODE-BETA. The comparison is made using 124 well characterized fuel rods with FGR measurements in the EPRI Fuel Performance Data Base.

  2. Mechanistic prediction of fission-product release under normal and accident conditions: key uncertainties that need better resolution. [PWR; BWR

    SciTech Connect (OSTI)

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.

  3. SPEAR-BETA fuel-performance code system: fission-gas-release module. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Christensen, R.

    1983-03-01

    The original SPEAR-BETA general description manual covers both mechanistic and statistical models for fuel reliability, but only mechanistic modeling of fission gas release. This addendum covers the SPEAR-BETA statistical model for fission gas release.

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

    SciTech Connect (OSTI)

    Pontillon, Y.; Noirot, J.; Caillot, L.

    2007-07-01

    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)

  5. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    SciTech Connect (OSTI)

    Swinhoe, Martyn Thomas; De Baere, Paul

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

  6. Implementation of the active neutron Coincidence Collar for the verification of unirradiated PWR and BWR fuel assemblies

    SciTech Connect (OSTI)

    Menlove, H.O.; Keddar, A.

    1982-01-01

    An active neutron interrogation technique has been developed for the measurement of the /sup 235/U content in fresh fuel assemblies. The method employs an AmLi neutron source to induce fission reactions in the fuel assembly and coincidence counting of the resulting fission reaction neutrons. When no interrogation source is present, the passive neutron coincidence rate gives a measure of the /sup 238/U by the spontaneous fission reactions. The system can be applied to the fissile content determination in fresh fuel assemblies for accountability, criticality control, and safeguards purposes. Field tests have been performed by International Atomic Energy Agency (IAEA) staff using the Coincidence Collar to verify the /sup 235/U content in light-water-reactor fuel assemblies. The results gave an accuracy of 1 to 2% in the active mode (/sup 235/U) and 2 to 3% in the passive mode (/sup 238/U) under field conditions.

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

    SciTech Connect (OSTI)

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

    1982-09-01

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

  8. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

    SciTech Connect (OSTI)

    Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

    1981-08-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations.

  9. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 4: seismic response analysis. Final report

    SciTech Connect (OSTI)

    Lu, S.C.; Ma, S.M.; Larder, R.A.

    1981-09-01

    This volume of the report gives a detailed account of the seismic response analysis of the primary coolant loop piping of Unit 1 of the Zion Nuclear Power Station. Because the purpose of this work was to perform a realistic simulation, best estimate loads and material properties were used for the calculation whenever possible. When such data were unavailable, conservative values were used. The calculation procedure included the generation of seismic input, the determination of dynamic soil properties, a three-part soil-structure-piping interaction analysis, and the post-response data procession. A large number of variables considered in the analysis can affect the seismic response stresses. This volume therefore describes a sensitivity study, as well as the method of analysis. The sensitivity study is included to establish confidence in the computed response stresses.

  10. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 6: failure mode analysis. Final report

    SciTech Connect (OSTI)

    Streit, R.D.

    1981-09-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic flow requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  11. Temperature estimates from the Zircaloy oxidation kinetics in the. cap alpha. plus. beta. phase region. [PWR; BWR

    SciTech Connect (OSTI)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of Zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of Zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near the thermocouple locations.

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

    SciTech Connect (OSTI)

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

    1984-01-01

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

  13. Evaulation of power-reactor fuel-rod-analysis capabilities. Phase 1 topical report. Volume 2. Code evaluation. [PWR; BWR

    SciTech Connect (OSTI)

    Coleman, D.R.

    1983-09-01

    FRAPCON-2 (V1M4) was applied to generate fuel performance predictions for 60 rods of a recently evaluated power reactor data sample. Rod design, operational, and performance data was obtained from the RPRI Fuel Performance Data Base. The data was systematically processed to generate code input parameters. FRAPCON was initially applied for scoping studies to identify the best estimate mechanical response and fission gas release modeling options. Based on final scoping results, the balance of rods were analyzed with FRACAS-2 mechanics and FASTGRASS gas release models. Comparisons between measured and calculated fuel and cladding deformation, fission gas release, internal pressure, and gas composition are presented and interpreted relative to code error magnitudes, distributions, and trends versus rod design and operating parameters. The results indicate the FRAPCON-2 has best estimate capability for analysis of moderate duty fuel rod performance, provided that rod fabrication parameters are well characterized, and the fuel is dimensionally stable.

  14. Evaluation of flow redistribution due to flow blockage in rod bundles using COBRA code simulation. Final report. [PWR

    SciTech Connect (OSTI)

    Prelewicz, D.A.; Caruso, M.A.

    1981-01-01

    During a Loss-of-Coolant Accident, fuel rod cladding may reach temperatures approaching 2200/sup 0/F. At these temperatures, swelling and rupture of the cladding may occur. The resulting flow blockage will affect steam flow and heat transfer in the bundle during the period of reflooding. The COBRA-IV-I subchannel computer code was used to simulate flow redistribution due to sleeve blockages in the FLECHT-SEASET 21-rod bundle and plate blockages in the JAERI Slab Core Test Facility. Sensitivity studies were conducted to determine the effects of spacer grid and blockage interaction, sleeve shape effects, sleeve length effects, blockage magnitude and distribution, thermally induced mixing and bundle average velocity on flow redistribution. Pressure drop due to sleeve blockages was also calculated for several blockage configurations.

  15. FRAP-T6: a computer code for the transient analysis of oxide fuel rods. [PWR; BWR

    SciTech Connect (OSTI)

    Siefken, L.J.; Shah, V.N.; Berna, G.A.; Hohorst, J.K.

    1983-06-01

    FRAP-T6 is a computer code which is being developed to calculate the transient behavior of a light water reactor fuel rod. This report is an addendum to the FRAP-T6/MODO user's manual which provides the additional user information needed to use FRAP-T6/MOD1. This includes model changes, improvements, and additions, coding changes and improvements, change in input and control language, and example problem solutions to aid the user. This information is designed to supplement the FRAP-T6/MODO user's manual.

  16. Temperature estimates from the zircaloy oxidation kinetics in the. cap alpha. plus. beta. phase region. [PWR; BWR

    SciTech Connect (OSTI)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near thermocouple locations.

  17. Description and assessment of structural and temperature models in the FRAP-T6 code. [PWR; BWR

    SciTech Connect (OSTI)

    Siefken, L.J.

    1983-01-01

    The FRAP-T6 code was developed at the Idaho National Engineering Laboratory (INEL) for the purpose of calculating the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to severe hypothetical loss-of-coolant accidents. An important application of the FRAP-T6 code is to calculate the structural performance of fuel rod cladding. The capabilities of the FRAP-T6 code are assessed by comparisons of code calculations with the measurements of several hundred in-pile experiments on fuel rods. The results of the assessments show that the code accurately and efficiently models the structural and thermal response of fuel rods.

  18. Analysis of fission gas release measurements using the COMETHE IIIJ and FCODE-Alpha computer codes. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Leppert, G.; Rayes, L.; Rumble, E.; Stuart, R.

    1981-07-01

    Fission gas release predictions from FCODE-Alpha and COMETHE IIIJ were compared with experimental data from a representative group of light water reactor (LWR) fuel rods and with each other. In the first phase of the study, standard versions of the codes obtained from the Electric Power Software Center were compared with data from 36 rods. A modified version of COMETHE was used in the second phase of the study, which compared measurements from some of the same rods studied in the first phase, as well as with an additional 27 rods. Fission gas release predictions from both codes show substantial deviation from experimental measurements, and additional well-qualified data from LWR's is needed for comparison. Unpressurized rods experience significant degradation in heat transfer across the fuel-to-cladding gap as the lower thermal conductivity fission gases mix with the helium.

  19. Heavy-section steel technology program. Quarterly progress report, October-December 1982. Volume 4. [PWR; BWR

    SciTech Connect (OSTI)

    Whitman, G.D.; Pugh, C.E.; Bryan, R.H.

    1983-05-01

    The investigation focuses on the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into seven tasks: (1) program administration and procurement, (2) fracture-mechanics analyses and investigations, (3) investigations of irradiated materials, (4) thermal-shock investigations, (5) pressure vessel investigations, (6) stainless steel cladding investigations, and (7) environmentally assisted crack growth studies. A superposition solution technique for determining stress-intensity factors for semielliptical surface cracks in cylinders was implemented in pressurized thermal-shock (PTS) analyses. Subcontractors continued studies on crack arrest, cleavage fracture initiation, and cleavage transition. Specimens of the ORNL single-wire cladding were fabricated for irradiation. Pretest analyses were carried out for the upcoming thermal-shock test, TSE-7, and posttest analyses and examinations were under way for intermediate vessel test ITV-8A. Preparations for the first PTS experiment continued with design, procurement, and construction of the test facility, test vessels, and experimental apparatus.

  20. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect (OSTI)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  1. FRAP-T6 uncertainty study of LOCA tests LOFT L2-3 and PBF LLR-3. [PWR

    SciTech Connect (OSTI)

    Chambers, R.; Driskell, W.E.; Resch, S.C.

    1983-01-01

    This paper presents the accuracy and uncertainty of fuel rod behavior calculations performed by the transient Fuel Rod Analysis Program (FRAP-T6) during large break loss-of-coolant accidents. The accuracy of the code was determined primarily through comparisons of code calculations with cladding surface temperature measurements from two loss-of-coolant experiments (LOCEs). These LOCEs were the L2-3 experiment conducted in the Loss-of-Fluid Test (LOFT) Facility and the LOFT Lead Rod 3 (LLR-3) experiment conducted in the Power Burst Facility (PBF). Uncertainties in code calculations resulting from uncertainties in fuel and cladding design variables, material property and heat transfer correlations, and thermal-hydraulic boundary conditions were analyzed.

  2. Effects of high temperature and flow blockage on the reflood behavior of a 4-rod bundle. Final report. [PWR

    SciTech Connect (OSTI)

    Drucker, M.; Dhir, V.K.

    1981-11-01

    It is usual in reactor safety analysis to assume that blocking or deforming the reactor core decreases the heat removal. This simplistic approach may not only penalize reactor power, but must be investigated experimentally to determine the real extent, if any. The experiments reported here examine quenching and heat removal in a blocked four-rod bundle. The local heat transfer in the blockage region is enhanced, despite the flow diversion away from the blockage. Additionally, data and correlations are given which compare the quenching rate of steel pins (typical of experiments) with Zircaloy (typical of reactor cladding). The Zircaloy bundle quenches faster when correlated on a local basis because of its smaller heat capacity. Additional work is under way to explain and correlate the intriguing results in more detail.

  3. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    SciTech Connect (OSTI)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  4. CX-014324: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Dome Tap-Wellton Mohawk Ligurta - Replace Structure CX(s) Applied: B1.3Date: 07/01/2015 Location(s): ArizonaOffices(s): Western Area Power Administration-Desert Southwest Region

  5. CX-014325: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Dome-Wellton-Mohawk-Ligurta Line Maintenance CX(s) Applied: B1.3Date: 09/14/2015 Location(s): ArizonaOffices(s): Western Area Power Administration-Desert Southwest Region

  6. CX-000034: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Saint Regis Mohawk Tribe Energy Efficiency RetrofitsCX(s) Applied: B5.1, B2.5Date: 11/02/2009Location(s): New YorkOffice(s): Energy Efficiency and Renewable Energy

  7. CX-004898: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

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

  8. operation_tbl3_October_2011M.xlsx

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

    Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 1011968 8171974 5202014 212000 6202001 5202034 Arkansas Nuclear One 2 PWR Combustion Eng. 711971 12261978 7...

  9. CASL-U-2015-0167-000 COBRA-TF

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

    Parallelization and Application to PWR Reactor Core Vefa Kucukboyaci and Yixing Sung ... PARALLELIZATION AND APPLICATION TO PWR REACTOR CORE SUBCHANNEL DNB ANALYSIS Vefa ...

  10. State Nuclear Profiles 2010

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

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

  11. Detection and monitoring of air emissions and emergency response planning within three geographic areas

    SciTech Connect (OSTI)

    Not Available

    1985-01-01

    This report gives the results of air emissions and emergency response planning in the following areas: Baton Rouge/New Orleans; Philadelphia/Wilmington/South Jersey; and Niagara Falls/Buffalo.

  12. PLEAEERUSH ANALYTICAL DA-~-A SHEET

    Office of Legacy Management (LM)

    ...lectedby-CESS-.Route to CBS LocationTITANIUM Type of Sample airnalyzed for F Alpham Remarks NIAGARA pALI+S* N.Y. U Beta Bldg. 103 - furnace room - -NO, Ra Oil PH Be Th Sample No. ...

  13. - United States Government

    Office of Legacy Management (LM)

    illiams, 903-8149) : NY 41 I .' 41 G I? SUBJECT: Elimination of the T itanium Alloy Manufacturing Co., Niagara Falls, New York TO: The F ile I have reviewed the attached site....

  14. Canrom Photovoltaics Inc | Open Energy Information

    Open Energy Info (EERE)

    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...

  15. U.S. Natural Gas Exports to Mexico

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

    MI St. Clair, MI Noyes, MN Warroad, MN Babb, MT Havre, MT Port of Morgan, MT Sherwood, ND Pittsburg, NH Buffalo, NY Grand Island, NY Massena, NY Niagara Falls, NY Waddington, NY...

  16. Department of Energy Announces Tougher Criteria for ENERGY STAR®...

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

    The Department of Energy projects that this new criteria could result in savings of over ... each year, equivalent to the amount of water that flows over Niagara Falls in four hours. ...

  17. THE AEROSPACE CORPORATION ,'

    Office of Legacy Management (LM)

    ... lands reputedly owned by the United States Air Force; and on the west by certain lands now ... Buffalo, N. Y. CHEI-TROL POLLUTION SERVICES, INC. (Reputed Tenant) 1 Niagara Square ...

  18. Unique nature of hydroplant complicates design

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    This article describes the unique nature of the Niagara Power Project as it affected upgrading of the Robert Moses powerplant and operation of the powerplant and pumped storage facility. To be taken into account are the variable flow of the Niagara River, treaties regarding division of river flow between Canada and USA and maintenance of river flow over the falls, and the level of Lake Erie.

  19. Microsoft Word - S06246_VP_Report

    Office of Legacy Management (LM)

    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

  20. Oak Ridge Associated VERIFICATION Universities OF OF

    Office of Legacy Management (LM)

    y /I/ i Prepared by Oak Ridge Associated VERIFICATION Universities OF OF - Prepared for the Decontamination and 1985 AND 1986 REMEDIAL ACTIONS Decommissioning Division NIAGARA FALLS STORAGE SITE U.S. Department VICINITY PROPERTIES of Energy of Energy LEWISTON, NEW YORK 3I~~~~ ~~~J. D. BERGER l I I I I I Environmental Survey and Site Assessment Program Energy/Environment Systems Division FINAL REPORT JULY 1990 VERIFICATION OF 1985 AND 1986 REMEDIAL ACTIONS NIAGARA FALLS STORAGE SITE VICINITY

  1. FORMERLY UTILIZED SITES REMEDIAL ACTION PROGRAM ELIMINATION REPORT

    Office of Legacy Management (LM)

    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),

  2. U.S. Spent Nuclear Fuel Data as of December 31,2002 -Table 2

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

    2024 Three Mile Island 1 PA PWR B&W 177 1974 2014 Trojan OR PWR WE 193 1975 2011 1992 Turkey Point 3 FL PWR WE 157 1972 2032 Turkey Point 4 FL PWR WE 157 1973 2033 Vermont Yankee...

  3. Instrument fieldsclose.sdr

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

    Radar Van GP Van BBSS Cart RWP Met Twr ECOR Radiometry Stands PWR Module 25m

  4. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  5. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    SciTech Connect (OSTI)

    Wissinger, G.; Klingenfus, J.

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  6. The Effect of Accident Conditions on the Molten Core Material Relocation into the Lower Head of a PWR Vessel with Application to TMI-2

    SciTech Connect (OSTI)

    An Xuegao; Dhir, Vijay K.; Okrent, David

    2000-11-15

    The damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out on a Three Mile Island Unit 2 configuration using the computer code SCDAP/RELAP5/MOD3.2.Different accident scenarios were simulated. The high-pressure injection and makeup flow rates were changed. The extreme case with no water being added during the accident was examined. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the power-operated relief valve opening was also changed. The effects of these accident scenarios on the accident progression and the core damage process were studied.It is concluded that, according to code MOD3.2, the molten material slumped to the lower head of the reactor vessel when the junction of the top and side crusts failed after the molten pool had reached the periphery of the core. When the effective stress caused by pressure imbalance inside and outside of the crust became larger than the ultimate strength of the crust, the crust failed mechanically.

  7. Response of PWR Baffle-Former Bolt Loading to Swelling, Irradiation Creep and Bolt Replacement as Revealed Using Finite Element Modeling

    SciTech Connect (OSTI)

    Simonen, Edward P.; Garner, Francis A.; Klymyshyn, Nicholas A.; Toloczko, Mychailo B.

    2005-10-01

    Baffle-former bolts in pressurized water reactors (PWRs) tend to degrade with aging, partially due to radiation-induced hardening and also due to the often complex stress history of the bolt in response to time-dependent and spatial gradients in temperature and neutron flux-spectra that can alter the stress distribution of the bolts. The time-integrated stresses must play some role in bolt cracking, however, and therefore it is of interest to study the time dependence of bolt stresses even for idealized cases. These stresses have been quantified in the present analysis using newly developed material constitutive equations for swelling and creep at light-water reactor (LWR)-relevant temperatures and dose rates. ABAQUS finite element calculations demonstrate that irradiation creep in the absence of void swelling tends to relax bolt tension before 10 dpa. Subsequent differential swelling leads to an increase in bolt tension, but only to stresses below the yield strength and usually below the initial bolt loading. Various assumed bolt replacement scenarios are considered with respect to their consequences on future failure possibilities.

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

    SciTech Connect (OSTI)

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

    1982-03-01

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

  9. Intergranular stress-corrosion cracking of Ni-Cr-Fe Alloy 600 tubes in PWR primary water - review and assessment for model development. Final report

    SciTech Connect (OSTI)

    Garud, Y.S.; Gerber, T.L.

    1983-05-01

    The service performance of Ni-Cr-Fe Alloy 600 tubes with respect to primary-side intergranular stress-corrosion cracking (IGSCC) and related experimental observations are reviewed, assessed, and summarized. The qualitative trends suggested by these observations and the role of primary variables of IGSCC are discussed from the point of view of mechanistic considerations. The need for a quantitative model to predict the IGSCC is established. The review and assessment indicate that, in addition to the water chemistry and metallurgical variables, explicit consideration of stress (and attendant deformation, i.e., strain-rate) is essential for any evaluation of the IGSCC to be complete. A promising approach of quantitative model development is presented with the above background. The analytical-experimental work needed for the model development is suggested, and implementation of the model and potential benefits to utilities is briefly discussed.

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

    SciTech Connect (OSTI)

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

    1996-05-01

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

  11. Stress relief to prevent stress corrosion in the transition region of expanded Alloy 600 steam-generator tubing. Final report. [PWR

    SciTech Connect (OSTI)

    Woodward, J.; van Rooyen, D.

    1983-05-01

    The feasibility of preventing primary side roll transition cracking has been investigated, using induction heating to attain stress relief of expanded Ni-Cr-Fe Alloy 600 steam generator tubing. Work on rolled tubing and U-bends has shown that temperatures with which stress relief can be obtained range from 700 to 850/sup 0/C, with lower temperatures in this range requiring longer times at temperature to provide the requisite reduction in residual stresses. No work has yet been done outside this range. Preliminary tests, using induction heating, have been carried out on a mock tube sheet assembly, designed to the dimensions of a typical steam generator, and have identified the type of heating/cooling cycle that would occur in the tube sheet during a stress relief operation. Preliminary results show that the times to reach the higher temperatures in the range observed to give stress relief, of the order of 850/sup 0/C, can be as short as 8 seconds, and less with optimum coil design and power control.

  12. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9: PRAISE computer code user's manual. Final report

    SciTech Connect (OSTI)

    Lim, E.Y.

    1981-08-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor.

  13. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 4. Seismic response analysis. Load Combination Program Project I final report

    SciTech Connect (OSTI)

    Lu, S.C.; Ma, S.M.; Larder, R.A.

    1981-06-01

    This volume of the report gives a detailed account of the seismic response analysis of the primary coolant loop piping of Unit 1 of the Zion Nuclear Power Station. Because the purpose of this work was to perform a realistic simulation, best estimate loads and material properties were used for the calculation whenever possible. When such data were unavailable, conservative values were used. The calculation procedure included the generation of seismic input, the determination of dynamic soil properties, a three-part soil-structure-piping interaction analysis, and the post-response data procession. A large number of variables considered in the analysis can affect the seismic response stresses. This volume therefore describes a sensitivity study, as well as the method of analysis. The sensitivity study is included to establish confidence in the computed response stresses. 19 refs., 28 figs., 6 tabs.

  14. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 7. System failure probability analysis. Load Combination Program Project I final report

    SciTech Connect (OSTI)

    George, L.; Mensing, R.

    1981-06-01

    This volume describes the computational methodology used to estimate the probability of a simultaneous occurrence of an earthquake and a primary coolant loop pipe fracture caused directly by an earthquake for a pressurized water reactor. Point estimates of this probability, based on a simulation experiment, and the probabilities of related events are included. Simulation is used to estimate weld fracture probabilities conditional on a crack initially existing and an earthquake of specified intensity occurring at a specified time in the life of the plant. These estimates are combined with probabilities associated with the occurrence of an earthquake and the existence of a crack to obtain an estimate of the probability of simultaneous earthquake and pipe fracture for the entire primary coolant loop piping system. A point estimate of probability, as outlined in this volume, does not fully take into consideration all of the uncertainties associated with an analysis of this type. Uncertainty analysis, confidence interval estimates, and sensitivity measures better reflect potential uncertainties. These topics are discussed. Finally, a discussion of the use of a risk-based, rather than a probability-based, decision criterion for deciding whether to decouple is included. 13 refs., 7 figs., 6 tabs.

  15. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 6. Failure mode analysis load combination program. Project I, final report

    SciTech Connect (OSTI)

    Streit, R.D.

    1981-06-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  16. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9. PRAISE computer code user's manual. Load Combination Program Project I final report

    SciTech Connect (OSTI)

    Lim, E.Y.

    1981-06-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor. PRAISE has a very modular structure and can be tailored to a variety of crack growth and piping reliability problems. Although PRAISE was developed on a CDC-7600 computer, it was, however, coded in standard FORTRAN IV and is readily transportable to other machines.

  17. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 3. Nonseismic stress analysis. Load Combination Program, Project I final report

    SciTech Connect (OSTI)

    Chan, A.L.; Lu, S.C.; Rybicki, E.F.; Curtis, D.J.

    1981-06-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations. 13 refs., 16 figs., 11 tabs.

  18. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO/sub 2/ burnup. [PWR; BWR

    SciTech Connect (OSTI)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500/sup 0/C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO/sub 2/ of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines.

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

    SciTech Connect (OSTI)

    Mincey, J.F.

    1980-01-01

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

  20. Application of analytical capability to predict rapid cladding cooling and quench during the blowdown phase of a large break loss-of-coolant accident. [PWR

    SciTech Connect (OSTI)

    Aksan, S.N.; Tolman, E.L.; Nelson, R.A.

    1983-01-01

    Large-break Experiments L2-2 and L2-3 conducted in the Loss-of-Fluid Test (LOFT) facility experienced core-wide rapid quenches early in the blowdown transients. To further investigate rapid cladding quenches, separate effects experiments using Semiscale solid-type electric heater rods were conducted in the LOFT Test Support Facility (LTSF) over a wide range of inlet coolant conditions. The analytical capability to predict the cladding temperature response from selected LTSF experiments estimated to bound the hydraulic conditions causing the LOFT early blowdown quenches was investigated using the RELAP4 computer code and was shown to be acceptable over the film boiling cooldown phase. This analytical capability was then used to investigate the behavior of nuclear fuel rods under the same hydraulic conditions. The calculations show that, under rapid cooling conditions, the behaviors of nuclear and electrical heater rods are significantly different because the nuclear rods are conduction limited, while the electrical rods are convection limited.

  1. Analysis of the PBF in-pile large-break LOCA test results with FRAP-T6/BALON-2. [PWR

    SciTech Connect (OSTI)

    Broughton, J.M.; Golden, D.W.; Hagrman, D.L.

    1982-01-01

    A series of four, large-break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility (PBF) at the Idaho Engineering Laboratory. These experiments have been analyzed by using out-of-pile data to understand the phenomenology of zircaloy cladding ballooning and to construct a mechanistic computer code to describe cladding deformation and failure. The code was then used to quantify the influence of rod internal pressure, cladding heatup, and cladding circumferential temperature differences on ballooning and rupture for fresh and irradiated test rods in the PBF. The analysis indicates that the timing and magnitude of cladding circumferential temperature differences are the primary controlling parameters. Both the experimental and the analytical results support the hypothesis that previously irradiated rods exhibit greater cladding strain at failure than do fresh rods because of small local temperature differences within the cladding.

  2. Corrosion and hydriding performance evaluation of three Zircaloy-2 clad fuel assemblies after continuous exposure in PWR cores 1 and 2 at Shippingport, PA. Addendum. LWBR Development Program

    SciTech Connect (OSTI)

    Hillner, E.

    1983-12-01

    The cladding from one additional Zircaloy-2 clad fuel rod from the pressurized water reactor at Shippingport, Pa. was destructively examined for corrosion film thickness and hydrogen accumulation. These additional examinations were conducted primarily to determine whether or not the hydrogen pickup ratio (..delta..H/..delta..O) increased with increasing neutron exposure, as had been suggested by the results from earlier studies on these fuel rods. The current results indicate that the hydrogen pickup ratio for Zircaloy-2 does not change with increasing neutron exposure and suggest that some of the earlier reported data may be anomolous.

  3. Comparison of GAPCON-THERMAL-3 and FRAPCON-2 fuel-performance codes to in-reactor measurement of elastic cladding deformation. [PWR; BWR

    SciTech Connect (OSTI)

    Lanning, D.D.; Rausch, W.N.; Williford, R.E.

    1981-01-01

    A revision of the GAPCON-3 computer code became part of the NRC-sponsored FRAPCON-2 code. This paper presents a comparison of both codes to in-reactor data from IFA-508, a 3-rod test rig in the Halden Reactor, Norway, which features simultaneous measurements of fuel temperature, power, axial elongation, and diametral strain. The modeling revisions included putting all regions of the fuel in contact with cladding at all time, but assigning non-linear, spatially dependent, anisotropic elastic moduli to the fuel on an incremental load step basis. The moduli are functions of the local available void within the cladding. These concepts bring demonstrable improvement to the code predictions.

  4. TRAC-PF1/MOD1 analysis of a 200% cold-leg break in a US/Japanese PWR with four loops and 15 x 15 fuel

    SciTech Connect (OSTI)

    Spore, J.W.; Cappiello, M.W.

    1986-01-01

    This report presents the results of a TRAC-PF1/MOD1 calculation that simulated a 200% double-ended cold-leg-break loss-of-coolant accident in a generic US/Japanese pressurized water reactor. This is a best-estimate analysis using conservative boundary conditions and minimum safeguards. The calculation shows that the peak cladding temperature (PCT) occurs during blowdown and that the core reheat is minimal during reflood. The results also show that for an evaluation-model peak rod linear power of 15.85 kW/ft, a PCT of 1084 K is reached at 3.5 s into the blowdown transient, which is approx.394 K below the design basis limit of 1478 K. 10 figs.

  5. OE/EV-0005/2 Formerly Utilized MED/AEC Sites Remedial Action Program

    Office of Legacy Management (LM)

    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

  6. Microsoft Word - NY.17-16.doc

    Office of Legacy Management (LM)

    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)

  7. / J8Y.17 I E(DE86008418)

    Office of Legacy Management (LM)

    /,, 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

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

    Office of Legacy Management (LM)

    '^ 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,

  9. # Energy Measuremenfs Group

    Office of Legacy Management (LM)

    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

  10. 1993index.PDF

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

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

  11. Extended Version of the WIMS 69-group Library.

    Energy Science and Technology Software Center (OSTI)

    1990-12-17

    Version 01 The WIMSLIB-IJSO data library can be used for PWR calculations. The materials - Ag, In, Cd - were added to the original data for PWR control rods.

  12. Arizona Nuclear Profile - Palo Verde

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

    expiration date" 1,"1,311","9,308",81.0,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" 2,"1,314","11,653",101.2,"PWR","applicationvnd.ms-excel","application...

  13. Virginia Nuclear Profile - Surry

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

    expiration date" 1,839,"6,206",84.4,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" 2,799,"6,966",99.5,"PWR","applicationvnd.ms-excel","application...

  14. Tennessee Nuclear Profile - Sequoyah

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

    Expiration Date" 1,"1,152","8,962",88.8,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" 2,"1,126","8,792",89.2,"PWR","applicationvnd.ms-excel","application...

  15. Arkansas Nuclear Profile - Arkansas Nuclear One

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

    ...vnd.ms-excel" 2,993,"8,416",96.7,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"1,835","15,023",93.5 "Data for 2010" "PWR Pressurized Light Water Reactor."

  16. South Carolina Nuclear Profile - Oconee

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

    ...vnd.ms-excel" 3,846,"6,779",91.5,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,538","20,943",94.2 "Data for 2010" "PWR Pressurized Light Water Reactor."

  17. Tennessee Nuclear Profile - Watts Bar Nuclear Plant

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

    expiration date" 1,"1,123","9,738",99.0,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"1,123","9,738",99.0 "Data for 2010" "PWR Pressurized Light Water Reactor."

  18. Georgia Nuclear Profile - Vogtle

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

    ...vnd.ms-excel" 2,"1,152","9,363",92.8,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,302","19,610",97.2 "Data for 2010" "PWR Pressurized Light Water Reactor."

  19. New York Nuclear Profile - Indian Point

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

    ...vnd.ms-excel" 3,"1,040","8,995",98.7,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,063","16,321",90.3 "Data for 2010" "PWR Pressurized Light Water Reactor."

  20. Connecticut Nuclear Profile - Millstone

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

    ...vnd.ms-excel" 3,"1,233","9,336",86.4,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,103","16,750",90.9 "Data for 2010" "PWR Pressurized Light Water Reactor."

  1. EIA - State Nuclear Profiles

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

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

  2. CX-000033: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Saint Regis Mohawk Tribe Energy Efficiency and Conservation Programs for Buildings and FacilitiesCX(s) Applied: B5.1, A9Date: 11/02/2009Location(s): New YorkOffice(s): Energy Efficiency and Renewable Energy

  3. CX-009805: Categorical Exclusion Determination

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  4. State Nuclear Profiles 2010

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

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

  5. Overview of the energy from a waste facility at Occidental Chemical

    SciTech Connect (OSTI)

    Blasius, G.F.

    1985-01-01

    The startup and operational problems and solutions concerned with processing and burning MSW to produce steam and electricity at Occidental's Niagara Falls chemical complex are reviewed. The facility was designed to burn 2000 tons per day of municipal waste, and produce 600,000number/HR steam and 37 mw of electricity.

  6. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect (OSTI)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    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.

  7. Adequacy of Power-to-Mass Scaling in Simulating PWR Incident Transient for Reduced-Height, Reduced-Pressure and Full-Height, Full-Pressure Integral System Test Facilities

    SciTech Connect (OSTI)

    Liu, T.-J.; Lee, C.-H

    2004-03-15

    A complete scheme of scaling methods to design the reduced-height, reduced-pressure (RHRP) Institute of Nuclear Energy Research Integral System Test (IIST) facility and to specify test conditions for incident simulation was developed. In order to preserve core decay power history and coolant mass inventory during a transient, a unique power-to-mass scaling method is proposed and utilized for RHRP and full-height, full-pressure (FHFP) systems. To validate the current scaling method, three counterpart tests done at the IIST facility are compared with the FHFP tests in small-break loss-of-coolant, station blackout, and loss-of-feedwater accidents performed at the Large-Scale Test Facility (LSTF) and the BETHSY test facility. Although differences appeared in design, scaling, and operation conditions among the IIST, LSTF, and BETHSY test facilities, the important physical phenomena shown in the facilities are almost the same. The physics involved in incident transient phenomena are well measured and modeled by showing the common thermal-hydraulic behavior of key parameters and the general consistency of chronological events. The results also confirm the adequacy of power-to-mass scaling methodology.

  8. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 8. Pipe fracture indirectly induced by an earthquake. Load Combination Program, Project I final report

    SciTech Connect (OSTI)

    Streit, R.D.

    1981-06-01

    This volume considers the probability that a double-ended guillotine break in the primary coolant loop of a pressurized water reactor occurs simultaneously with (and is indirectly caused by) a seismic event. The pipe break is a consequence of a seismically initiated failure in a system other than the primary piping itself. Events studied that can lead to an indirectly induced pipe break include structural and mechanical failures, missile impact, pressure transients, jet impingement, fire, and explosion. Structural failures of the supports for the reactor pressure vessel, reactor coolant pump, and steam generator have the highest probability of causing a double-ended pipe break. Furthermore, we found that structural failure of the containment dome and failure of the reactor coolant pump flywheel have the highest potential for a missile-caused pipe break. Since structural failure proved to be a major factor, we developed a model to estimate the probability of structural failure; this model is based on the engineering factors of safety and seismic hazard. preliminary results indicate that the probability of a double-ended pipe break indirectly caused by a seismic event during the plant life is on the order of 10/sup -9/.

  9. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 1011968 8171974 5202014 212000 6202001 5202034 Arkansas ...

  10. Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R. 11 NUCLEAR...

    Office of Scientific and Technical Information (OSTI)

    21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; SPENT FUEL STORAGE; PWR TYPE REACTORS; FUEL POOLS; STORAGE FACILITIES; ACCIDENTS; FAILURES; FISSION...

  11. TITLE AUTHORS SUBJECT SUBJECT RELATED DESCRIPTION PUBLISHER AVAILABILI...

    Office of Scientific and Technical Information (OSTI)

    REACTORS SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS BWR TYPE REACTORS SPENT FUEL STORAGE PWR TYPE REACTORS FUEL POOLS STORAGE FACILITIES ACCIDENTS FAILURES FISSION PRODUCTS...

  12. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    mechanics (14) pwr type reactors (14) water cooled reactors (14) reactor components ... heat transfer (10) reactors 210500* -- power reactors, breeding (10) Filter by Author ...

  13. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Preface: Fourth International Conference on Advances in Nuclear Science and Engineering ... Optimization of small long-life PWR based on thorium fuel Subkhi, Moh Nurul, E-mail: ...

  14. U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3

    Gasoline and Diesel Fuel Update (EIA)

    permanently discharged in previous years, the historical totals change. BWR Boiling-water reactor; PWR Pressurized-water reactor; HTGR High-temperature gas cooled reactor....

  15. DATE

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

    6-002 SECTION A. Project Title: Experimental Verification of Post-Accident iPWR Aerosol Behavior - Electric Power Research Institute, Inc. SECTION B. Project Description EPRI ...

  16. Kansas Nuclear Profile - Wolf Creek Generating Station

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

    April 2012" "Next Release Date: February 2013" "Wolf Creek Generating Station" ...0","9,556",94.0,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ...

  17. City of Shelbyville, Tennessee (Utility Company) | Open Energy...

    Open Energy Info (EERE)

    Twitter: @ShelbyvillePwr Facebook: https:www.facebook.compagesShelbyville-Power-Water-Sewerage396497130511649 Outage Hotline: 1-931-684-7171 References: EIA Form EIA-861...

  18. EC Publications

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

    ... and hard-rock unsaturated (waste package sizes up to 32-PWR) - Thermal analysis for mined, "open" concepts - Cost estimation for 5 disposal concepts - Summary and conclusions

  19. NEUTRONICS STUDIES OF URANIUM-BEARING ...

    Office of Scientific and Technical Information (OSTI)

    MICROENCAPSULATED FUEL FOR PRESSURIZED WATER REACTORS Nathan Michael George and Ivan ... based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). ...

  20. Metal Matrix Microencapsulated (M3) fuel neutronics performance...

    Office of Scientific and Technical Information (OSTI)

    ... in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors ... HOT PRESSING; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; STEAM; SWELLING; THERMAL ...