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Sample records for salem nuclear generating station unit-1

  1. 76 FR 19148 - PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1...

    Science.gov (United States)

    2011-04-06

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311, 50-354; NRC-2009-0390 and NRC-2009-0391] PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2..., DPR-70, and DPR-75 for an additional 20 years of operation for the Hope Creek Generating Station (HCGS...

  2. Salem Nuclear Generating Station, Unit 1. Annual operating report for 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Initial reactor criticality was achieved 12/11/76 and power generation began 12/25/76. Information is presented concerning operation, maintenance, procedure and specification changes, power generation, unit shutdowns and forced power reductions, testing, and personnel radiation exposures

  3. 77 FR 66484 - PSEG Nuclear LLC; Hope Creek Generating Station and Salem Generating Station, Units 1 and 2...

    Science.gov (United States)

    2012-11-05

    ... Power Plant Personnel,'' endorses the Nuclear Energy Institute (NEI) report, NEI 06-11, Revision 1... exclusion, set forth in 10 CFR 51.22(c)(25). Pursuant to 10 CFR 51.22(b), no environmental impact statement... Commission (NRC or the Commission) now or hereafter in effect. The facilities consist of one boiling-water...

  4. NRC Fact-Finding Task Force report on the ATWS event at Salem Nuclear Generating Station, Unit 1, on February 25, 1983

    International Nuclear Information System (INIS)

    1983-03-01

    An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1 on February 25, 1983. The charter of the Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the reactor trip breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions

  5. 75 FR 6223 - PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1...

    Science.gov (United States)

    2010-02-08

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311 and 50-354; NRC-2010-0043] PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 and 2...-70, and DPR-75, issued to PSEG Nuclear LLC (PSEG, the licensee), for operation of the Hope Creek...

  6. FIND: Douglas Point Nuclear Generating Station, Units 1 and 2

    International Nuclear Information System (INIS)

    Moore, M.M.

    1975-12-01

    This index is presented as a guide to microfiche items 1 through 136 in Docket 50448, which was assigned to Potomac Electric Power Company's Application for Licenses to construct and operate Douglas Point Nuclear Generating Station, Units 1 and 2. Information received from August, 1973 through July, 1975 is included

  7. Internal fire analysis screening methodology for the Salem Nuclear Generating Station

    International Nuclear Information System (INIS)

    Eide, S.; Bertucio, R.; Quilici, M.; Bearden, R.

    1989-01-01

    This paper reports on an internal fire analysis screening methodology that has been utilized for the Salem Nuclear Generating Station (SNGS) Probabilistic Risk Assessment (PRA). The methodology was first developed and applied in the Brunswick Steam Electric Plant (BSEP) PRA. The SNGS application includes several improvements and extensions to the original methodology. The SNGS approach differs significantly from traditional fire analysis methodologies by providing a much more detailed treatment of transient combustibles. This level of detail results in a model which is more usable for assisting in the management of fire risk at the plant

  8. 75 FR 36700 - Exelon Generation Company, LLC; Three Mile Island Nuclear Station, Unit 1; Environmental...

    Science.gov (United States)

    2010-06-28

    ...; Three Mile Island Nuclear Station, Unit 1; Environmental Assessment and Finding of No Significant Impact... Company, LLC (the licensee), for operation of Three Mile Island Nuclear Station, Unit 1 (TMI-1), located... Three Mile Island Nuclear Station, Units 1 and 2, NUREG-0552, dated December 1972, and Generic...

  9. 78 FR 77726 - Exelon Generation Company, LLC Three Mile Island Nuclear Station, Unit 1

    Science.gov (United States)

    2013-12-24

    ... Three Mile Island Nuclear Station, Unit 1 AGENCY: Nuclear Regulatory Commission. ACTION: Exemption... License No. DPR-50, which authorizes operation of the Three Mile Island Nuclear Station, Unit 1 (TMI-1... Facility Operating License No. DPR-50, which authorizes operation of the Three Mile Island Nuclear Station...

  10. 76 FR 24064 - Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Notice...

    Science.gov (United States)

    2011-04-29

    ... Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Notice of Issuance of Renewed... Company (licensee), the operator of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS... Plants: Supplement 43, Regarding Palo Verde Nuclear Generating Station,'' issued January 2011, discusses...

  11. Monitoring equipment environment during nuclear plant operation at Salem and Hope Creek generating stations

    International Nuclear Information System (INIS)

    Blum, A.; Smith, R.J.

    1991-01-01

    Monitoring of environmental parameters has become a significant issue for operating nuclear power plants. While the long-term benefits of plant life extension programs are being pursued with comprehensive environmental monitoring programs, the potential effect of local hot spots at various plant locations needs to be evaluated for its effect on equipment degradation and shortening of equipment qualified life. A significant benefit can be experienced from temperature monitoring when a margin exists between the design versus actual operating temperature. This margin can be translated into longer equipment qualified life and significant reduction in maintenance activities. At PSE and G, the immediate need for monitoring environmental parameters is being accomplished via the use of a Logic Beach Bitlogger. The Bitlogger is a portable data loggings system consisting of a system base, input modules and a communication software package. Thermocouples are installed on selected electrical equipment and cables are run from the thermocouples to the input module of the Bitlogger. Temperature readings are taken at selected intervals, stored in memory, and downloaded periodically to a PC software program, i.e., Lotus. The data is formatted into tabular or graphical documents. Because of their versatility, Bitloggers are being used differently at the authors Nuclear facility. At the Salem Station (2 Units-4 loop Westinghouse PWR), a battery powered, fully portable, calibrated Bitlogger is located in an accessible area inside Containment where it monitors the temperature of various electrical equipment within the Pressurizer Enclosure. It is planned that close monitoring of the local hot spot temperatures in this area will allow them to adjust and reconcile the environmental qualification of the equipment

  12. 75 FR 8149 - Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

    Science.gov (United States)

    2010-02-23

    ...] Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3... NPF-74, issued to the Arizona Public Service Company (APS, or the licensee), for operation of the Palo Verde Nuclear Generating Station (PVNGS, the facility), Units 1, 2, and 3, respectively, located in...

  13. 33 CFR 165.553 - Security Zone; Salem and Hope Creek Generation Stations, Delaware River, Salem County, New Jersey.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Security Zone; Salem and Hope... Limited Access Areas Fifth Coast Guard District § 165.553 Security Zone; Salem and Hope Creek Generation...: the waters of the Delaware River in the vicinity of the Salem and Hope Creek Generation Stations...

  14. 75 FR 9956 - PSEG Nuclear LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1...

    Science.gov (United States)

    2010-03-04

    ... letter number LR-N09- 0248 and the second enclosure is an environmental impact statement. Based on a... November 3, and November 20, 2009, submittals, with the exception of the environmental impact statement... M. S. Fertel, Nuclear Energy Institute, ADAMS Accession No. ML091410309). The licensee's request for...

  15. 75 FR 15745 - Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

    Science.gov (United States)

    2010-03-30

    ...] Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Exemption 1.0 Background The Arizona Public Service Company (APS, the licensee) is the holder of Facility... Generating Station (PVNGS), Units 1, 2, and 3, respectively. The licenses provide, among other things, that...

  16. 75 FR 13606 - Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

    Science.gov (United States)

    2010-03-22

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. STN 50-528, STN 50-529, and STN 50-530; NRC-2010-0114] Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Environmental...-74, issued to Arizona Public Service Company (APS, the licensee), for operation of the Palo Verde...

  17. Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

    International Nuclear Information System (INIS)

    Brown, R.; Cobanoglu, M.M.

    1995-01-01

    Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

  18. Technical evaluation of RETS-required reports for Rancho Seco Nuclear Generating Station, Unit 1 for 1983

    International Nuclear Information System (INIS)

    Magleby, E.H.; Young, T.E.

    1985-01-01

    A review of the reports required by Federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted during 1983 was performed. The periodic reports reviewed for the Rancho Seco Nuclear Generating Station, Unit 1 were the Semiannual Effluent Release Report, January 1, 1983 to June 30, 1983 and the Radiation Exposure, Environmental Protection, Effluent and Waste Disposal Report. The principal review guidelines were the plant's specific RETS which were based on NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

  19. Management of main generator condition during long term plant shut down at Higashidori Nuclear Power Station Unit 1

    International Nuclear Information System (INIS)

    Kato, Seiji

    2014-01-01

    Higashidori Nuclear Power Station Unit 1 shut down on February 6, 2011 to start 4th refuel outage. On March 11, 2011, we keep going refuel outage on this moment a large earthquake occurred and tsunami was generated following it which called 'Great East Japan Earthquake'. Refuel outage takes 3 ∼ 5 months normally but Higashidori NPS still keeping shut down over 3 years due to some issues. In this paper, we introduce about management of Main generator condition during long term plant shut down situation in addition to normal plant shut down situation to keep well. (author)

  20. Lessons learned from the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1

    International Nuclear Information System (INIS)

    Russell, M.J.; Shieh, L.C.; Tsai, N.C.; Cheng, T.M.

    1987-01-01

    A seismic reevaluation program was conducted for the San Onofre Nuclear Generating Station, Unit No. 1 (SONGS 1). SEP was created by the NRC to provide (1) an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The Systematic Evaluation Program (SEP) seismic review for SONGS 1 was exacerbated by the results of an evaluation of an existing capable fault near the site during the design review for Units 2 and 3, which resulted in a design ground acceleration of 0.67g. Southern California Edison Company (SCE), the licensee for SONGS 1, realized that a uniform application of existing seismic criteria and methods would not be feasible for the upgrading of SONGS 1 to such a high seismic requirement. Instead, SCE elected to supplement existing seismic criteria and analysis methods by developing criteria and methods closer to the state of the art in seismic evaluation techniques

  1. Critical evaluation of the nonradiological environmental technical specifications. Volume 4. San Onofre Nuclear Generating Station, Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.M.; Cunningham, P.A.; Gray, D.D.; Kumar, K.D.

    1976-08-10

    A comprehensive study of the data collected as part of the environmental Technical Specifications program for Unit 1 of the San Onofre Nuclear Generating Station (SONGS 1) was conducted for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The program included an analysis of the hydrothermal and ecological monitoring data collected during 1975. The hydrothermal analysis includes a discussion of models used in plume predictions prior to plant operation and an evaluation of the present hydrothermal monitoring program. The ecological evaluation was directed toward reviewing the strengths and weaknesses of the various sampling programs designed to monitor the planktonic, benthic, and nektonic communities inhabiting the inshore coastal area in the vicinity of San Onofre.

  2. Critical evaluation of the nonradiological environmental technical specifications. Volume 4. San Onofre Nuclear Generating Station, Unit 1

    International Nuclear Information System (INIS)

    Adams, S.M.; Cunningham, P.A.; Gray, D.D.; Kumar, K.D.

    1976-01-01

    A comprehensive study of the data collected as part of the environmental Technical Specifications program for Unit 1 of the San Onofre Nuclear Generating Station (SONGS 1) was conducted for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The program included an analysis of the hydrothermal and ecological monitoring data collected during 1975. The hydrothermal analysis includes a discussion of models used in plume predictions prior to plant operation and an evaluation of the present hydrothermal monitoring program. The ecological evaluation was directed toward reviewing the strengths and weaknesses of the various sampling programs designed to monitor the planktonic, benthic, and nektonic communities inhabiting the inshore coastal area in the vicinity of San Onofre

  3. Seismic structural fragility investigation for the San Onofre Nuclear Generating Station, Unit 1 (Project I); SONGS-1 AFWS Project

    International Nuclear Information System (INIS)

    Wesley, D.A.; Hashimoto, P.S.

    1982-04-01

    An evaluation of the seismic capacities of several of the San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) structures was conducted to determine input to the overall probabilistic methodology developed by Lawrence Livermore National Laboratory. Seismic structural fragilities to be used as input consist of median seismic capacities and their variabilities due to randomness and uncertainty. Potential failure modes were identified for each of the SONGS-1 structures included in this study by establishing the seismic load-paths and comparing expected load distributions to available capacities for the elements of each load-path. Particular attention was given to possible weak links and details. The more likely failure modes were screened for more detailed investigation

  4. San Onofre Nuclear Generating Station, Unit 1. Annual operating report for 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Gross electrical energy generated was 2,610,000 MWH with the generator on line 6,162.9 hrs. Information is presented concerning operations, power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, reportable occurrences, steam generator tube inspections, primary coolant chemistry, containment penetration leak tests, and radiological environmental monitoring

  5. Rancho Seco Nuclear Generating Station, Unit 1. Annual report for 1976

    International Nuclear Information System (INIS)

    1977-03-01

    Net electrical energy generated was 2,205,091 MWH with the generator on line 2,662 hrs. Information is presented concerning operations, changes, tests, maintenance, fuel performance, refueling, shutdowns and outages, containment local leak rate testing, and power generation

  6. Marble Hill Nuclear Generating Station, Units 1 and 2. License application, PSAR, general information

    International Nuclear Information System (INIS)

    1975-01-01

    An application is presented for two PWR reactors to be constructed in Salud Township, Jefferson County, Indiana, about six miles northeast of New Washington on the Ohio River. Each unit will have a rated core power level of 3411 MW(t) with a corresponding electrical output of 1130 MW(e). Mechanical draft cooling towers will be provided. The facility, which will replicate the Byron facility will be employed for the generation of electricity for transmission, sale for resale, and distribution

  7. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1. Design basis criteria used to evaluate the acceptability of the system include operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  8. Evaluation of Shoreham Nuclear Power Station, Unit 1 technical specifications

    International Nuclear Information System (INIS)

    Baxter, D.E.; Bruske, S.J.

    1985-08-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Shoreham Nuclear Power Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Shoreham T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Shoreham Nuclear Power Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

  9. Palo Verde Nuclear Generating Station, Units 1, 2 and 3 (Docket Nos. STN 50-528, STN 50-529 and STN 50-530): Final environmental statement

    International Nuclear Information System (INIS)

    1975-09-01

    The proposed action is the issuance of construction permits to the Arizona public Service Company for the construction of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3. Preparation of the 3800-acre site will involve the clearing of up to 2500 acres of land, 1500 of which will be prominently devoted to station facilities. An additional 1200- to 1300-acre evaporation pond will ultimately be developed during the lifetime of the station. About 2200 site acres, previously devoted to agriculture, will be excluded from this land use. Soil disturbance during construction of the station, transmission lines, and water conveyance pipeline will tend to promote erosion and increase siltation in local ephemeral water courses. Stringent measures will be taken to minimize these effects. The total radiation dose to construction workers is estimated to be 15 man-rem. This dose is a small fraction of the approximately 470 man-rem which will be received by the construction force over the same period from natural background radiation. Station, transmission line, and water pipeline construction will kill, remove, displace, or otherwise disturb involved flora and fauna, and will eliminate varying amounts of wildlife breeding, nesting, and forage habitat. These will not be important permanent impacts to the population stability and structure of the involved local ecosystems of the Sonoran desert; however, measures will be taken to minimize such effects as do result form the proposed action

  10. 76 FR 40754 - Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units...

    Science.gov (United States)

    2011-07-11

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0100; Docket Nos. 50-413 and 50-414; Docket Nos. 50-369 and 50-370; Docket Nos. 50-269, 50-270, And 50-287] Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; Notice...

  11. 76 FR 24538 - Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station...

    Science.gov (United States)

    2011-05-02

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-413 and 50-414; NRC-2011-0100; Docket Nos. 50-369 and 50-370; Docket Nos. 50-269, 50-270, and 50-287] Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3...

  12. 75 FR 9623 - Arizona Public Service Company, et al.; Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

    Science.gov (United States)

    2010-03-03

    ... pressurized-water reactors located in Maricopa County, Arizona. 2.0 Request/Action Title 10 of the Code of... pressure boundary during normal operating and hydrostatic or leak rate testing conditions. Specifically, 10... NRC's March 16, 2001, SE, the staff noted, ``[t]he CE NSSS [nuclear steam supply system] methodology...

  13. Palo Verde Nuclear Generating Station Units 1, 2, and 3: Draft environmental statement (Docket Nos. STN 50-528, 529, and 530)

    International Nuclear Information System (INIS)

    1975-04-01

    The proposed action is the issuance of construction permits to the Arizona Public Service Corporation for the construction of the Palo Verde Nuclear Generating Station, Units 1, 2 and 3. Preparation of the 3880-acre site will involve the clearing of up to 2500 acres of land, 1500 of which will be permanently devoted to station facilities. An additional 1200- to 1300-acre evaporation pond will ultimately be developed during the lifetime of the station. About 2200 site acres, previously devoted to agriculture, will be excluded from this land use (Sec. 4. 1). Soil disturbance during construction of the station, transmission lines, and water conveyance pipeline will tend to promote erosion and increase siltation in local ephemeral water courses. Stringent measures will be taken to minimize these effects (Sec. 4.5). Station, transmission line, and water pipeline construction will kill, remove displace, or otherwise disturb involved flora and fauna, and will eliminate varying amounts of wildlife breeding, nesting, and forage habitat. These will not be important impacts to the population stability and structure of the involved local ecosystems of the Sonoran desert; however, measures will be taken to minimize such effects (Sec. 4.3 and 4.5). Approximately 60 acres of agricultural land will be temporarily affected by construction in transmission corridors. The great majority can be returned to that use upon completion of construction, thus the impact is considered minor. Similarly, most grazing lands affected along these corridors, as well as along the water pipeline corridor, can eventually be returned to that use. New archaeological resources could be discovered along the path of final transmission corridor alignments. The applicant will take measures to locate and protect such resources if they exist. 75 refs., 24 figs., 65 tabs

  14. Integrated plant safety assessment: Systematic Evaluation Program, San Onofre Nuclear Generating Station, Unit 1 (Docket No. 50-206): Final report

    International Nuclear Information System (INIS)

    1986-12-01

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues; (2) a basis for deciding on how these differences should be resolved in an integrated plant review; and (3) a documented evaluation of plant safety. This report documents the review of San Onofre Nuclear Generating Station, Unit 1, operated by Southern California Edison Company. The San Onofre plant is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. This report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the draft report issued in April 1985

  15. Marble Hill Nuclear Generating Station, Units 1 and 2: Final environmental statement (Docket Nos. STN 50-546 and STN 50-547)

    International Nuclear Information System (INIS)

    1976-09-01

    The proposed action is the issuance of construction permits to the Public Service Company of Indiana, Inc., Northern Indiana Public Service Company, Inc., East Kentucky Power Cooperative, Inc., and Wabash Valley Power Association for the construction of the Marble Hill Nuclear Generating Station, Units 1 and 2 (MH 1and2). The 987-acre site is predominately forest and cropland. Construction-related activities on the site would disturb about 250 acres. The portion of this land not be used for the plant facilities, parking lots, roads, etc., will be restored by seeding and landscaping. The temporary removal of vegetation will tend to promote erosion. Increased siltation and turbidity can be expected in local streams during construction, but measures will be taken to minimize these effects. A maximum of 69 cfs of cooling water will be withdrawn from the Ohio River of which cfs will be returned to the river via pipeline with the dissolved solids concentration increased by a factor of about 6. About cfs will be evaporated to the atmosphere by the cooling towers. The volume of discharge (9 cfs) is very small compared with the river flow (annual mean is about 110,000 cfs) and the thermal effect on the river ecosystem is not expected to be significant. Chemical discharges from the plant will be diluted to concentrations below those which might adversely affect aquatic biota. The risk associated with accidental radiation exposure will be very low. 43 figs., 115 tabs

  16. Station black out of Fukushima Daiichi Nuclear Power Station Unit 1 was not caused by tsunamis

    International Nuclear Information System (INIS)

    Ito, Yoshinori

    2013-01-01

    Station black out (SBO) of Fukushima Daiichi Nuclear Power Station Unit 1 would be concluded to be caused before 15:37 on March 11, 2011 because losses of emergency ac power A system was in 15:36 and ac losses of B system in 15:37 according to the data published by Tokyo Electric Power Co. (TEPCO) in May 10, 2013. Tsunami attacked the site of Fukushima Daiichi Nuclear Power Station passed through the position of wave amplitude meter installed at 1.5 km off the coast after 15:35 and it was also recognized tsunami arrived at the coast of Unit 4 sea side area around in 15:37 judging from a series of photographs taken from the south side of the site and general knowledge of wave propagation. From a series of photographs and witness testimony, tsunami didn't attack Fukushima Daiichi Nuclear Power Station uniformly and tsunami's arrival time at the site of Unit 1 would be far later than arrival time at the coast of Unit 4 sea side area, which suggested it would be around in 15:39. TEPCO insisted tsunami passed through 1.5 km off the coast around in 15:33 and clock of wave amplitude meter was incorrect, which might be wrong. Thus SBO of Fukushima Daiichi Nuclear Power Station Unit 1 occurred before tsunami's arrival at the site of Unit 1 and was not caused by tsunami. (T. Tanaka)

  17. Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530): Final environmental statement

    International Nuclear Information System (INIS)

    1982-02-01

    The proposed action is the issuance of operating licenses to the Arizona Public Service Company (APS, applicant) for the startup and operation of PVNGS, Units 1, 2, and 3, located in Maricopa County, about 24 km (15 mi) west of Buckeye, Arizona. The information in this statement represents the second assessment of the environmental impact associated with PVNGS Units 1, 2, and 3 pursuant to the guidelines of the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations (10 CFR) Part 51 of the Commissions's Regulations. After receiving an application in July 1974 to construct this station, the staff carried out a review of impacts that would occur during its construction and operation. That evaluation was issued as a Final Environmental Statement/emdash/Construction Phase (FES-CP). After this environmental review, a safety review, an evaluation by the Advisory Committee on Reactor Safeguards, and public hearings in Phoenix, Arizona, the US Nuclear Regulatory Commission issued Construction Permits Nos. CPPR-141, CPPR-142, and CPPR-143 for the construction of PVNGS Units 1, 2, and 3. As of September 1981, the construction of Unit 1 was about 92 percent complete, Unit 2 was 68 percent complete, and Unit 3 was 26 percent complete. 11 figs., 21 tabs

  18. Final environmental statement related to the proposed construction of Douglas Point Nuclear Generating Station, Units 1 and 2: (Docket Nos. 50-448 and 50-449)

    International Nuclear Information System (INIS)

    1976-03-01

    The proposed action is the issuance of construction permits to the Potomac Electric Power Company for the construction of the Douglas Point Nuclear Generating Station, Units 1 and 2, located in Charles County, Maryland. The exhaust steam will be cooled via a closed-cycle mode incorporating natural-draft wet cooling towers. The water used in the cooling system will be obtained from the Potomac River. Construction-related activities on the site will convert about 290 acres of the 1390 acres of forested land at the Douglas Point site to industrial use. In addition to acreage at the site, approximately 4.5 miles of transmission corridor will require about 211 acres of land for rights-of-way. This corridor will connect with 27 miles of existing rights-of-way over which a line connecting Possum Point to Burches Hill has already been approved. The installation of new transmission line, uniquely identified with Douglas Point, along the existing right-of-way will involve approximately 464 additional acres. As described in the application, the maximum river water intake will be about 97,200 gpm. Of this, a maximum of about 28,000 gpm will be lost in drift or evaporation from the cooling towers. About 700 gpm maximum of fresh well water will be consumed. It is conservatively assumed that all aquatic organisms entrained in the service water system will be killed due to thermal and mechanical shock. It is further estimated that at 97,200 gpm maximum total river water intake, the maximum impact on the striped bass fishery will be a reduction of <5%. The risk associated with accidental radiation exposure is very low. 32 figs., 59 tabs

  19. Audit of Wolf Creek Generating Station, Unit 1 technical specifications. Final technical evaluation report

    International Nuclear Information System (INIS)

    Stromberg, H.M.

    1985-07-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Wolf Creek Generating Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, the requirements of the Safety Evaluation Report (SER) as supplemented, and the Comments and Responses to the Wolf Creek Technical Specification Draft Inspection Report. A comparative audit of the FSAR as amended, the SER as supplemented, and the Draft Inspection Report was performed with the Wolf Creek T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Wolf Creek Generating Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR, SER, and Draft Inspection Report

  20. Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530)

    International Nuclear Information System (INIS)

    1984-10-01

    Supplement No. 6 to the Safety Evaluation Report for the application filed by Arizona Public Service Company, et al., for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528/529/530), located in Maricopa County, Arizona, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of (1) additional information submitted by the applicant since Supplement No. 5 was issued and (2) matters that the staff had under review when Supplement No. 5 was issued

  1. Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530). Supplement No. 7

    International Nuclear Information System (INIS)

    1984-12-01

    Supplement No. 7 to the Safety Evaluation Report for the application filed by Arizona Public Service Company et al. for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528/529/530), located in Maricopa County, Arizona, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of: (1) additional information submitted by the applicant since Supplement No. 6 was issued; and (2) matters that the staff had under review when Supplement No. 6 was issued

  2. 75 FR 8757 - Nebraska Public Power District, Cooper Nuclear Station, Unit 1; Notice of Availability of the...

    Science.gov (United States)

    2010-02-25

    ..., Cooper Nuclear Station, Unit 1; Notice of Availability of the Draft Supplement 41 to the Generic... Renewal of Cooper Nuclear Station, Unit 1 Notice is hereby given that the U.S. Nuclear Regulatory... operating license DPR-46 for an additional 20 years of operation for Cooper Nuclear Station, Unit 1 (CNS-1...

  3. Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

    International Nuclear Information System (INIS)

    Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

    2015-01-01

    As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

  4. Pilgrim Nuclear Power Station, Unit 1. Annual operating report, 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electric power generated in 1975 was 1,074,401 MW(e) with the generator on line 4,680.7 hrs. Information is presented concerning operations, maintenance, radioactive effluents and waste shipments, health physics, shutdowns, and personnel exposures

  5. Pilgrim Nuclear Power Station, Unit 1. Annual operating report, 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 2,587,248 MWH(e) with the reactor on line 6,242.4 hr. Information is presented concerning operations, power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, and reportable occurrences

  6. 76 FR 45301 - PSEG Nuclear LLC, Hope Creek Generating Station; Notice of Issuance of Renewed Facility Operating...

    Science.gov (United States)

    2011-07-28

    ... NUCLEAR REGULATORY COMMISSION Docket No. 50-354 [NRC-2009-0391] PSEG Nuclear LLC, Hope Creek... operator of the Hope Creek Generating Station (HCGS). Renewed Facility Operating License No. NPF- 57... Renewal of Nuclear Power Plants, Supplement 45, Regarding Hope Creek Generating Station and Salem Nuclear...

  7. Safety Evaluation Report related to the restart of Rancho Seco Nuclear Generating Station, Unit 1, following the event of December 26, 1985 (Docket No. 50-312)

    International Nuclear Information System (INIS)

    1987-10-01

    On December 26, 1985, the Rancho Seco Nuclear Generating Station, owned and operated by the Sacramento Municipal Utility District (SMUD), experienced a loss of dc power within the integrated control system (ICS) while the plant was at 76% power. The ensuing reactor trip was followed by a rapid overcooling transient and automatic initiation of the safety features actuation system (SFAS). The overcooling transient continued until ICS dc power was restored 26 minutes after its loss. Two letters from the NRC Region V Administrator (dated December 26, 1985) confirmed that the Rancho Seco plant would not be returned to power operation until SMUD (the licensee) had provided the NRC with an assessment of the root cause of the transient and a justification as to why the Rancho Seco facility is ready to resume power operation. In response, the licensee submitted the ''Rancho Seco Action Plan for Performance Improvement'' on July 3, 1986; revisions to that action plan were submitted on December 15, 1986 and February 28, 1987. The NRC staff has reviewed the action plan and numerous other supporting documents submitted by the licensee. The staff's evaluation of the information supporting restart of Rancho Seco is presented in this safety evaluation report

  8. Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Dockets Nos. STN 50-528, STN 50-529, and STN 50-530)

    International Nuclear Information System (INIS)

    1987-03-01

    Supplement No. 11 to the Safety Evaluation Report for the application filed by Arizona Public Service Company et al. for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528/529/530), located in Maricopa County, Arizone, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of (1) additional information submitted by the applicant since Supplement No. 10 was issued and (2) other matters requiring staff review since Supplenent No. 10 was issued, specifically those issues that required resolution before Unit 3 low-power licensing

  9. Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530). Supplement No. 9

    International Nuclear Information System (INIS)

    1985-12-01

    Supplement No. 9 to the Safety Evaluation Report for the application filed by Arizona Public Service Company et al. for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2 and 3 (Docket Nos. STN 50-528/529/530), located in Maricopa County, Arizona, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of: (1) additional information submitted by the applicant since Supplement No. 8 was issued; and (2) matters that the staff had under review when Supplement No. 8 was issued, specifically those issues which required resolution prior to Unit 2 fuel loading and testing up to 5% of full power

  10. Radiological Effluent Technical Specifications (RETS) implementation: Zion Generating Station Units 1 and 2

    International Nuclear Information System (INIS)

    Serrano, W.; Akers, D.W.; Duce, S.W.; Mandler, J.W.; Simpson, F.B.; Young, T.E.

    1985-06-01

    A review of the Radiological Effluent Technical Specifications (RETS) of the Zion Generating Station Units 1 and 2 was performed. The principal review guidelines used were NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,'' and Draft 7 of NUREG-0472, Revision 3, ''Radiological Effluent Technical Specifications for Pressurized Water Reactors.'' Draft submittals were discussed with the Licensee by both EG and G and the NRC staff until all items requiring changes to the Technical Specifications were resolved. The Licensee then submitted final proposed RETS to the NRC which were evaluated and found to be in compliance with the NRC review guidelines. The proposed Offsite Dose Calculation Manual was reviewed and generally found to be consistent with the NRC review guidelines. 35 refs., 2 figs., 1 tab

  11. Palo Verde Nuclear Generating Station, Units 1, 2 and 3 (Docket Nos. STN 50-528, STN 50-529 and STN 50-530): Draft supplement to the Final environmental satement

    International Nuclear Information System (INIS)

    1975-11-01

    The proposed action is the issuance of construction permits to the Arizona Public Service Company for the construction of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3. Preparation of the 3800-acre site will involve the clearing of up to 2500 acres of land, 1500 of which will be permanently devoted to station facilities. An additional 1200- to 1300-acre evaporation pond will ultimately be developed during the lifetime of the station. About 2200 site acres, previously devoted to agriculture, will be excluded from this land use. Soil disturbance during construction of the station, transmission lines, and water conveyance pipeline will tend to promote erosion and increase siltation in local ephemeral water courses. Stringent measures will be taken to minimize these effects. Station, transmission line, and water pipeline construction will kill, remove, displace, or otherwise disturb involved flora and fauna, and will eliminate varying amounts of wildlife breeding, nesting, and forage habitat. These will not be important permanent impacts to the population stability and structure of the involved local ecosystems of the Sonoran desert; however, measures will be taken to minimize such effects as do result from the proposed action. The pumping of groundwater will cause a local drawdown of about 1 ft/yr, less than that currently occurring; hence, the impact is considered acceptable. 1 fig., 20 tabs

  12. 76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

    Science.gov (United States)

    2011-11-21

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-295 and 50-304; NRC-2011-0244] ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor...

  13. 75 FR 43572 - Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2; Environmental Assessment and...

    Science.gov (United States)

    2010-07-26

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-369 and 50-370; NRC-2010-0259] Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2; Environmental Assessment and Finding of No Significant... Energy Carolinas, LLC (the licensee), for operation of the McGuire Nuclear Station, Units 1 and 2...

  14. 75 FR 43571 - Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; Environmental Assessment And...

    Science.gov (United States)

    2010-07-26

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-413 and 50-414; NRC-2010-0260] Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; Environmental Assessment And Finding of No Significant... Energy Carolinas, LLC (the licensee), for operation of the Catawba Nuclear Station, Units 1 and 2...

  15. Final supplement to the final environmental statement related to construction of Palo Verde Nuclear Generating Station Units 1, 2 and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530)

    International Nuclear Information System (INIS)

    1976-02-01

    The proposed action is the issuance of construction permits to the Arizona Public Service Company for the construction of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3. Preparation of the 3800-acre site will involve the clearing of up to 2500 acres of land, 1500 of which will ultimately be developed during the lifetime of the station. About 2200 site acres, previously devoted to agriculture, will be excluded from this land use. Soil disturbance during construction of the station, transmission lines, and water conveyance pipeline will tend to promote erosion and increase siltation local ephemeral water courses. Stringent measures will be taken to minimize these effects (Sec. 4.5). Station, transmission line, and water pipeline construction will kill, remove, displace, or otherwise disturb involved flora and fauna, and will eliminate varying amounts of wildlife breeding, nesting, and forage habitat. These will not be important permanent impacts to the population stability and structure of the involved local ecosystems of the Sonoran desert; however, measures will be taken to minimize such effects as do results from the proposed action. 26 refs., 1 fig., 20 tabs

  16. Second periodic safety review of Angra Nuclear Power Station, unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M., E-mail: ottoncf@tecnatom.com.br, E-mail: emfreire46@gmail.com, E-mail: robcrepaldi@hotmail.com [Tecnatom do Brasil Engenharia e Servicos Ltda, Rio de Janeiro, RJ (Brazil); Campello, Sergio A., E-mail: sacampe@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

  17. Second periodic safety review of Angra Nuclear Power Station, unit 1

    International Nuclear Information System (INIS)

    Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M.; Campello, Sergio A.

    2015-01-01

    This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

  18. Final environmental statement for Shoreham Nuclear Power Station, Unit 1: (Docket No. 50-322)

    International Nuclear Information System (INIS)

    1977-10-01

    The proposed action is the issuance of an Operating License to the Long Island Lighting Company (LILCO) for the startup and operation of the Shoreham Nuclear Power Station, Unit 1 (the plant) located on the north shore of Long Island, the State of New York, County of Suffolk, in the town of Brookhaven. The Shoreham station will employ a boiling-water reactor (BWR), which will operate at a thermal output of 2436 MW leading to a gross output of 846 MWe and a net output of about 820 MWe. The unit will be cooled by once-through flow of water from the Long Island Sound. One nuclear unit with a net capacity of 820 MWe will be added to the generating resources of the Long Island Lighting Company. This will have a favorable effect on reserve margins and provide a cost savings of approximately $62.1 million (1980 dollars) in production costs in 1980 if the unit comes on line as scheduled; additional cost savings will be realized in subsequent years. Approximately 100 acres (40 hectares) of the 500-acre (202-hectare) site of rural (mostly wooded) land owned by the applicant have been cleared. Most of this will be unavailable for other uses during at least the 40-year life of the plant. No offsite acreage has been or will be cleared. Land in the vicinity of the site has undergone some residential development that is typical for all of this area of Long Island. The operation of Shoreham Unit 1 will have insignificant impacts on this and other types of land uses in the vicinity of the site. 33 figs., 56 tabs

  19. Safety evaluation report related to the full-term operating license for San Onofre Nuclear Generating Station, Unit 1 (Docket No. 50-206)

    International Nuclear Information System (INIS)

    1991-07-01

    The safety evaluation report for the full-term operating license application filed by the Southern California Edison Company and the San Diego Gas and Electric Company has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in San Diego County, California. The staff has evaluated the issues related to the conversion of the provisional operating license to a full-term operating license and concluded that the facility can continue to be operated without endangering the health and safety of the public following the license conversion. 43 refs., 3 figs., 3 tabs

  20. North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

    International Nuclear Information System (INIS)

    Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

    1993-01-01

    The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

  1. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1993-01-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  2. Results of the 4th regular inspection in Unit 1 of the Mihama Nuclear Power Station

    International Nuclear Information System (INIS)

    1981-01-01

    The 4th regular inspection of Unit 1 in the Mihama Nuclear Power Station was made from July, 1975, to December, 1980, on its reactor and associated facilities. The respective stages of inspection during the years are described. The inspection by external appearance examination, disassembling leakage inspection and performance tests indicated crackings in piping for fuel-replacement water tank, the container penetration of recirculation pipe for residual-heat removal, and main steam-relief valve, and leakage in one fuel assembly. Radiation exposure of the personnel during the inspection was less than the permissible dose. Radiation exposure data for the personnel are given in tables. The improvements and repairs done accordingly were as follows: reapir of the piping for a fuel-replacement tank and recirculation piping for residual-heat removal, replacement of the main steam-relief valve, plugging of heating tubes for the steam-generator, replacement of pins and covers for control-rod guide pipes, improvement of safety protection system and installation of rare gas monitor. (J.P.N.)

  3. Probabilistic fire risk assessment for Koeberg Nuclear Power Station Unit 1

    International Nuclear Information System (INIS)

    Grobbelaar, J.F.; Foster, N.A.S.; Luesse, L.J.

    1995-01-01

    A probabilistic fire risk assessment was done for Koeberg Nuclear Power Station Unit 1. Areas where fires are likely to start were identified. Equipment important to safety, as well as their power and/or control cable routes were identified in each fire confinement sector. Fire confinement sectors where internal initiating events could be caused by fire were identified. Detection failure and suppression failure fault trees and event trees were constructed. The core damage frequency associated with each fire confinement sector was calculated, and important fire confinement sectors were identified. (author)

  4. Generic implications of ATWS events at the Salem Nuclear Power Plant. Licensee and staff actions

    International Nuclear Information System (INIS)

    1983-08-01

    This report, Volume 2 of two volumes of NUREG-1000, describes the intermediate term actions to be taken by licensees and applicants of the US Nuclear Regulatory Commission (NRC), on the one hand, and by NRC staff, on the other, to address the generic issues raised by two anticipated transients without scram (ATWS) at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983. These actions came about as a result of the findings of NUREG-1000, Volume 1, and of reviews by the NRC Committee to Review Generic Requirements, the NRC Program Offices, and the Commission. The actions to be taken by licensees and applicants have been detailed in a letter pursuant to 10 CFR 50.54(f)

  5. 76 FR 79228 - Combined Licenses at William States Lee III Nuclear Station Site, Units 1 and 2; Duke Energy...

    Science.gov (United States)

    2011-12-21

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-018 and 52-019; NRC-2008-0170] Combined Licenses at William States Lee III Nuclear Station Site, Units 1 and 2; Duke Energy Carolinas, LLC AGENCY: Nuclear.... SUMMARY: Notice is hereby given that the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Army Corps...

  6. Summary of plant life management evaluation for Onagawa Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    Nodate, Kazumi

    2014-01-01

    The Onagawa Nuclear Power Station Unit-1 (Onagawa NPS-1) began commercial operation on June 1, 1984, and has reached 30-year from starting of operation on June of 2014. To that end, we implemented the Plant Life Management (PLM) evaluation for Onagawa NPS-1 as our first experience. We decided on a Long-term Maintenance Management Policy from result of the evaluation, and then applied the Safety-Regulations change approval application on November 6, 2013 and its correcting application on April 16, 2014. Our application was approved on May 21, 2014 through investigation by the Nuclear Regulatory Agency. Also at implementation of the PLM evaluation, we considered effects of the Great East Japan Earthquake that occurred on March 11, 2011 against ageing phenomena. In this paper, we introduce summary of PLM evaluation for Onagawa NPS-1 and the evaluation that considered effects of the Great East Japan Earthquake. (author)

  7. Draft environmental impact statement. River Bend Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Federal financing of an undivided ownership interest of River Bend Nuclear Power Station Unit 1 on a 3293-acre site near St. Francisville, Louisiana is proposed in a supplement to the final environmental impact statement of September 1974. The facility would consist of a boiling-water reactor that would produce a maximum of 2894 megawatts (MW) of electrical power. A design level of 3015 MW of electric power could be realized at some time in the future. Exhaust steam would be cooled by mechanical cooling towers using makeup water obtained from and discharged to the Mississippi River. Power generated by the unit would be transmitted via three lines totaling 140 circuit miles traversing portions of the parishes of West Feliciana, East Feliciana, East Baton Rouge, West Baton Rouge, Pointe Coupee, and Iberville. The unit would help the applicant meet the power needs of rural electric consumers in the region, and the applicant would contribute significanlty to area tax base and employment rolls during the life of the unit. Construction related activities would disturb 700 forested acres on the site and 1156 acres along the transmission routes. Of the 60 cubic feet per second (cfs) taken from the river, 48 cfs would evaporate during the cooling process and 12 cfs would return to the river with dissolved solids concentrations increased by 500%. The terrace aquifer would be dewatered for 16 months in order to lower the water table at the building site, and Grants Bayou would be transformed from a lentic to a lotic habitat during this period. Fogging and icing due to evaporation and drift from the cooling towers would increase slightly. During the construction period, farming, hunting, and fishing on the site would be suspended, and the social infractructure would be stressed due to the influx of a maximum of 2200 workers

  8. 78 FR 77508 - Duke Energy Carolinas, LLC; William States Lee III Nuclear Station, Units 1 and 2; Combined...

    Science.gov (United States)

    2013-12-23

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-018 and 52-019; NRC-2008-0170] Duke Energy Carolinas, LLC; William States Lee III Nuclear Station, Units 1 and 2; Combined Licenses Application Review AGENCY: Nuclear Regulatory Commission. ACTION: Final environmental impact statement; availability...

  9. 78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

    Science.gov (United States)

    2013-07-03

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 72-1004, 72-40, 50-269, 50-270, and 50-287; NRC-2013-0135] Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel Storage Installation; Environmental Assessment and Finding of No Significant Impact AGENCY: Nuclear...

  10. 78 FR 45575 - Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

    Science.gov (United States)

    2013-07-29

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos.: 72-1004, 72-40, 50-269, 50-270, 50-287; and NRC-2013- 0135] Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory Commission. ACTION: Exemption; issuance. SUMMARY: The NRC...

  11. Results of the 5th regular inspection of Unit 1 in the Hamaoka Nuclear Power Station

    International Nuclear Information System (INIS)

    1983-01-01

    The 5th regular inspection of Unit 1 in the Hamaoka Nuclear Power Station was carried out from March 27 to July 27, 1982. Inspection was made on the reactor proper, reactor cooling system, instrumentation/control system, radiation control facility, etc. By the examinations of external appearance, leakage, performance, etc., no abnormality was observed. In the regular inspection, personnel exposure dose was all below the permissible level. The works done during the inspection were the following: the replacement of control rod drives, the replacement of core support-plate plugs, the repair of steam piping, steam extraction pipes and feed water heaters, the repair of a waste-liquid concentrator, the installation of barriers and leak detectors, the installation of drain sump monitors in a containment vessel, the replacement of concentrated liquid waste pumps, the employment of type B fuel. (Mori, K.)

  12. Data book of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2002-03-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The result of the present examination has already been reported to NISA and has also been published as the JAERI-Tech report No.2001-94. This report is a data book containing the detailed data obtained by the present examination. (author)

  13. Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2001-12-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

  14. Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement 2

    International Nuclear Information System (INIS)

    1984-10-01

    In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the applicant) for licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania. This supplement addresses further issues that require resolution and closes them out

  15. Oconee Nuclear Station, Units 1, 2, and 3. Semiannual operating report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented concerning operations, performance characteristics, changes, tests, inspections, containment leak tests, maintenance, primary coolant chemistry, station staff changes, reservoir investigations, plume mapping, and operational environmental radioactivity monitoring data for oconee Units 1, 2, and 3. The non-radiological environmental surveillance program is also described. (FS)

  16. Dresden Nuclear Power Station, Units 1, 2, and 3. Annual operating report: January thru December 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical energy generated by Unit 1 was 953,015.5 MWH with the generator on line 7,399.37 hrs. Unit 2 generated 4,371,553.689 MWH with the generator on line 6,664.58 hrs while Unit 3 generated 4,034,251 MWH with the generator on line 7,234.86 hrs. Information is presented concerning operations, maintenance, and shutdowns

  17. 78 FR 29158 - In the Matter of Zion Solutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Order Approving...

    Science.gov (United States)

    2013-05-17

    ... and DPR-48] In the Matter of Zion Solutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Order... formed for the purpose of acquiring ES, Inc. and is held by certain investment fund entities organized by... Environmental Management Programs, in writing, of such receipt no later than one (1) business day prior to the...

  18. 76 FR 30204 - Exelon Nuclear, Dresden Nuclear Power Station, Unit 1; Exemption From Certain Security Requirements

    Science.gov (United States)

    2011-05-24

    ... contained in the Responsibility Matrix of the safeguards contingency plan.'' Part 73 of Title 10 of the Code... organization, which will have as its objective to provide high assurance that activities involving special... structures) for DNPS Unit 1 is in a form that does not pose a risk of removal (i.e., an intact reactor...

  19. Safety evaluation report related to the operation of Palo Verde nuclear generating station, Units 1, 2, and 3. Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Arizona Public Service Company

    International Nuclear Information System (INIS)

    1982-05-01

    On November 13, 1981, the Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) relating to the application for licenses to operate the Palo Verde Nuclear Generating Station, Unit Nos. 1, 2 and 3 (PVNGS 1-3); Supplement No. 1 to the SER was issued on February 4, 1982. In the SER and Supplement No. 1, the staff identified certain issues where either further information was required of the applicant or additional staff effort was necessary to complete the review of the application. The purpose of this supplement is to update the SER by providing (1) the evaluation of additional information submitted by the applicant since Supplement No. 1 to the SER was issued, and (2) the evaluation of the matters that the staff had under review and Supplement No. 1 was issued

  20. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

    International Nuclear Information System (INIS)

    Huerta B, A.; Lopez M, R.

    1995-01-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and.analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  1. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

    International Nuclear Information System (INIS)

    Huerta B, A.; Lopez M, R.

    1995-01-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  2. Report on Darlington nuclear generating station

    International Nuclear Information System (INIS)

    1985-12-01

    The Select Committee on Energy was appointed on July 10, 1985 by the Legislative Assembly of the Province of Ontario in order to inquire into and report on Ontario Hydro affairs within ten months. Two sessions were planned the first of which was a review of the Darlington Nuclear Generating Station. Darlington is a large, 4 unit nuclear-powered electricity generating station currently under construction on the shore of Lake Ontario in the town of Newcastle. At the time the Committee met, construction had been underway for over four years. The first two units are scheduled to become operational in 1988 and 1989 with the second two scheduled to become operational in 1991 and 1992. The total estimated cost of the station is $10.895 billion of which $3.66 billion has been spent and $3.385 billion has been committed. Though the nuclear industry has been a major area of investment in Ontario over the past decade, the demand for electrical power from nuclear stations has been significantly decreased. This report focusses on the need for Darlington and public policy issues involved in planning and completing it. The Committee proposed the following recommendations: 1) The relationship between the Government of Ontario and Ontario Hydro and their individual responsibilities should be clarified. 2) An independent review of the Ontario Hydro demand/supply options should be carried out. 3) No further significant contracts for Darlington units 3 and 4 should be let for materials not required for construction during the next 6 months while the Committee studies demand and supply options

  3. Technical evaluation of RETS-required reports for Browns Ferry Nuclear Power Station, Units 1, 2, and 3, for 1983

    International Nuclear Information System (INIS)

    Young, T.E.; Magleby, E.H.

    1985-01-01

    A review was performed of reports required by federal regulations and the plant-specific radiological effluent technical specifications (RETS) for operations conducted at Tennessee Valley Authority's Browns Ferry Nuclear Station, Units 1, 2, and 3, during 1983. The two periodic reports reviewed were (a) the Effluents and Waste Disposal Semiannual Report, First Half 1983 and (b) the Effluents and Waste Disposal Semiannual Report, Second Half 1983. The principal review guidelines were the plant's specific RETs and NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

  4. Structural design of the turbine building of Angra Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Varella, L.N.; Reis, F.J.C.; Jurkiewicz, W.J.

    1978-01-01

    The Turbine Building of the Angra Nuclear Power Plant, Unit 1, and particularly its structure and structural design are described. The Turbine Building, as far as its structure is concerned, deviates from the standard structure of any turbine building due to the fact that huge ducts are provided in the foundation mat as to accomodate the circulating water system. This aspect and the fact that the building is founded upon a very deep strata of compacted and controlled fill, makes out of the building structure 'a concrete ship floating in the sea of sand', and by the same reason presents by itself an interesting structure, worth to be known to all engineers involved in design of power plants. This pape, suplemented by a few slides shown during presentation of the paper at the conference, covers the subject mainly from the designers' point of view. (Author)

  5. Technical evaluation report on the proposed design modifications and technical-specification changes on grid voltage degradation for the San Onofre Nuclear Genetating Station, Unit 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the proposed design modifications and Technical Specification changes for protection of Class 1E equipment from grid voltage degradation for the San Onofre Nuclear Generating Station, Unit 1. The review criteria are based on several IEEE standards and the Code of Federal Regulations. The evaluation finds that the proposed design modifications and Technical Specification changes will ensure that the Class 1E equipment will be protected from sustained voltage degradation

  6. Final environmental statement related to the operation of Limerick Generating Station, Units 1 and 2, Docket Nos. 50-352 and 50-353, Philadelphia Electric Company

    International Nuclear Information System (INIS)

    1989-08-01

    In April 1984 the staff of the Nuclear Regulatory Commission issued its Final Environmental Statement (NUREG-0974) related to the operation of Limerick Generating Station, Units 1 and 2, (Docket Nos. 50-352 and 50-353), located on the Schuylkill River, near Pottstown, in Limerick Township, Montgomery and Chester Counties, Pennsylvania. The NRC has prepared this supplement to NUREG-0974 to present its evaluation of the alternative of facility operation with the installation of further severe accident mitigation design features. The NRC staff has discovered no substantial changes in the proposed action as previously evaluated in the Final Environmental Statement that are relevant to environmental concerns nor significant new circumstances or information relevant to environmental concerns and bearing on the licensing of Limerick Generating Station, Units 1 and 2. 15 refs., 10 tabs

  7. Limerick Nuclear Generating Station vibration monitoring system

    International Nuclear Information System (INIS)

    Mikulski, R.

    1988-01-01

    Philadelphia Electric Company utilizes a vibration monitoring computer system at its Limerick Nuclear Generating Station to evaluate machine performance. Performance can be evaluated through instantaneous sampling, online static and transient data. The system functions as an alarm monitor, displaying timely alarm data to the control area. The passage of time since the system's inception has been a learning period. Evaluation through continuous use has led to many enhancements in alarm handling and in the acquisition and display of machine data. Due to the system's sophistication, a routine maintenance program is a necessity. This paper describes the system's diagnostic tools and current utilization. System development and maintenance techniques will also be discussed

  8. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  9. Montague Nuclear Power Station, Units 1 and 2: Final environmental statement (Docket Nos. 50-496 and 50-497)

    International Nuclear Information System (INIS)

    1977-02-01

    The proposed action is the issuance of construction permits to the Northeast Nuclear Energy Company for the construction of the Montague Nuclear Power Station, Units 1 and 2, located on the Connecticut River in the Town of Montague, Massachusetts. The plant will employ two identical boiling-water reactors to produce up to 3579 megawatts thermal (MWt) each. Two steam turbine-generators will use this heat to provide 1150 MWe (net) of electrical power capacity from each turbine-generator. A design power level of 3759 MWt (1220 Mwe net) for each unit is anticipated at a future date and is considered in the assessments contained in this statement. The waste heat will be rejected through natural-draft cooling towers using makeup water obtained from and discharged to the Connecticut River. The 1900-acre site is about 90% forest, with the remaining acreage in transmission-line corridor and old-field vegetation. The total loss of mixed-age forest will be 1273 acres. Nodesignated scenic areas will be crossed. Sixty acres of public lands, State forests, and parks will be lost to transmission facilities as well as losses associated with crossings of 2.0 miles of water bodies and 11.9 miles of wetlands. The maximum estimated potential loss of salable wood products will be $849,600. A maximum of 85.8 cfs of cooling water will be withdrawn from the Connecticut River. A maximum of 17.2 cfs will be returned to the river with the dissolved solids concentration increased by a factor of about 5. A maximum of 68.6 cfs will be evaporated to the atmosphere by the cooling towers. 143 refs., 58 figs., 69 tabs

  10. Experience and development of on-line BWR surveillance system at Onagawa nuclear power station unit-1

    International Nuclear Information System (INIS)

    Kishi, A.; Chiba, K.; Kato, K.; Ebata, S.; Ando, Y.; Sakamoto, H.

    1986-01-01

    ONAGAWA nuclear power station Unit-1 (Tohoku Electric Power Co.) is a BWR-4 nuclear power station of 524 MW electric power which started commercial operation in June 1984. To attain high reliability and applicability for ONAGAWA-1, Tohoku Electric Power Co. and Toshiba started a Research and Development project on plant surveillance and diagnosis from April 1982. Main purposes of this project are to: (1) Develop an on-line surveillance system and acquire its operating experience at a commercial BWR, (2) Assist in plant operation and maintenance by data acquisition and analysis, (3) Develop a new technique for plant surveillance and diagnosis. An outline of the project, operating experience gained from the on-line surveillance system and an introduction to new diagnosis techniques are reported in this paper. (author)

  11. Safety-evaluation report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353)

    International Nuclear Information System (INIS)

    1983-08-01

    The Safety Evaluation Report for the application filed by the Philadelphia Electric Company, as applicant and owner, for licenses to operate the Limerick Generating Station Units 1 and 2 (Docket Nos. 50-352 and 50-353), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Pottstown, Pennsylvania. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  12. Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement No. 6

    International Nuclear Information System (INIS)

    1985-08-01

    In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania. A license for the operation of Limerick Unit 1 was issued on October 26, 1984. The license, which was restricted to a five percent power level, contained conditions which required resolution prior to proceeding beyond the five percent power level. Supplement 4, issued in May 1985, addressed some of these issues. Supplement 4 also contained the comments made by the Advisory Committee on Reactor Safeguards in its report dated November 6, 1984, regarding full power operation of Limerick Unit 1. Supplement 5, issued in July 1985, and this Supplement 6 address further issues, principally the status of offsite emergency planning, that require resolution prior to proceeding beyond the five percent power level

  13. AECL's participation in the commissioning of Point Lepreau generating station unit 1

    International Nuclear Information System (INIS)

    Chawla, S.; Singh, K.; Yerramilli, S.

    1983-05-01

    Support from Atomic Energy of Canada Ltd. (AECL) to Point Lepreau during the commissioning program has been in the form of: seconded staff for commissioning program management, preparation of commissioning procedures, and hands-on commissioning of several systems; analysis of test results; engineering service for problem solving and modifications; design engineering for changes and additions; procurement of urgently-needed parts and materials; technological advice; review of operational limits; interpretation of design manuals and assistance with and preparation of submissions to regulatory authorities; and development of equipment and procedures for inspection and repairs. This, together with AECL's experience in the commissioning of other 600 MWe stations, Douglas Point and Ontario Hydro stations, provides AECL with a wide range of expertise for providing operating station support services for CANDU stations

  14. Safety evaluation report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353)

    International Nuclear Information System (INIS)

    1989-06-01

    In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for licenses to operate the Limerick Generating Station, Units 1 and 2, located on site in Montgomery and Chester Counties, Pennsylvania. Supplement 1 to NUREG-0991 was issued in December 1983. Supplements 2 and 3 were issued in October 1984. License NPF-27 for the low-power operation of Limerick Unit 1 was issued on October 26, 1984. Supplement 4 was issued in May, 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed further issues that required resolution before Unit 1 proceeded beyond the 5-percent power level. The full-power operating license for Limerick Unit 1 (NPF-39) was issued August 8, 1985, and the unit has completed two cycles of operation. Supplement 7 was issued April 1989 to address some of the few significant design differences between Units 1 and 2, the resolution of issues that remained open when the Unit 1 full-power license was issued and an assessment of some of the issues that required resolution before issuance of an operating license for Unit 2. This supplement addresses the remaining issues that required resolution before issuance of and operating license for Unit 2

  15. Damage of the Unit 1 reactor building overhead bridge crane at Onagawa Nuclear Power Station caused by the Great East Japan Earthquake and its repair works

    International Nuclear Information System (INIS)

    Sugamata, Norihiko

    2014-01-01

    The driving shaft bearings of the Unit 1 overhead bridge crane were damaged by the Great East Japan Earthquake at Onagawa Nuclear Power Station. The situation, investigation and repair works of the bearing failure are introduced in this paper. (author)

  16. 78 FR 14361 - In the Matter of Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Units 1 and...

    Science.gov (United States)

    2013-03-05

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0310; Docket Nos. 50-445 and 50-446; License Nos. NPF-87 and NPF-89] In the Matter of Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Units... Nuclear Power Plant, Units 1 and 2 (CPNPP), and its Independent Spent Fuel Storage Installation Facility...

  17. Browns Ferry Nuclear Power Station, Units 1, 2, and 3. Semiannual report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Browns Ferry units 1 and 2 operated at maximum power from January 1 to March 22 except as limited by thermal margins, fuel preconditioning, optimum power shape, maintenance, and Unit 2 start-up tests. On March 22 a cable tray fire started causing spurious starting of equipment due to faulted control cables. The reactors were manually scrammed and placed in cold shutdown for fire investigation, clean up, and fuel removal. Information is also presented concerning maintenance, radiochemistry, occupational radiation exposure, release of radioactive materials, and non-radiological environmental monitoring

  18. Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement 3

    International Nuclear Information System (INIS)

    1984-10-01

    In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the applicant) for licenses to operate the Limerick Generating Station, Units 1 and 2. Supplement 1 was issued in December 1983 and addressed several outstanding issues. Supplement 1 also contains the comments made by the Advisory Committee on Reactor Safeguards in its report dated October 18, 1983. Supplement 2 was issued in October 1984 and addressed fourteen outstanding and fifty-three confirmatory issues and closed them put. This Supplement 3 addresses the remaining issues that require resolution before issuance of the operating license for Unit 1 and closes them out

  19. Nuclear power generation modern power station practice

    CERN Document Server

    1971-01-01

    Nuclear Power Generation focuses on the use of nuclear reactors as heat sources for electricity generation. This volume explains how nuclear energy can be harnessed to produce power by discussing the fundamental physical facts and the properties of matter underlying the operation of a reactor. This book is comprised of five chapters and opens with an overview of nuclear physics, first by considering the structure of matter and basic physical concepts such as atomic structure and nuclear reactions. The second chapter deals with the requirements of a reactor as a heat source, along with the diff

  20. Pilgrim Nuclear Power Station, Unit 1. Annual operating report for 1976

    International Nuclear Information System (INIS)

    Net electrical power generated was 2,415,511 MWH with the generator on line 5,333.6 hrs. Information is presented concerning operations, procedure changes, tests, experiments, maintenance, unit shutdowns and power reductions, and radiation doses to personnel

  1. Technical specifications for Grand Gulf Nuclear Station, Unit 1 (Docket No. 50-416). Appendix A to License No. NPF-13

    International Nuclear Information System (INIS)

    1984-08-01

    The Grand Gulf Nuclear Station, Unit 1 Technical Specifications were prepared by the US Nuclear Regulatory Commission to set forth the limits, operating conditions and other requirements applicable to a nuclear facility as set forth in Section 50.36 of 10 CFR part 50 for the protection of the health and safety of the public

  2. Millstone Nuclear Power Station, Unit 1. Annual operating report: 1 January--31 December 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical power generated was 3,752,445 MWh with the generator on line 6,682 hrs. Information is presented concerning operations; shutdowns and power reductions; refueling; maintenance; failed fuel elements; power generation; occupational personnel radiation exposure; and changes in technical specifications, plant design, and key supervisory personnel

  3. Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530). Supplement No. 8

    International Nuclear Information System (INIS)

    1985-05-01

    Purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of additional information submitted by the applicant since Supplement No. 7 was issued and matters that the staff had under review when Supplement No. 7 was issued, specifically those issues which required resolution prior to plant operation of Unit 1 above 5% full power

  4. Integrated Plant Safety Assessment: Systematic Evaluation Program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245. Final report

    International Nuclear Information System (INIS)

    1983-02-01

    This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in November 1982

  5. Nine Mile Point Nuclear Station, Unit 1. Annual report of operation, 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 3,044,948 MWh(e) with the reactor on line 6,238 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, reportable occurrences, effluent and waste disposal, meteorological summary, and environmental monitoring

  6. Nine Mile Point Nuclear Station, Unit 1. Annual report of operation: January--December 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical power generated was 4,112,827 MWH with the reactor on line 7,727.67 hrs. Information is presented concerning operations, power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, reportable occurrences, and fuel performance

  7. Millstone Nuclear Power Station, Unit 1. Semiannual operating report, July--December 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 1,525,943 MWh(e) with the reactor on line 2,682 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, environmental effects monitoring, release of radioactive materials, and reportable occurrences. Occupational personnel radiation exposures will be submitted later

  8. Millstone Nuclear Power Station, Unit 1. Semiannual operating report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Net electrical power generated was 2,373,130 MWH(e) with the reactor on line 3,915 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, abnormal occurrences, and environmental radiation monitoring. (FS)

  9. 76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

    Science.gov (United States)

    2011-08-08

    ... steam generators remain in place. The facility is permanently shut down in a SAFSTOR condition, defueled..., ``Domestic Licensing of Production and Utilization Facilities,'' to possess but not operate the facility. All...

  10. Three Mile Island Nuclear Station, Unit 1. 1976 annual operating report

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical power generated was 4,335,625 MWh with the reactor on line 5,747.5 hrs. Information is presented concerning operations; specification, procedures, and FSAR changes; corrective maintenance; irradiated fuel examinations; radioactive effluent releases; personnel radiation exposures; shutdowns; and forced power reductions

  11. Pilgrim Nuclear Power Station, Unit 1. Fifth semiannual operating and maintenance report, July--December 1974

    International Nuclear Information System (INIS)

    1974-01-01

    During this period the reactor was critical for 3,550.3 hrs and the net electrical power generated was 1,973,033 MWH. Information is presented concerning operations, maintenance, radioactive effluents, environmental monitoring, and radioactive materials released to unrestricted areas. (U.S.)

  12. Fulton Generating Station Units 1 and 2 (Docket Nos. 50-463 and 50-464): Final environmental statement

    International Nuclear Information System (INIS)

    1975-04-01

    The proposed action is the issuance of construction permits to the Philadelphia Electric Company for the construction of the Fulton Generating Station, Units 1 and 2, located in Fulton and Drumore Townships, Lancaster County, Pennsylvania. Makeup water for cooling will be withdrawn form Conowingo Pond at a maximum rate of 43,000 gpm. The dissolved solids content of the blowdown water will be increased by a factor of about two. The remainder of the water will be evaporated to the atmosphere by cooling towers. About 10 acres offsite, some 7 acres of which is woodland, will be used for railroad-spur construction. About 0.25 mile of new transmission-line rights-of-way (9 acres) will be needed, although 49 miles of new transmission line, which will require about 3 miles of selective clearing, will be constructed on existing rights-of-way. An unestablished amount of land will be used for access-road construction, but the applicant will use existing roadway corridors where feasible. A small loss of consumer species will result from loss of habitat. Some loss of benthic and pelagic organisms in Conowingo Pond will be caused by intake and discharge construction. The Station's thermal and chemical discharges will meet the State water-quality standards. The duration of additional ground-level fog caused by Station operation is expected to be less than 3 hr/year. (Sect. 5.3.3). No observable effects are expected from salt deposition from cooling-tower drift. (Sect. 5.3.3). Decomposers, primary producers, and zooplankton will be entrained and killed in the cooling-tower system; they, as well as benthic organisms, will be affected by the heated-water discharge. This loss will have little effect on the pond food web. 30 figs., 76 tabs

  13. Safety-evaluation report related to operation of McGuire Nuclear Station, Units 1 and 2. Docket Nos. 50-369 and 50-370

    International Nuclear Information System (INIS)

    1983-05-01

    This report supplements the Safety Evaluation Report Related to the Operation of McGuire Nuclear Station, Units 1 and 2 (SER (NUREG-0422)) issued in March 1978 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, as applicant and owner, for licenses to operate the McGuire Nuclear Station, Units 1 and 2 (Docket Nos. 50-369 and 50-370). The facility is located in Mecklenburg County, North Carolina, about 17 mi north-northwest of Charlotte, North Carolina. This supplement provides information related to issuance of a full-power authorization for Unit 2. The staff concludes that the McGuire Nuclear Station can be operated by the licensee without endangering the health and safety of the public

  14. Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement No. 4

    International Nuclear Information System (INIS)

    1985-05-01

    In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the applicant) for licenses to operate the Limerick Generating Station, Units 1 and 2 located on a site in Montgomery and Chester Counties, Pennsylvania. A license (NPF-27) for the operation of Limerick Unit 1 was issued on October 26, 1984. The license, which was restricted to a five percent power level, contained conditions which required resolution prior to proceeding beyond the five percent power level. This Supplement 4 to the SER addresses some of those technical issues and their associated license conditions which require resolution prior to proceeding beyond the five percent power level. The remaining issues to be addressed prior to proceeding beyond the five percent power level will be addressed in a later supplement to this report. This Supplement 4 to the SER also contains the comments made by the Advisory Committee on Reactor Safeguards in its report dated November 6, 1984, regarding full power operation of Limerick Unit 1

  15. Safety evaluation report related to the operation of Limerick Generating Station, Units 1 and 2, Docket nos. 50-352 and 50-353

    International Nuclear Information System (INIS)

    1989-08-01

    In August 1983 the staff of the Nuclear Regulatory commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for the licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania. Supplement 1 was issued in December 1983. Supplement 2 was issued in October 1984. Supplement 3 was issued in October 1984. Supplement 4 was issued in May 1985. Supplement 5 was issued in July 1985. Supplement 6 was issued in August 1985 and Supplement 7 was issued in April 1989. Supplement 7 addresses the major design differences between Units 1 and 2, the resolution of all issues that remained open when the Unit 1 full-power license was issued, the staff's assessment regarding the application by the licensee to operate Unit 2 and issues that require resolution before issuance of an operating license for Unit 2. Supplements 8 and 9 address further issues that require resolution prior to issuance of an operating license. 1 tab

  16. 75 FR 44292 - Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2; Notice of...

    Science.gov (United States)

    2010-07-28

    ... and DPR-60] Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2... assessment, and behavioral observation) of the unescorted access authorization program when making the... under consideration to determine whether it met the criteria established in NRC Management Directive (MD...

  17. Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417). Supplement 6

    International Nuclear Information System (INIS)

    1984-08-01

    Supplement 6 to the Safety Evaluation Report for Mississippi Power and Light Company et al. joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the NRC staff's evaluation of open items from previous supplements and Technical Specification changes required before authorizing operation of Unit 1 above 5% of rated power

  18. Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417). Supplement No. 5

    International Nuclear Information System (INIS)

    1984-08-01

    Supplement 5 to the Safety Evaluation Report for Mississippi Power and Light Company, et al., joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status on the resolution of those issues that require further evaluation before authorizing operation of Unit 1 above 5% of rated power

  19. Safety-Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2. Docket Nos. 50-416 and 50-417

    International Nuclear Information System (INIS)

    1983-05-01

    Supplement 4 to the Safety Evaluation Report for Mississippi Power and Light Company, et. al., joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status on the resolution of those issues that required further evaluation before authorizing operation of Unit 1 above 5% rated power and other issues that were to be evaluated during the first cycle of power operation

  20. Draft environmental statement related to the operation of Limerick Generating Station, Units 1 and 2. Docket Nos. 50-352 and 50-353

    International Nuclear Information System (INIS)

    1983-06-01

    This Draft Environmental Statement contains the second assessment of the environmental impact associated with the operation of the Limerick Generating Station, Units 1 and 2, pursuant to the National Environment Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs. Land use and terrestrial and aquatic ecological impacts will be small. Operational impacts to historic and archeological sites will be negligible. The effects of routine operations, energy transmission, and periodic maintenance of rights of way and transmission facilities should not jeopardize any populations of endangered or threatened species. No significant impacts are anticipated from normal operational releases of radioactivity

  1. Integrated-plant-safety assessment Systematic Evaluation program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245

    International Nuclear Information System (INIS)

    1982-11-01

    The Systematic Evaluation Program was initiated in February 1977 to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

  2. The paperclip and the nuclear generating station

    International Nuclear Information System (INIS)

    Mussard, J.A.

    1981-01-01

    The article presents some reflections upon the circumstances of the recent rejection by the Swiss Federal Government of a project for building a nuclear generating plant at Kaiseraugst. The following points are made: The use of conventional publicity and public relations techniques to try to convince the public of the desirability or at least of the harmlessness of such projects may very well be counter-productive, given the public's not altogether ill-founded suspicion of such types of pressure. Nor is it helpful to accuse opponents of nuclear developments of indulging in emotional reaction, emotion being entirely legitimate. The proponents of such schemes should confine themselves to objective discussion of the questions Where How Why and At what cost (cost being interpreted in the widest, not merely financial, sense). They should avoid the trap of appearing to be for (as distinct from against) nuclear energy. Finally both sides should abjure (and so far as possible the community should outlaw) methods of conducting disputes that border on lying, charlatanism, demagogy and above all, illegality, and confine themselves to serious discussion of the questions that arise, which are far from being confined to technology and economics. (C.J.O.G.)

  3. Browns Ferry Nuclear Power Station, Units 1, 2, and 3. Annual operating report: January--December 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Units 1 and 2 were down for the first half of the year caused by the fire of March 1975. Net electrical power generated by Unit 1 this year was 1,301,183 MWH with the generator on line 2,175.25 hrs. Unit 2 generated 1,567,170 MWH with the generator on line 2,548.73 hrs. Unit 3 began full power operation on November 20th and generated 1,416,891 MWH with the generator on line 2,058.20 hrs. Information is presented concerning operations, fuel performance, surveillance testing, containment leak testing, changes, power generation, shutdown and forced reductions, coolant chemistry, occupational radiation exposures, and maintenance

  4. Dresden Nuclear Power Station, Units 1, 2, and 3. Semiannual report on operating and maintenance, July--December 1974

    International Nuclear Information System (INIS)

    1975-01-01

    Unit 1 generated 388,882 MWH(e) and was on line 3111.2 hours, Unit 2 generated 1,204,106 MWH(e) and was on line 2013.4 hours, and Unit 3 generated 2,250,810 MWH(e) and was on line 3836 hours. Information is presented concerning operations, shutdowns, maintenance, changes, tests, and experiments for the three units. (U.S.)

  5. Revised draft environmental statement related to construction of Atlantic Generating Station Units 1 and 2 (Docket Nos. STN 50-477 and STN 50-478)

    International Nuclear Information System (INIS)

    1976-10-01

    The proposed action is the issuance of a construction permit to Public Service Electric and Gas Company (PDE and G) for the construction of the Atlantic Generating Station (AGS), Units 1 and 2. The AGS is the first nuclear power station in the United States proposed for construction in the offshore waters on the continental shelf. The AHS will be located in the Atlantic Ocean 2.8 miles offshore of Atlantic and Ocean countries. New Jersey, 11 miles northeast of Atlantic City, and will consist of two floating nuclear power plants enclosed in a protective rubble-mound breakwater. Both plants will be identical, of standardized design, and will employ pressurized water reactors to produce up to approximately 3425 megawatts thermal (MWt) each. Steam turbine generators will use this heat to produce up to approximately 1150 megawatts of electrical power (MWe) per unit. The main condensers will be cooled by the flow of seawater drawn from within the breakwater and discharged shoreward and external to the breakwater. This statement identifies various environmental aspects and potential adverse effects associated with the construction and operation of the AGS. Based upon an approximate two-year review period which included a multidisciplined assessment of extensive survey and modeling data, these effects are considered by the staff to be of a generally acceptable nature. Breakwater construction will result in the destruction of 100 acres of benthic infauna (burrowing animals) and the development of a reef-type community on the breakwater. The production of new biomass (standing crop) by the reef community is expected to compensate for the infaunal biomass destroyed by dredging and will contribute mainly to the local sport fishery. 93 figs., 110 tabs

  6. Activity transport in nuclear generating stations

    International Nuclear Information System (INIS)

    Mitchell, A.B.

    1975-01-01

    The objective of this paper is to give a basic understanding of the operational limitations caused by radiation fields in the present design of CANDU-PHW reactors. A simple model of activity transport is described, and the significance of various radioisotopes identified. The impact which radiation fields have at the Divisional, Station Manager and Operation levels, is outlined in the context of typical work situations. (author)

  7. Assessment of the impacts of the Salem and Hope Creek Stations on shortnose sturgeon, 'Acipenser brevirostrum' lesueur. Technical report

    International Nuclear Information System (INIS)

    Masnik, M.T.; Wilson, J.H.

    1980-04-01

    Acipenser brevirostrum, a federally recognized endangered species, is known from the Delaware River in the vicinity of the Salem and Hope Creek Nuclear Stations. The potential for adverse impact to this species due to the completion of construction and the operation of these stations is assessed. A review of the literature pertaining to the distribution, life history, and tolerance of this species is presented. Both historical records and current status of the shortnose sturgeon population in the Delaware River are discussed

  8. Technical evaluation of RETS-required reports for Zion Nuclear Power Station, Units 1 and 2 for 1983

    International Nuclear Information System (INIS)

    Young, T.E.; Magleby, E.H.

    1985-01-01

    A review was performed on the reports required by Federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted at Commonwealth Edison's Zion Nuclear Power Station during 1983. The two periodic reports reviewed: (1) were the Effluent and Waste Disposal Semiannual Report, July-December 1983, and (2) the Radioactive Waste and Environmental Monitoring Annual Report - 1983. The principal review guidelines were the plant's specific RETS and NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

  9. Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414)

    International Nuclear Information System (INIS)

    1984-07-01

    The report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for fuel loading and precriticality testing for Unit 1

  10. Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414)

    International Nuclear Information System (INIS)

    1984-12-01

    This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power opertion for Unit 1

  11. Valve maintainability in CANDU-PHW nuclear generating stations

    International Nuclear Information System (INIS)

    Pothier, N.E.; Crago, W.A.

    1977-09-01

    Design, application, layout and administrative factors which affect valve maintainability in CANDU-PHW power reactors are identified and discussed. Some of these are illustrated by examples based on prototype reactor operation experience. Valve maintainability improvements resulting from laboratory development and maintainability analysis, have been incorporated in commercial CANDU-PHW nuclear generating stations. These, also, are discussed and illustrated. (author)

  12. A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station

    International Nuclear Information System (INIS)

    Vo, T.; Gore, B.; Simonen, F.; Doctor, S.

    1994-08-01

    As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish in-service inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction (i.e., 5%) of the total PRA-estimated risk for core damage. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilistics are maintained

  13. Iron turbidity removal from the active process water system of the Kaiga Generating Station Unit 1 using an electrochemical filter

    International Nuclear Information System (INIS)

    Venkateswaran, G.; Gokhale, B.K.

    2007-01-01

    Iron turbidity is observed in the intermediate cooling circuit of the active process water system (APWS) of Kaiga Generating Station (KGS). Deposition of hydrous/hydrated oxides of iron on the plate type heat exchanger, which is employed to transfer heat from the APWS to the active process cooling water system (APCWS), can in turn result in higher moderator D 2 O temperatures due to reduced heat transfer. Characterization of turbidity showed that the major component is γ-FeOOH. An in-house designed and fabricated electrochemical filter (ECF) containing an alternate array of 33 pairs of cathode and anode graphite felts was successfully tested for the removal of iron turbidity from the APWS of Kaiga Generating Station Unit No. 1 (KGS No. 1). A total volume of 52.5 m 3 water was processed using the filter. At an average inlet turbidity of 5.6 nephelometric turbidity units (NTU), the outlet turbidity observed from the ECF was 1.6 NTU. A maximum flow rate (10 L . min -1 ) and applied potential of 18.0-20.0 V was found to yield an average turbidity-removal efficiency of ∝ 75 %. When the experiment was terminated, a throughput of > 2.08 . 10 5 NTU-liters was realized without any reduction in the removal efficiency. Removal of the internals of the filter showed that only the bottom 11 pairs of felts had brownish deposits, while the remaining felts looked clean and unused. (orig.)

  14. Technical evaluation report on the monitoring of electric power to the reactor protection system for the Nine Mile Point Nuclear Station, Unit 1 (Docket No. 50-220)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1984-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Nine Mile Point Nuclear Station, Unit 1. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  15. Development and application of the lancing system of delta-60 steam generator-Kori nuclear power plant unit 1

    International Nuclear Information System (INIS)

    Jeong, W. T.; Han, D. Y.; Ahn, N. S.; Jo, B. H.; Hong, Y. W.

    2001-01-01

    A lancing system for removing the deposits on the tube sheet of a nuclear steam generator using high pressure water was developed and applied to Kori Nuclear Power Plant( NPP) Unit 1. As the place where the lancing system is to be installed is relatively high radioactive area, every part consisting the equipment is carefully selected to be radiation resistant. The lancing robot was designed to be water proof to aviod possible malfunction of the lancing robot because of high pressure water. To minimize radiation exposure to operators, the system was designed considering easy installation and maintenance in mind. Water ejection nozzle are designed to have high strength with special material and heat treatment so as to lessen abrasion caused by high pressure ejection. The lancing system showed good performance during the on-site lancing using the system for Delta-60 steam generator of Kori NPP No. 1 in October 2000

  16. A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

    International Nuclear Information System (INIS)

    Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A.; Doctor, S.R.

    1997-02-01

    As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems

  17. Tritium in groundwater investigation at the Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    DeWilde, J.; Yu, L.; Wootton, R.; Belanger, D.; Hansen, K.; McGurk, E.; Teare, A.

    2001-01-01

    Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radionuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identity the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

  18. Monticello Nuclear Generating Plant, Unit 1. Semiannual operating report, July--December 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 1,103,754 MWH(e) with the reactor on line 2,743 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, and environmental radiation monitoring

  19. Monticello Nuclear Generating Plant, Unit 1. Semiannual operating report No. 9, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Net electrical power generated was 1,775,704 MWH with the reactor on line 3,583 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, abnormal occurrences, and radiation environmental monitoring. (FS)

  20. Improving nuclear generating station response for electrical grid islanding

    International Nuclear Information System (INIS)

    Chou, Q.B.; Kundur, P.; Acchione, P.N.; Lautsch, B.

    1989-01-01

    This paper describes problems associated with the performance characteristics of nuclear generating stations which do not have their overall plant control design functions co-ordinated with the other grid controls. The paper presents some design changes to typical nuclear plant controls which result in a significant improvement in both the performance of the grid island and the chances of the nuclear units staying on-line following the disturbance. This paper focuses on four areas of the overall unit controls and turbine governor controls which could be modified to better co-ordinate the control functions of the nuclear units with the electrical grid. Some simulation results are presented to show the performance of a typical electrical grid island containing a nuclear unit with and without the changes

  1. Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1981-07-01

    This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected

  2. Technical evaluation report on the proposed design modifications and technical specification changes on grid voltage degradation for the Millstone Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the proposed design modifications and Technical Specification change for protection of Class 1E equipment from grid voltage degradation for the Millstone Nuclear Power Station, Unit 1. The review criteria are based on several IEEE standards and the Code of Federal Regulations. The evaluation finds that the licensee has not provided sufficient information on the undervoltage protection system to allow a complete evaluation into the adequacy of protecting the Class 1E equipment from sustained voltage degradation

  3. Safety evaluation report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352-50-353). Supplement 1

    International Nuclear Information System (INIS)

    1983-12-01

    This report supplements the Safety Evaluation Report (NUREG-0991, August 1983) for the application filed by the Philadelphia Electric Company, as applicant and owner, for licenses to operate the Limerick Generating Station Units 1 and 2 (Docket Nos. 50-352 and 50-353). The facility is located near Pottstown, Pennsylvania. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  4. Online control loop tuning in Pickering Nuclear Generating Stations

    International Nuclear Information System (INIS)

    Yu, K.X.; Harrington, S.

    2008-01-01

    Most analog controllers in the Pickering B Nuclear Generating Stations adopted PID control scheme. In replacing the analog controllers with digital controllers, the PID control strategies, including the original tuning parameters were retained. The replacement strategy resulted in minimum effort on control loop tuning. In a few cases, however, it was found during commissioning that control loop tuning was required as a result of poor control loop performance, typically due to slow response and controlled process oscillation. Several factors are accounted for the necessities of control loop re-tuning. Our experience in commissioning the digital controllers showed that online control tuning posted some challenges in nuclear power plant. (author)

  5. Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414. Suppl. 1

    International Nuclear Information System (INIS)

    1983-04-01

    This reort supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report, and discusses the recommendations of the Advisory Committee on Reactor Safeguards in its report dated March 15, 1983

  6. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  7. Safety Evaluation Report related to the operation of Cartawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement No. 5

    International Nuclear Information System (INIS)

    1986-02-01

    This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, Saluda River Electric Cooperative, Inc., and Piedmont Municipal Power Agency, as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power operation for Unit 2

  8. Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417)

    International Nuclear Information System (INIS)

    1984-10-01

    This report supplements the Safety Evaluation Report (NUREG-0831) issued in September 1981 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Mississippi Power and Light (MP and L) Company, Middle South Energy, Inc., and South Mississippi Electric Power Association as applicants and owners, for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417, respectively). The facility is located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi. This supplement provides information on the NRC staff's evaluation of requests for exemptions to NRC regulations pursuant to the Commission's direction in CLI-84-19, dated October 25, 1984

  9. Safety Evaluation Report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement No. 6

    International Nuclear Information System (INIS)

    1986-05-01

    This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, Saluda River Electric Cooperative, Inc., and Piedmont Municipal Power Agency, as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 miles) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for operation above 5% power and power ascension to full-power operation for Unit 2

  10. Foreign Material Exclusion Program at CNE Cernavoda Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Urjan, Daniel [S.N. ' Nuclearelectrica' SA, CNE Cernavoda Nuclear Power Plant, Medgidiei 2 Street, 905200 Cernavoda, Constanta (Romania)

    2008-07-01

    In the face of a continuing attention to operations and maintenance costs at nuclear power plants, the future of the industry depends largely upon increasing plant availability and improving operating efficiency. The success in achieving these objectives is dependent upon the success of each plant's equipment maintenance program. Preventing the introduction of foreign materials into a nuclear power plant system or component requires a careful, thoughtful, and professional approach by all site personnel. This paper describes a proactive approach to prevent the introduction of foreign material into systems and components, by providing an overview of technical considerations required to develop, implement, and manage a foreign material exclusion program at CNE Cernavoda Unit 1 and 2 Nuclear Power Station. It is also described an example of Foreign Material Intrusion which happened during the 2003 planned maintenance outage at Cernavoda Unit no.1. This paper also defines personnel responsibilities and key nomenclature and a means for evaluating prospective work tasks and activities against standardized criteria, in order to identify the appropriate level of the required FME controls. (author)

  11. Foreign Material Exclusion Program at CNE Cernavoda Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Urjan, Daniel [S.N. ' Nuclearelectrica' SA, CNE Cernavoda Nuclear Power Plant, Medgidiei 2 Street, 905200 Cernavoda, Constanta (Romania)

    2008-07-01

    In the face of a continuing attention to operations and maintenance costs at nuclear power plants, the future of the industry depends largely upon increasing plant availability and improving operating efficiency. The success in achieving these objectives is dependent upon the success of each plant's equipment maintenance program. Preventing the introduction of foreign materials into a nuclear power plant system or component requires a careful, thoughtful, and professional approach by all site personnel. This paper describes a proactive approach to prevent the introduction of foreign material into systems and components, by providing an overview of technical considerations required to develop, implement, and manage a foreign material exclusion program at CNE Cernavoda Unit 1 and 2 Nuclear Power Station. It is also described an example of Foreign Material Intrusion which happened during the 2003 planned maintenance outage at Cernavoda Unit no.1. This paper also defines personnel responsibilities and key nomenclature and a means for evaluating prospective work tasks and activities against standardized criteria, in order to identify the appropriate level of the required FME controls. (author)

  12. Foreign Material Exclusion Program at CNE Cernavoda Nuclear Generating Station

    International Nuclear Information System (INIS)

    Urjan, Daniel

    2008-01-01

    In the face of a continuing attention to operations and maintenance costs at nuclear power plants, the future of the industry depends largely upon increasing plant availability and improving operating efficiency. The success in achieving these objectives is dependent upon the success of each plant's equipment maintenance program. Preventing the introduction of foreign materials into a nuclear power plant system or component requires a careful, thoughtful, and professional approach by all site personnel. This paper describes a proactive approach to prevent the introduction of foreign material into systems and components, by providing an overview of technical considerations required to develop, implement, and manage a foreign material exclusion program at CNE Cernavoda Unit 1 and 2 Nuclear Power Station. It is also described an example of Foreign Material Intrusion which happened during the 2003 planned maintenance outage at Cernavoda Unit no.1. This paper also defines personnel responsibilities and key nomenclature and a means for evaluating prospective work tasks and activities against standardized criteria, in order to identify the appropriate level of the required FME controls. (author)

  13. Technical evaluation report on the adequacy of station electric-distribution-system voltages for the Millstone Nuclear Power Station, Units 1 and 2. Docket Nos. 50-245, 50-336

    International Nuclear Information System (INIS)

    Selan, J.C.

    1983-01-01

    This report documents the technical evaluation of the adequacy of the station electric-distribution-system voltages for the Millstone Nuclear Power Station, Units 1 and 2. The evaluation is to determine if the onsite distribution system, in conjunction with the offsite power sources, has sufficient capacity to automatically start and operate all Class 1E loads within the equipment voltage ratings under certain conditions established by the Nuclear Regulatory Commission. The analyses submitted demonstrate that adequate voltages will be supplied to the Class 1E equipment under the worst-case conditions analyzed

  14. Vinten exposure measurements of the Salem Unit 1 lower core barrel

    International Nuclear Information System (INIS)

    Glennon, P.T.

    1988-01-01

    On November 6, 1987, the lower core barrel of Salem Unit I was removed from the reactor vessel and placed in the refueling pool as part of the unit's ten year inspection program. This paper deals with the supporting actions of the dosimetry group of PSE ampersand G. Prior to the move of the lower core barrel, Westinghouse predicted dose rates at one foot in water as a function of axial distance along the core barrel. This prediction was used in planning the health physics requirements associated with the move. It was agreed that a measurement of the axial dose rates would either lend confidence to the predictions or identify weaknesses in them

  15. Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement No. 5

    International Nuclear Information System (INIS)

    1985-07-01

    In August 1983 the NRC issues its Safety Evaluation Report regarding the application for licenses to operate the Limerick Generating Station, Units 1 and 2 located on a site in Montgomery and Chester Counties, Pennsylvania. Supplement 1 was issued in December 1983 and addressed several outstanding issues. SSER 1 also contains the comments made by the Advisory Committee on Reactor Safeguards in its interim report dated October 18, 1983. Supplement 2 was issued in October 1984. Supplement 3 was issued in October 1984 and addressed the remaining issues that required resolution before issuance of the operating licence for Unit 1. On October 26, 1984 a license (NPF-27) for Unit 1 was issued which was restricted to a five percent power level and contained conditions which required resolution prior to proceeding beyond the five percent power level. Supplement 4 issued in May 1985 addressed some of the technical issues and their associated license conditions, which required resolution prior to proceeding beyond the five percent power level. SSER 4 also contained the comments made by the Advisory Committee on Reactor Safeguards in its report dated November 6, 1984. This Supplement 5 to the SER addresses further issues that require resolution prior to proceeding beyond the five percent power level

  16. 78 FR 41425 - In the Matter of Duke Energy Carolinas, LLC; (Oconee Nuclear Station, Units 1, 2, and 3...

    Science.gov (United States)

    2013-07-10

    ... approved plant procedures and the completion of operator training necessary to accomplish this lineup, and a combination of testing and engineering evaluation in accordance with station procedures which... credited. This modification along with approved plant procedures and the completion of operator training...

  17. Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement 2

    International Nuclear Information System (INIS)

    1984-06-01

    This report supplements the Safety Evaluation Report (NUREG-0954) and Supplement 1 with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos., 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory issues and license conditions identified in the Safety Evaluation Report

  18. 75 FR 12312 - South Carolina Electric and Gas Company; Virgil C. Summer Nuclear Station, Unit 1; Exemption

    Science.gov (United States)

    2010-03-15

    ... cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation. The... of energy release, hydrogen concentration, and cladding oxidation from the metal-water reaction shall... pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or...

  19. Steam generator replacement at the Obrigheim nuclear power station

    International Nuclear Information System (INIS)

    Pickel, E.; Schenk, H.; Huemmler, A.

    1984-01-01

    The Obrigheim Nuclear Power Station (KWO) is equipped with a dual-loop pressurized water reactor of 345 MW electric power; it was built by Siemens in the period 1965 to 1968. By the end of 1983, KWO had produced some 35 billion kWh in 109,000 hours of operation. Repeated leaks in the heater tubes of the two steam generators had occurred since 1971. Both steam generators were replaced in the course of the 1983 annual revision. Kraftwerk Union AG (KWU) was commissioned to plant and carry out the replacement work. Despite the leakages the steam generators had been run safely and reliably over a period of 14 years until their replacement. Replacing the steam generators was completed within twelve weeks. In addition to the KWO staff and the supervising crew of KWU, some 400 external fitters were employed on the job at peak work-load periods. For the revision of the whole plant, work on the emergency systems and replacement of the steam generators a maximum number of approx. 900 external fitters were employed in the plant in addition to some 250 members of the plant crew. The exposure dose of the personnel sustained in the course of the steam generator replacement was 690 man-rem, which was clearly below previous estimates. (orig.) [de

  20. 76 FR 30206 - Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant, Unit 1 and 2; Notice...

    Science.gov (United States)

    2011-05-24

    ... Operating Company, Inc., Vogtle Electric Generating Plant, Unit 1 and 2; Notice of Consideration of Issuance..., http://www.regulations.gov . Because your comments will not be edited to remove any identifying or... received from other persons for submission to the NRC inform those persons that the NRC will not edit their...

  1. Browns Ferry Nuclear Power Station, Units 1, 2, and 3. Annual operating report: 1 January--31 December 1977

    International Nuclear Information System (INIS)

    1978-01-01

    The three reactors operated at near full power generating 17,622,500 MWH gross electrical power. There were 153 major power reductions including 50 scrams mostly caused by equipment failures. Data are presented concerning operations, modifications, occupational exposures, effluent activity, and waste disposal

  2. Evaluation of River Bend Station Unit 1 Technical Specifications

    International Nuclear Information System (INIS)

    Baxter, D.E.; Bruske, S.J.

    1985-08-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the River Bend Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the River Bend T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The River Bend Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

  3. Cernavoda NPP Unit 1 - a plant of several generations

    International Nuclear Information System (INIS)

    Rotaru, I.; Metes, M.; Anghelescu, M.S.

    2001-01-01

    The paper reflects some key aspects related to the shift of generations during the project's development, including the present stage. Further, the place of Cernavoda NPP Unit 1 in the Romanian power sector and among other nuclear stations in the world is presented. The operational performances achieved 'in service' up to the end of 1999, with reference to the performance indicators for electrical energy production, nuclear safety, radiation protection, radioactive wastes and nuclear fuel are illustrated. For all of these items, comparisons are performed with similar indicators reported by other worldwide nuclear power plants, in order to assess our results. Finally, some comments about Cernavoda NPP Unit 2 project status and need to completion and commissioning it are included. (authors)

  4. Steam generator maintenance and life management at Embalse Nuclear Station

    International Nuclear Information System (INIS)

    Sainz, R.; Diaz, G.; Sveruga, H.; Ramakrishnan, T.K.; Azeez, S.

    2004-01-01

    The Embalse Nuclear Station has four steam generators (SGs) with inverted vertical U tubes manufactured by Babcock and Wilcox Canada (B and W). These are main components, both from the operative point of view as the heat transfer from the Primary Heat Transport System (PHTS) to the Secondary System, and from the point of view of safety, as they are the part of the PHTS and its radioactive inventory pressure barrier. In addition, they are one of the most important cost-related elements for potential life extensions. Maintenance and inspections are carried out in order to maintain a high availability of the SGs, as they have had a positive impact on the operational availability of the plant, and to reduce the tube failure probabilities, thus minimizing the amount of radioactive effluents and taking care of the condition of the main components in order to enable the plant life management and the planning of the plant life extension. The most relevant maintenance activities performed have been the inspections performed on 100% of the tubes every 3 years. the mechanical cleaning of the inside of the tubes, the sludge removal from the secondary side tubesheet, the divider plate replacement, and the inspection of internals of the secondary side.Thanks to the latter and to the eddy current inspections, the degradation in the U-bend supports was detected early and every effort is being made to repair them shortly. Besides, a life management program has been started covering the entire plant starting with this important component. The Embalse Nuclear Station's SGs show a low percentage of plugged tubes compared to other stations in similar conditions, but they must be monitored continually and systematically if a life extension is intended. (author)

  5. Final environmental statement: Related to the operation of Davis-Besse Nuclear Power Station, Unit 1 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of an operating license to the Toledo Edison Company and the Cleveland Electric Illuminating Company for the startup and operation of the Davis-Besse Nuclear Power Station Unit 1 (the station) located near Port Clinton in Ottawa County, Ohio. The total site area is 954 acres of which 160 acres have been removed from production of grain crops and converted to industrial use. Approximately 600 acres of the area is marshland which will be maintained as a wildlife refuge. The disturbance of the lake shore and lake bottom during construction of the station water intake and discharge pipes resulted in temporary turbidity, silting, and destruction of bottom organisms. Since completion of these activities, evidence of improvement in turbidity and transparency measurements, and the reestablishment of the bottom organism has been obtained. The cooling tower blowdown and service water which the station discharges to Lake Erie, via a submerged jet, will be heated no more than 20/degrees/F above the ambient lake water temperature. Although some small fish and plankton in the discharge water plume will be disabled as a result of thermal shock, exposure to chlorine and buffeting, few adult fish will be affected. The thermal plume resulting from the maximum thermal discharge is calculated to have an area of less than one acre within the 3/degrees/F isotherm (above lake ambient). Approximately 101 miles of transmission lines have been constructed, primarily over existing farmland, requiring about 1800 acres of land for the rights-of-way. Land use will essentially be unchanged since only the land required for the base of the towers is removed from production. Herbicides will not be used to maintain the rights-of-way. 14 figs., 34 refs

  6. Public response to the Diablo Canyon Nuclear Generating Station

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-01-01

    The authors examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response

  7. Public response to the Diablo Canyon Nuclear Generating Station

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-01-01

    We examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response. (author)

  8. Public response to the Diablo Canyon Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Pijawka, K D [Arizona State Univ., Tempe (USA)

    1982-08-01

    We examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response.

  9. Perkins Nuclear Station, Units 1, 2, and 3: Final environmental statement (Docket Nos. STN 50-488, STN 50-489, and STN 50-490

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of a construction permit to the Duke Power Company for the construction of the Perkins Nuclear Station (PNS) Units 1, 2, and 3 located in Davie County, North Carolina. A total of 2402 acres will be used for the PNS site; another 1401 acres will be used for the Carter Creek Impoundment. Construction-related activities on the primary site will disturb about 617 acres. Approximately 631 acres of land will be required for transmission line right-of-way, and a railroad spur will affect 77 acres. This constitutes a minor local impact. The heat dissipation system will require a maximum water makeup of 55,816 gpm, of which 50,514 gpm will be consumed due to drift and evaporative losses. This amount represents 4% of the mean monthly flow of the Yadkin River. The cooling tower blowdown and chemical effluents from the station will increase the dissolved solids concentration in the Yadkin River by a maximum of 18 ppm. The thermal alterations and increases in total dissolved solids concentration will not significantly affect the aquatic productivity of the Yadkin River. 26 figs., 51 tabs

  10. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

  11. 76 FR 1197 - Arizona Public Service Company, Palo Verde Nuclear Generating Station; Notice of Availability of...

    Science.gov (United States)

    2011-01-07

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-528, 50-529, 50-530; NRC-2009-0012] Arizona Public Service Company, Palo Verde Nuclear Generating Station; Notice of Availability of the Final Supplement 43... of operation for the Palo Verde Nuclear Generating Station (PVNGS). Possible alternatives to the...

  12. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-15

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century.

  13. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-01

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  14. Socio-economic impacts of nuclear generating stations: Arkansas Nuclear One Station case study

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Arkansas Nuclear One nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  15. Phytotoxicology section investigation in the vicinity of the Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the Darlington Nuclear Generating Station, in October, 1989

    International Nuclear Information System (INIS)

    1991-02-01

    The Phytotoxicology Section, Air Resources Branch is a participant in the Pickering and Bruce Nuclear Contingency Plans. The Phytotoxicology Emergency Response Team is responsible for collecting vegetation samples in the event of a nuclear emergency at any of the nuclear generating stations in the province. As part of its responsibility the Phytotoxicology Section collects samples around the nuclear generating stations for comparison purposes in the event of an emergency. Because of the limited frequency of sampling, the data from the surveys are not intended to be used as part of a regulatory monitoring program. These data represent an effort by the MOE to begin to establish a data base of tritium concentrations in vegetation. The Phytotoxicology Section has carried out seven surveys in the vicinity of Ontario Hydro nuclear generating stations since 1981. Surveys were conducted for tritium in snow in the vicinity of Bruce Nuclear Power Development (BNPD), February, 1981; tritium in cell-free water of white ash in the vicinity of BNPD, September, 1981; tritium in snow in the vicinity of BNPD, March, 1982; tritium in tree sap in the vicinity of BNPD, April, 1982; tritium in tree sap in the vicinity of BNPD, April, 1984, tritium in the cell-free water of white ash in the vicinity of BNPD, September, 1985; and, tritium in cell-free water of grass in the vicinity of Pickering Nuclear Generation Station (PNGS), October 1986. In all cases a pattern of decreasing tritium levels with increasing distance from the stations was observed. In October, 1989, assessment surveys were conducted around Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the new Darlington Nuclear Generating Station (DNGS). The purpose of these surveys was to provide baseline data for tritium in cell-free water of grass at all three locations at the same time of year. As none of the reactor units at DNGS had been brought on line at the time of the survey, this data was to be

  16. Chemistry technician performance evaluation program Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Shawver, J.M.

    1992-01-01

    The Arizona Nuclear Power Project (ANPP), a three-reactor site located 50 miles west of Phoenix, Arizona, has developed and implemented a program for evaluating individual chemistry technician analytical performance on a routine basis. About 45 chemistry technicians are employed at the site, 15 at each operating unit. The technicians routinely perform trace level analyses for impurities of concern to PWRs. Each month a set of blind samples is provided by an outside vendor. The blind samples contain 16 parameters which are matrixed to approximate the PWR's primary and secondary cycles. Nine technicians receive the samples, three from each operating unit, and perform the required analyses. Acceptance criteria for successful performance on the blind parameters is based on the values found in the Institute of Nuclear Power Operations (INPO) Document 83-016, Revision 2, August 1989, Chemistry Quality Control Program. The goal of the program is to have each technician demonstrate acceptable performance on each of 16 analytical parameters. On completion of each monthly set, a summary report of all of the analytical results for the sample set is prepared. From the summary report, analytical bias can be detected, technician performance is documented, and overall laboratory performance can be evaluated. The program has been very successful at satisfying the INPO requirement that the analytical performance of each individual technician should be checked on at least a six-month frequency for all important parameters measured. This paper describes the program as implemented at the Palo Verde Nuclear Generating Station and provides a summary report and trend and bias graphs for illustrative purposes

  17. Decontamination and disposal of Sb-124 at Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Miller, A.D.; Hillmer, T.P.; Kester, J.W.; Hensch, J.R.

    1988-01-01

    Palo Verde Nuclear Generating Station (PVNGS) is a three unit Combustion Engineering pressurized water reactor site. Each unit consists of an identical, self contained 1270 MWe reactor. This standardized design allows sharing of design improvements and equipment leading to optimum operation of the individual units. One design improvement, identified early into the operation of Unit 1, involved the elemental antimony content of the seals and bearings within the reactor coolant pumps. Normal wear of these components releases small amounts of elemental antimony. This antimony in turn deposits on in-core surfaces and activates to produce the isotopes Sb-122 and Sb-124. These isotopes emit highly energetic gamma rays which contribute significantly to the exposure and radwaste disposal charges at PVNGS. For these reasons, the Antimony Removal Program was undertaken to remove the radioactive and elemental antimony from the nuclear steam supply system at all three units. The work presented here describes the antimony decontamination and disposal

  18. Experience with reactor power cutback system at Palo Verde nuclear generating station

    International Nuclear Information System (INIS)

    Chari, D.R.; Rec, J.R.; Simoni, L.P.; Eimar, R.L.; Sowers, G.W.

    1987-01-01

    Palo Verde Nuclear Generating Station (PVNGS) is a three unit site which illustrates System 80 nuclear steam supply system (NSSS) design. The System 80 NSSS is the Combustion Engineering (C-E) standard design rated at 3817 Mwth. PVNGS Units 1 and 2 achieved commercial operation on February 13, 1986 and September 22, 1986, respectively, while Unit 3 has a forecast date for commercial operation in the third quarter of 1987. The System 80 design incorporates a reactor power cutback system (RPCS) feature which reduces plant trips caused by two common initiating events: loss of load/turbine trip (LOL) and loss of one main feedwater pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety system

  19. Draft environmental statement related to the proposed Jamesport Nuclear Power Station, Units 1 and 2: (Docket Nos. STN-50-516 and STN-50-517)

    International Nuclear Information System (INIS)

    1975-02-01

    The proposed action is the issuance of construction permits to the Long Island Lighting Company for the construction of the Jamesport Nuclear Power Station, Units 1 and 2, located on Long Island Sound in the Town of Riverhead, New York. Operation of the proposed once-through cooling system will result in the heating of 4180 cfs of water by 18/degree/F. Phytoplankton, zooplankton, and ichthyoplankton will be entrained by the cooling system. The organisms in approximately 40 /times/ 10 9 ft 3 of water per year will suffer direct mortality due to chlorination and heat shock. The risk associated with accidental radiation exposure will be very low. A bottom area of some 105.5 acres will be affected by dredging and jetty construction. Of this total, 4.35 acres will be permanently replaced by jetties, and 101.15 acres will be temporarily (a total of four years) disrupted by dredging. This impact will be temporary, since the dredged areas will be recolonized by the benthic community upon completion of work. There is a potential for substantial impingement loss of fishes on the intake screens which will require additional data to quantify. Approximately 39 miles of transmission lines will be constructed. Ninety-two percent of the total routings utilize existing rights-of-way, railroad rights-of-way, or new rights-of-way immediately adjacent to railroad rights-of-way. The rights-of-way will require approximately 621 acres. 65 refs., 64 tabs

  20. Operation of Grand Gulf Nuclear Station, Units 1 and 2, Dockets Nos. 50-416 and 50-417: Mississippi Power and Light Company, Middle South Energy, Inc., South Mississippi Electric Power Association. Final environmental statement

    International Nuclear Information System (INIS)

    1981-09-01

    The information in this Final Environmental Statement is the second assessment of the environmental impacts associated with the construction and operation of the Grand Gulf Nuclear Station, Units 1 and 2, located on the Mississippi River in Claiborne County, Mississippi. The Draft Environmental Statement was issued in May 1981. The first assessment was the Final Environmental Statement related to construction, which was issued in August 1973 prior to issuance of the Grand Gulf Nuclear Station construction permits. In September 1981 Grand Gulf Unit 1 was 92% complete and Unit 2 was 22% complete. Fuel loading for Unit 1 is scheduled for December 1981. The present assessment is the result of the NRC staff review of the activities associated with the proposed operation of the Station, and includes the staff responses to comments on the Draft Environmental Statement

  1. AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro`s Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations` safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs.

  2. AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

    International Nuclear Information System (INIS)

    1995-06-01

    The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro's Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations' safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs

  3. 75 FR 52045 - Arizona Public Service Company, Palo Verde Nuclear Generating Station, Unit 3; Environmental...

    Science.gov (United States)

    2010-08-24

    ... Company, Palo Verde Nuclear Generating Station, Unit 3; Environmental Assessment and Finding of No.... NPF-74, issued to Arizona Public Service Company (APS, the licensee), for operation of Palo Verde... Statement for the Palo Verde Nuclear Generating Station, NUREG-0841, dated February 1982. Agencies and...

  4. US central station nuclear electric generating units: significant milestones

    International Nuclear Information System (INIS)

    1979-09-01

    Listings of US nuclear power plants include significant dates, reactor type, owners, and net generating capacity. Listings are made by state, region, and utility. Tabulations of status, schedules, and orders are also presented

  5. Final environmental statement for William B. McGuire Nuclear Station, Units 1 and 2: (Docket Nos. 50-369 and 50-370)

    International Nuclear Information System (INIS)

    1976-04-01

    The proposed action is the issuance of operating licenses to the Duke Power Company for the startup and operation of the William B. McGuire Nuclear Station, Units 1 and 2 (the plant) located on the Lake Norman in Mecklenburg County, 17 miles north-northwest of Charlotte, North Carolina. The units will be cooled by once-through flow of water from Lake Norman. Two units, each with a net electrical capacity of 1180 MWe will be added to the resources of the Duke Power Company. This will have a favorable effect on reserve margins and provide a cost savings of $77 to $122 million in production costs in 1979 if the units come on line as scheduled, and cost savings in subsequent years. Approximately 200 acres of rural, partially wooded land owned by the applicant will be unavailable for other uses during the 40-year life of the plant. Approximately 61.6 acres of additional land will be utilized for transmission line corridors and/or switchyard and maintained under controlled conditions. Land-use patterns will necessarily conform to the needs of the application but will not be changed significantly from present usage. At full power, condenser cooling water could be heated to as high as 96/degree/F (35.6/degree/C) as a monthly average and will be discharged at a rate of up to 4492 cfs. The temperature rise of the water will be 16/degree/F (8.8/degree/C) to 32/degree/F (17.8/degree/C) above ambient. The heated water will mix with the cooler water of Lake Norman, where the heat will be dissipated to the atmosphere. The increase in temperature will cause a loss of approximately 31 cfs of water as a result of increased evaporation. 26 figs., 46 tabs

  6. Generic implications of ATWS events at the Salem Nuclear Power Plant: generic implications. Vol. 1

    International Nuclear Information System (INIS)

    1983-04-01

    This report is the first of two volumes. It documents the work of an interoffice, interdisciplinary NRC Task Force established to determine the generic implications of two anticipated transients without scram (ATWS) at the Salem Nuclear Power Plant, Unit 1 on February 22 and 25, 1983. A second report will document the NRC actions to be taken based on the work of the Task Force. The Task Force was established to address three questions: (1) Is there a need for prompt action for similar equipment in other facilities. (2) Are NRC and its licensees learning the sefety-management lessons, and, (3) How should the priority and content of the ATWS rule be adjusted. A number of short-term actions were taken through Bulletins and an Information Notice. Intermediate-term actions to address the generic issues will be addressed in the separate report and implemented through appropriate regulatory mechanisms

  7. Thermal efficiency improvements - an imperative for nuclear generating stations

    International Nuclear Information System (INIS)

    Hassanien, S.; Rouse, S.

    1997-01-01

    A one and a half percent thermal performance improvement of Ontario Hydro's operating nuclear units (Bruce B, Pickering B, and Darlington) means almost 980 GWh are available to the transmission system (assuming an 80% capacity factor). This is equivalent to the energy consumption of 34,000 electrically-heated homes in Ontario, and worth more than $39 million in revenue to Ontario Hydro Nuclear Generation. Improving nuclear plant thermal efficiency improves profitability (more GWh per unit of fuel) and competitiveness (cost of unit energy), and reduces environmental impact (less spent fuel and nuclear waste). Thermal performance will naturally decrease due to the age of the units unless corrective action is taken. Most Ontario Hydro nuclear units are ten to twenty years old. Some common causes for loss of thermal efficiency are: fouling and tube plugging of steam generators, condensers, and heat exchangers; steam leaks in the condenser due to valve wear, steam trap and drain leaks; deposition, pitting, cracking, corrosion, etc., of turbine blades; inadequate feedwater metering resulting from corrosion and deposition. This paper stresses the importance of improving the nuclear units' thermal efficiency. Ontario Hydro Nuclear has demonstrated energy savings results are achievable and affordable. Between 1994 and 1996, Nuclear reduced its energy use and improved thermal efficiency by over 430,000 MWh. Efficiency improvement is not automatic - strategies are needed to be effective. This paper suggests practical strategies to systematically improve thermal efficiency. (author)

  8. Safety Evaluation Report related to the restart of Davis-Besse Nuclear Power Station, Unit 1, following the event of June 9, 1985 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    1986-06-01

    On June 9, 1985, the Davis-Besse Nuclear Power Station, operated by the Toledo Edison Company, experienced a partial loss of main feedwater while the plant was at 90% power. The ensuing reactor trip was followed by spurious isolation of the steam geneators which initiated a chain of events involving a number of equipment malfunctions and several operator errors ultimately interrupting all feedwater for a short period of time. By the time operators were able to restore feedwater, both steam generators had dried out. A letter from the Director of the Office of Nuclear Reactor Regulation, pursuant to 10 CFR 50.54(f) of the Commission's regulations, confirmed that the Davis-Besse facility would not be restarted without NRC approval. The letter also requested that Toledo Edison submit its program for resolving numerous concerns identified by the staff. In response, the license submitted the Davis-Besse Course of Action report. The staff has reviewed that document and other supporting material submitted by the licensee; the staff's evaluation of that information is presented in this report

  9. Science, society, and America's nuclear waste: Unit 1, Nuclear waste

    International Nuclear Information System (INIS)

    1992-01-01

    This is unit 1 in a four-unit secondary curriculum. It is intended to provide information about scientific and societal issues related to the management of spent nuclear fuel from generation of electricity at nuclear powerplants and high-level radioactive waste from US national defense activities. The curriculum, supporting classroom activities, and teaching materials present a brief discussion of energy and electricity generation, including that produced at nuclear powerplants; information on sources, amounts, location, and characteristics of spent nuclear fuel and high-level radioactive waste; sources, types and effects of radiation; US policy for managing and disposing of spent nuclear fuel and high-level radioactive waste and what other countries are doing; and the components of the nuclear waste management system

  10. Socio-economic impacts of nuclear generating stations

    International Nuclear Information System (INIS)

    Weisiger, M.L.; Pijawka, K.D.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the St. Lucie nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  11. Technical evaluation report on the adequacy of station electric distribution system voltages for the Point Beach Nuclear Plant, Units 1 and 2. (Docket Nos. 50-266, 50-301)

    International Nuclear Information System (INIS)

    White, R.L.

    1983-01-01

    This report documents the technical evaluation of the adequacy of the station electric distribution system voltages for the Point Beach Nuclear Plant, Units 1 and 2. The evaluation is to determine if the onsite distribution system, in conjunction with the offsite power sources, has sufficient capacity to automatically start and operate all Class 1E loads within the equipment voltage ratings under certain conditions established by the Nuclear Regulatory Commission. For the worst case conditions study submitted by the licensee, it was shown that the station electric distribution system voltages would be adequate to start and operate 4160-volt and 480-volt Class 1E loads and their associated low voltage controls

  12. Three Mile Island Nuclear Station steam generator chemical cleaning

    International Nuclear Information System (INIS)

    Hansen, C.A.

    1992-01-01

    The Three Mile Island-1 steam generators were chemically cleaned in 1991 by the B and W Nuclear Service Co. (BWNS). This secondary side cleaning was accomplished through application of the EPRI/SGOG (Electric Power Research Institute - Steam Generator Owners Group) chemical cleaning iron removal process, followed by sludge lancing. BWNS also performed on-line corrosion monitoring. Corrosion of key steam generator materials was low, and well within established limits. Liquid waste, subsequently processed by BWNS was less than expected. 7 tabs

  13. Snubber reduction program at the Byron Station, Unit 1

    International Nuclear Information System (INIS)

    Arterburn, J.; Bakhtiari, S.

    1987-01-01

    Commonwealth Edison Company's (CECo's) Byron Station, unit 1, was originally designed with approximately 1200 snubbers supporting the plant's large- and small-bore piping systems. This relatively large number of snubbers is attributed to excessive conservatism in nuclear piping codes and regulations effective during the original piping design. A recent pilot program at CECo's LaSalle County Station, a boiling water reactor plant, demonstrated that a 50% or greater reduction in total snubber population is achievable in plants of this design vintage. Based on the successful results of the pilot program, CECo initiated a full scale snubber reduction program at Byron, a pressurized water reactor plant of the same vintage at the LaSalle County Station. The benefits from a reduced snubber population are described. To realize the maximum potential benefits, all snubbers in the plant were prioritized in order of desirability for removal. The priority designations are discussed. The major results from phase 1 of the Byron program are summarized. The NRC inspection of the project addressed a variety of issues and is discussed. The conclusions that can be drawn from the phase 1 program are summarized

  14. Reliable, fault tolerant control systems for nuclear generating stations

    International Nuclear Information System (INIS)

    McNeil, T.O.; Olmstead, R.A.; Schafer, S.

    1990-01-01

    Two operational features of CANDU Nuclear Power Stations provide for high plant availability. First, the plant re-fuels on-line, thereby eliminating the need for periodic and lengthy refuelling 'outages'. Second, the all plants are controlled by real-time computer systems. Later plants are also protected using real-time computer systems. In the past twenty years, the control systems now operating in 21 plants have achieved an availability of 99.8%, making significant contributions to high CANDU plant capacity factors. This paper describes some of the features that ensure the high degree of system fault tolerance and hence high plant availability. The emphasis will be placed on the fault tolerant features of the computer systems included in the latest reactor design - the CANDU 3 (450MWe). (author)

  15. Nuclear generating station and heavy water plant cost estimates for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-07-01

    Nuclear generating station capital, operating and maintenance costs are basic input data for strategy analyses of alternate nuclear fuel cycles. This report presents estimates of these costs for natural uranium CANDU stations, CANDU stations operating on advanced fuel cycles, and liquid metal fast breeder reactors. Cost estimates for heavy water plants are also presented. The results show that station capital costs for advanced fuel cycles are not expected to be significantly greater than those for natural uranium stations. LMFBR capital costs are expected to be 25-30 percent greater than for CANDU's. (auth)

  16. Heat exchanger tubing materials for CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Taylor, G.F.

    1977-07-01

    The performance of steam generator tubing (nickel-chromium-iron alloy in NPD and nickel-copper alloy in Douglas Point and Pickering generating stations) has been outstanding and no corrosion-induced failures have occurred. The primary coolant will be allowed to boil in the 600 MW (electrical) CANDU-PHW reactors. An iron-nickel-chromium alloy has been selected for the steam generator tubing because it will result in lower radiation fields than the alloys used before. It is also more resistant than nickel-chromium-iron alloy to stress corrosion cracking in the high purity water of the primary circuit, an unlikely but conceivable hazard associated with higher operating temperatures. Austenitic alloy and ferritic-austenitic stainless steel tubing have been selected for the moderator coolers in CANDU reactors being designed and under construction. These materials will reduce the radiation fields around the moderator circuit while retaining the good resistance to corrosion in service water that has characterized the copper-nickel alloys now in use. Brass and bronze tubes in feedwater heaters and condensers have given satisfactory service but do, however, complicate corrosion control in the steam cycle and, to reduce the transport of corrosion products from the feedtrain to the steam generator, stainless steel is preferred for feedwater heaters and stainlss steel or titanium for condensers. (author)

  17. Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

    International Nuclear Information System (INIS)

    1987-04-01

    The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

  18. Safety evaluation report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2. Dockets Nos. 50-416 and 50-417, Mississippi Power and Light Company; Middle South Energy, Inc., South Mississippi Electric Power Association

    International Nuclear Information System (INIS)

    1982-06-01

    Supplement 2 to the Safety Evaluation Report for Mississippi Power and Light Company, et. al, joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson, in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  19. Technical evaluation of the proposed design modifications and technical specification changes on grid voltage degradation (Part A) for the Pilgrim Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    White, R.L.

    1980-01-01

    This report documents the technical evaluation of the proposed design modifications and Technical Specification changes for protection of Class 1E equipment from grid voltage degradation for the Pilgrim Nuclear Power Station. The review criteria are based on several IEEE standards and the Code of Federal Regulations. The evaluation compares the submittals made by the licensee with the NRC staff positions and the review criteria and presents the reviewer's conclusion on the acceptability of the proposed system

  20. Duke Power Company - McGuire Nuclear Station: steam-generator hideout return and cleanup

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    McGuire Nuclear Station steam generator hideout return and cleanup are discussed. Hideout return data are presented for Unit 1 shutdowns that occurred on November 23, 1984, and April 19, 1985, and a Unit 2 shutdown on January 25, 1985. The data are presented as the concentrations of various species as a function of time after power reduction and primary water temperature. The steam generator blowdown as a function of time after power reduction is also presented. The concentrations of sodium, potassium, calcium, magnesium, aluminum, iron, and copper cations, and chloride, fluoride, sulfate, phosphate and nitrite anions were monitored during the each shutdown. Silica was also measured in the two 1985 shutdowns. The return of sulfate, phosphate, calcium and magnesium showed retrograde solubility. Silica concentrations showed an increase as the temperature decreased to about 450 to 500 0 F and then they decreased as the temperature decreased. McGuire has a holf point at 300 at 350 0 F to clean up the steam generator secondary water. The return of sulfates should occur within 4 to 6 hours. The blowdown is maximized to reduce the secondary water impurity concentrations. Cleanup continues until the sulfate concentration is reduced to below 100 ppb. At that point cooldown is continued

  1. Progress on the decommissioning of Zion nuclear generating station

    International Nuclear Information System (INIS)

    Moloney, B. P.; Hess, J.

    2013-01-01

    The decommissioning of the twin 1040 MWe PWRs at Zion, near Chicago USA is a ground breaking programme. The original owner, Exelon Nuclear Corporation, transferred the full responsibility for reactor dismantling and site license termination to a subsidiary of EnergySolutions. The target end state of the Zion site for return to Exelon will be a green field with the exception of the dry fuel storage pad. In return, ZionSolutions has access to the full value of the decommissioning trust fund. There are two potential attractions of this model: lower overall cost and significant schedule acceleration. The Zion programme which commenced in September 2010 is designed to return the cleared site with an Independent Spent Fuel Storage Installation (ISFSI) pad in 2020, 12 years earlier than planned by Exelon. The overall cost, at $500 M per full size power reactor is significantly below the long run trend of $750 M+ per PWR. Implementation of the accelerated programme has been underway for nearly three years and is making good progress. The programme is characterised by numerous projects proceeding in parallel. The critical path is defined by the inspection and removal of fuel from the pond and transfer into dry fuel storage casks on the ISFSI pad and completion of RPV segmentation. Fuel loading is expected to commence in mid- 2013 with completion in late 2014. In parallel, ZionSolutions is proceeding with the segmentation of the Reactor Vessel (RV) and internals in both Units. Removal of large components from Unit 1 is underway. Numerous other projects are underway or have been completed to date. They include access openings into both containments, installation of heavy lift crane capacity, rail upgrades to support waste removal from the site, radiological characterization of facilities and equipment and numerous related tasks. As at February 2013, the programme is just ahead of schedule and within the latest budget. The paper will provide a fuller update. The first two

  2. Progress on the decommissioning of Zion nuclear generating station

    Energy Technology Data Exchange (ETDEWEB)

    Moloney, B. P. [EnergySolutions EU Ltd, Swindon, Wiltshire (United Kingdom); Hess, J. [EnergySolutions LLC, Oak Ridge, TN (United States)

    2013-07-01

    The decommissioning of the twin 1040 MWe PWRs at Zion, near Chicago USA is a ground breaking programme. The original owner, Exelon Nuclear Corporation, transferred the full responsibility for reactor dismantling and site license termination to a subsidiary of EnergySolutions. The target end state of the Zion site for return to Exelon will be a green field with the exception of the dry fuel storage pad. In return, ZionSolutions has access to the full value of the decommissioning trust fund. There are two potential attractions of this model: lower overall cost and significant schedule acceleration. The Zion programme which commenced in September 2010 is designed to return the cleared site with an Independent Spent Fuel Storage Installation (ISFSI) pad in 2020, 12 years earlier than planned by Exelon. The overall cost, at $500 M per full size power reactor is significantly below the long run trend of $750 M+ per PWR. Implementation of the accelerated programme has been underway for nearly three years and is making good progress. The programme is characterised by numerous projects proceeding in parallel. The critical path is defined by the inspection and removal of fuel from the pond and transfer into dry fuel storage casks on the ISFSI pad and completion of RPV segmentation. Fuel loading is expected to commence in mid- 2013 with completion in late 2014. In parallel, ZionSolutions is proceeding with the segmentation of the Reactor Vessel (RV) and internals in both Units. Removal of large components from Unit 1 is underway. Numerous other projects are underway or have been completed to date. They include access openings into both containments, installation of heavy lift crane capacity, rail upgrades to support waste removal from the site, radiological characterization of facilities and equipment and numerous related tasks. As at February 2013, the programme is just ahead of schedule and within the latest budget. The paper will provide a fuller update. The first two

  3. Main feedwater valve diagnostics at Waterford 3 nuclear generating station

    International Nuclear Information System (INIS)

    Fitzgerald, W.V.

    1991-01-01

    Pneumatically-operated control valves are coming under increasing scrutiny in nuclear power plants because of their relatively high incident rate. The theory behind a device that could make performance evaluation of these valves simpler and more effective was first described at the original EPRI Power Plant Valve Symposium. The development of this Diagnostic System was completed in 1989, and it was recently used to troubleshoot two main feedwater valves at Louisiana Power and Light's Waterford 3 Power Station. During a cold snap last December, these valves failed to respond to the input signal and, as a result, the plant came off line. An incident report had to be filed, and the plant chose to contact the original equipment manufacturer (OEM) for assistance. This paper describes the original incident involving these valves and then gives a brief description of the diagnostic system and how it works. The balance of the paper then reviews how the OEM and plant personnel utilized the system to evaluate each component of the control valve assembly (I/P transducer, positioner, volume boosters, actuator, and valve body assembly). By simply stroking the valve and monitoring pneumatic signals and valve position, the problem was traced to a malfunctioning positioner and a volume booster that was leaking. The problems were corrected and new performance signatures run for the valves using the system to document their improved operation. This case study demonstrates how new Diagnostic Technology along with OEM involvement can effectively address problems with pneumatically-operated control valves so that root-cause solutions can be implemented

  4. Demonstration of containment purge and vent valve operability for the Hope Creek Generating Station, Unit 1 (Docket No. 50-354)

    International Nuclear Information System (INIS)

    Kido, C.

    1985-05-01

    The containment purge and vent valve qualification program for the Hope Creek Generating Station has been reviewed by the NRC Licensing Support Section. The review indicates that the licensee has demonstrated the dependability of containment isolation against the buildup of containment pressure due to a LOCA/DBA with the restrictions that during operating conditions 1, 2, and 3 all purge and vent valves will be sealed closed and under administrative control, and during power ascension and descension conditions the 26 in. inboard valve (1-GS-HV-4952) will be used in series with the 2 in. bypass valve (1-GS-HV-4951) to control the release of containment pressure

  5. Electricity Generation Through the Koeberg Nuclear Power Station of Eskom in South Africa

    International Nuclear Information System (INIS)

    Dladla, G.; Joubert, J.

    2015-01-01

    The poster provides information on the process of nuclear energy generation in a nuclear power plant in order to produce electricity. Nuclear energy currently provides approximately 11% of the world’s electricity needs, with Koeberg Nuclear Power Station situated in the Western Cape providing 4.4% of South Africa’s electricity needs. As Africa’s first nuclear power station, Koeberg has an installed capacity of 1910 MW of power. Koeberg’ s total net output is 1860 MW. While there are significant differences, there are many similarities between nuclear power plants and other electrical generating facilities. Uranium is used for fuel in nuclear power plants to make electricity. With the exception of solar, wind, and hydroelectric plants, all others including nuclear plants convert water to steam that spins the propeller-like blades of a turbine that spins the shaft of a generator. Inside the generator coils of wire and magnetic fields interact to create electricity. The energy needed to boil water into steam is produced in one of two ways: by burning coal, oil, or gas (fossil fuels) in a furnace or by splitting certain atoms of uranium in a nuclear energy plant. The uranium fuel generates heat through a controlled fission process fission, which is described in this poster presentation. The Koeberg Nuclear Power Station is a Pressurised water reactor (PWR). The operating method and the components of the Koeberg Power Station are also described. The nuclear waste generated at a nuclear power station is described under three headings— low-level waste, intermediate-level waste and used or spent fuel, which can be solid, liquid or gaseous. (author)

  6. Aerial radiological survey of the Rancho Seco Nuclear Generating Station, Clay Station, California, 18 January 1980 to 1 February 1980

    International Nuclear Information System (INIS)

    1980-11-01

    An airborne radiological survey of 260 km 2 area centered over the Rancho Seco Nuclear Generating Station was made 18 January through 1 February 1980. Detected radioisotopes and their associated gamma ray exposure rates were consistent with that expected from normal background emitters, except directly over the station. Count rates observed at 90 m altitude were converted to exposure rates at 1 m above the ground and are presented in the form of an isopleth map

  7. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1993-07-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  8. Steam generator management at Ontario Hydro Nuclear Stations

    International Nuclear Information System (INIS)

    Nickerson, J.; Maruska, C.C.

    1998-01-01

    Managing ageing steam generators involves costly decisions for the utility, both in terms of the cost of the maintenance activities and in terms of having the unit shutdown and consequent power loss while performing these activities. The benefits of these activities are seldom guaranteed and are sometimes very intangible. For nuclear utilities the most pertinent questions that arise are have we identified all the problem(s), can we predict the risk due to these problems? Can we implement corrective and preventive activities to manage the problem and what is the optimum timing of implementation? Is the money spent worthwhile, i.e. has it given us a return in production and safety? Can we avoid surprises? How can we tangibly measure success? This paper touches briefly on all the questions mentioned above but it mainly addresses the last question: 'how can we tangibly measure success?' by using several success indicators proposed by EPRI and by applying them to actual Ontario Hydro experience. The appropriateness of these success indicators as the means to assess the success of these programs, to feed back the results, and to enhance or revise the programs will be discussed. (author)

  9. Steam generator management at Ontario Hydro Nuclear Stations

    Energy Technology Data Exchange (ETDEWEB)

    Nickerson, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Maruska, C.C. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-07-01

    Managing ageing steam generators involves costly decisions for the utility, both in terms of the cost of the maintenance activities andin terms of having the unit shutdown and consequent power loss while performing these activities. The benefits of these activities are seldom guaranteed and are sometimes very intangible. For nuclear utilities the most pertinent questions that arise are have we identified all the problem(s), can we predict the risk due to these problems? Can we implement corrective and preventive activities to manage the problem and what is the optimum timing of implementation? Is the money spent worthwhile, i.e. has it given us a return in production and safety? Can we avoid surprises? How can we tangibly measure success? This paper touches briefly on all the questions mentioned above but it mainly addresses the last question: 'how can we tangibly measure success?' by using several success indicators proposed by EPRI and by applying them to actual Ontario Hydro experience. The appropriateness of these success indicators as the means to assess the success of these programs, to feed back the results, and to enhance or revise the programs will be discussed. (author)

  10. Technical specifications: Seabrook Station, Unit 1 (Docket No. 50-443)

    International Nuclear Information System (INIS)

    1990-03-01

    The Seabrook Station, Unit 1 Technical Specifications were prepared by the US Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  11. Commercial grade item (CGI) dedication of generators for nuclear safety related applications

    International Nuclear Information System (INIS)

    Das, R.K.; Hajos, L.G.

    1993-01-01

    The number of nuclear safety related equipment suppliers and the availability of spare and replacement parts designed specifically for nuclear safety related application are shrinking rapidly. These have made it necessary for utilities to apply commercial grade spare and replacement parts in nuclear safety related applications after implementing proper acceptance and dedication process to verify that such items conform with the requirements of their use in nuclear safety related application. The general guidelines for the commercial grade item (CGI) acceptance and dedication are provided in US Nuclear Regulatory Commission (NRC) Generic Letters and Electric Power Research Institute (EPRI) Report NP-5652, Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications. This paper presents an application of these generic guidelines for procurement, acceptance, and dedication of a commercial grade generator for use as a standby generator at Salem Generating Station Units 1 and 2. The paper identifies the critical characteristics of the generator which once verified, will provide reasonable assurance that the generator will perform its intended safety function. The paper also delineates the method of verification of the critical characteristics through tests and provide acceptance criteria for the test results. The methodology presented in this paper may be used as specific guidelines for reliable and cost effective procurement and dedication of commercial grade generators for use as standby generators at nuclear power plants

  12. AECB staff annual assessment of the Darlington Nuclear Generating Station for the year 1996

    International Nuclear Information System (INIS)

    1997-06-01

    The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of safety at the Darlington Nuclear Generating Station for 1996. Ontario Hydro operated the station in a safe manner in 1996. All four special safety systems were fully available 100 percent of the time. There were more problems that affected the safety support systems in 1996 than in the previous year

  13. Determination of reliability criteria for standby diesel generators at a nuclear power station

    International Nuclear Information System (INIS)

    Evans, M.G.K.

    1987-01-01

    The requirement for standby diesel generators at nuclear power stations is developed and a probabilistic approach used to define the reliability parameters. The present criteria used when ordering a diesel generator are compared with the testing required by the regulatory body and the most likely requirement following an accident. The impact of this on the diesels at a particular station and the root cause of failures are discussed. (orig.)

  14. Argentine activities on fuels for nuclear generation stations

    International Nuclear Information System (INIS)

    Olezza, R.L.; Valesi, J.

    1995-01-01

    In the last six years, significant changes have taken place in the nuclear fuel activity field in Argentina, therefore all the areas of the nuclear fuel cycle have been strongly influenced by these. The strategies carried out by CNEA to give an initial answer to the modifications of the domestic and international context of the nuclear fuel cycle were described in the previous Conference. Three years later, it is possible to appreciate the first results of the application of those strategies, and also that the frame has continued not only evolving and requiring new answers, but adapting and accentuating some strategies as well. A brief review of those results is presented here, together with a summary of the condition of the current situation and of the proposals to face it. (author)

  15. AECB staff annual assessment of the Bruce B Nuclear Generating Station for the year 1996

    International Nuclear Information System (INIS)

    1997-06-01

    The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Bruce Nuclear Generating Station B for 1996. It was concluded that Ontario Hydro operated Bruce B safely in 1996. Although the Bruce B plant is safe,it was noted that the number of outages and the number of secondary and tertiary equipment failures during reactor unit upsets increased. Ontario Hydro needs to pay special attention to prevent such a decrease in the safety performance at Bruce B

  16. AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1996

    International Nuclear Information System (INIS)

    1997-06-01

    The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Bruce Nuclear Generating Station A for 1996. Ontario Hydro operated Bruce A safely in 1996, maintaining the risk to workers and the public at an acceptably low level. Special safety system performance at Bruce A was adequate. Availability targets were all met. Improvement is needed to reduce the number of operating licence non-compliances

  17. Spent fuel storage at the Rancho Seco Nuclear Generation Station

    International Nuclear Information System (INIS)

    Miller, K.R.; Field, J.J.

    1995-01-01

    The Sacramento Municipal Utility District (SMUD) has developed a strategy for the storage and transport of spent nuclear fuel and is now in the process of licensing and manufacturing a Transportable Storage System (TSS). Staff has also engaged in impact limiter testing, non-fuel bearing component reinsertion, storage and disposal of GTCC waste, and site specific upgrades in support of spent fuel dry storage

  18. Nuclear power generating station equipment qualification method and apparatus

    International Nuclear Information System (INIS)

    Fero, A.H.; Potochnik, L.M.; Riling, R.W.; Semethy, K.F.

    1990-01-01

    This patent describes a method of monitoring an object piece of qualified equipment in a nuclear power plant. It comprises providing a first passive mimic means for mimicking the effect of radiation received by the object piece on the object piece; providing a second mimic means for mimicking the effect of a thermal history of the object piece on the object piece and mounting the first passive mimic means and the second mimic means in close proximity to the object piece

  19. Review of 12-hour shifts at nuclear generating stations

    International Nuclear Information System (INIS)

    Smiley, A.; Moray, N.P.

    1989-04-01

    This project reviewed the practice of 12-hour shift work schedules at nuclear power plants, and its relationship to safety. The current literature was examined for information on accidents, fatigue and personal preferences. Interviews with operators and maintainers showed that these groups had attitude and preference differences related to both 12 hour shift schedules and overtime work opportunities. Several factors related to 12-hour schedules were identified which could affect safety, but which have not been adequately considered. (24 refs.)

  20. Palo Verde nuclear generating station EASEplus SIMULATE model

    International Nuclear Information System (INIS)

    McDonald, W.F.; Reed, M.L.; Fauste, J.L.

    1992-01-01

    The Palo Verde on-site reactor engineers have an extremely powerful and accurate tool for quickly predicting the effects of reactor power maneuvers on core axial shape index (ASI) and xenon worth. They can analyze postulated future power maneuvers quickly and supply the reactor operators with valuable predictions without having to consult with the off-site nuclear analysis group. The tool developed by the nuclear analysis group was an advanced nodal code with a graphic user interface (GUI) driver for ease of use. The advanced nodal code used was the Studsvik of America SIMULATE-3 Version 2.20-DSI. This SIMULATE version was compiled for use on a personal computer (PC) with a Definicon Systems' 50-MHz coprocessor board. The GUI face used was Expert-EASE Systems' EASE+SIM3 Version 3.0 pre-/postprocessor. The system was installed on Compaq Deskpro 386/20e PCs located in the control room of each of the three units, in the reactor engineering office, in the nuclear analysis office, and in the control room of the training simulator

  1. Strategy for success in maintenance management at Point Lepreau nuclear generating station

    International Nuclear Information System (INIS)

    White, R.M.

    1987-01-01

    Improvements in availability of the station and in productivity of workers were achieved at Point Lepreau Nuclear Generating Station through implementation of a Maintenance Management System, which incorporates work planning and outage management techniques. Eight software systems on a VAX 11/70 minicomputer control work orders, temporary and permanent design changes, parts and material inventories, time keeping, and project management. All maintenance is coordinated through a regular planning meeting

  2. Criteria for the design of the control room complex for a nuclear power generating station

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This Standard addresses the central control room of a nuclear power generating station and the overall complex in which this room is housed. It is not intended to cover special or normally unattended control rooms, such as those provided for radioactive waste handling or for emergency shutdown operations. The nuclear power generating station control room complex provides a protective envelope for plant operating personnel and for instrument and control equipment vital to the operation of the plant during normal and abnormal conditions. In this capacity, the control room complex must be designed and constructed to meet the following criteria contained in Appendix A of 10CFR50, General Design criteria for Nuclear Power Plants: (1) Criterion 2: design bases for protection against natural phenomena; (2) Criterion 3: fire protection; (3) Criterion 4: environmental and missile design bases; (4) Criterion 5: sharing of structures, systems and components (multiunit stations only); and (5) Criterion 19: control room

  3. Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414, Duke Power Company, et al

    International Nuclear Information System (INIS)

    1983-02-01

    The Safety Evaluation Report for the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Electric Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  4. Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

    International Nuclear Information System (INIS)

    Yu, A.S.; Kukreti, B.M.

    1987-01-01

    The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

  5. Improvements to feed water system of vapor generators of nuclear power stations

    International Nuclear Information System (INIS)

    Byerlex, W.M.

    1976-01-01

    The description is given of a feed water system related to the steam generators for nuclear power stations and which have a water feed ring around their upper part. This water intake system enables water hammer to be avoided even during operation under low load [fr

  6. Manufacture of the 300 MW steam generator and pressure stabilizer for Qinshan Nuclear Power Station

    International Nuclear Information System (INIS)

    Qian Yi; Miao Deming.

    1989-01-01

    A brief description of the manufacturing process of the steam generator and pressure stabilizer for 300 MWe Qinshan Nuclear Power Station in Shanghai Boiler Works is presented, with special emphasis on fabrication facilities, test procedures and technological evaluations during the manufaturing process-imcluding deep driling of tubesheets, welding of tubes to tube-sheets and tube rolling tests

  7. Radioactive release data from Canadian nuclear generating stations 1872-1987

    International Nuclear Information System (INIS)

    1989-03-01

    All nuclear generating stations emit small quantities of radioactive effluent both into the atmosphere and in the form of liquid effluent, into the adjoining water body, be it river, lake or sea. The purpose of this document is to report on the magnitude of these emissions for each nuclear generating station in Canada and to indicate how these emissions compare with the relevant limitations imposed by the Atomic Energy Control Board as part of its regulatory and licensing program. This report incorporates histograms indicating the annual releases of tritium in air, noble gases, iodine-131, airborne particulates, tritium in water and waterborne gross beta activity for each nuclear generating station. In addition, for Pickering NGS 'A', annual released of carbon-14 are depicted for the years 1986 and 1987. In each case the emission data are compared to the Derived Emission Limit (DEL) in order that the data may be placed in perspective. At present, only Pickering NGS 'A' is required to monitor and report carbon-14 emissions. Environmental monitoring for C-14 is conducted around the Bruce site to determine the environmental impact of its emission and whether effluent monitoring will be necessary in future years. Three nuclear generating stations have been permanently taken out of service during the last few years (Gentilly NGS-1, Douglas Point NGS and NPD NGS). Some small emissions from these sites do still occur, however, due to decontamination and decommissioning operations. (11 tabs., 26 figs.)

  8. Lessons learned from the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1

    International Nuclear Information System (INIS)

    Russell, M.J.; Shieh, L.C.; Tsai, N.C.; Cheng, T.M.

    1987-01-01

    SCE developed site specific ground response spectra based on the Housner response spectra. These were modified as a result of NRC review to account for uncertainties in a limited frequency range, and were applied in a conventional manner to the development of time histories for structural analysis. Frequency-dependent soil structure interaction analyses were performed for the reactor building. Frequency-independent soil stiffness and damping values were used in conventional analyses of the turbine building. Conventional acceptance criteria were applied to concrete structures. Supplemental acceptance criteria were developed for the steel structures; these criteria allowed a limited amount of ductility in isolated members and were applied only if the connections were maintained in the elastic range. Consideration of the effect of the ductility on the total structural response was required. Application of these criteria and methods required only a moderate level of reinforcement of steel structures in order to assure adequate structural integrity. The development, review and approval of supplemental criteria and methods was extented to the reevaluation of piping and equipment, with an emphasis on piping. (orig./HP)

  9. Steam turbine generators for Sizewell 'B' nuclear power station

    International Nuclear Information System (INIS)

    Hesketh, J.A.; Muscroft, J.

    1990-01-01

    The thermodynamic cycle of the modern 3000 r/min steam turbine as applied at Sizewell 'B' is presented. Review is made of the factors affecting thermal efficiency including the special nature of the wet steam cycle and the use of moisture separation and steam reheating. Consideration is given to the optimization of the machine and cycle parameters, including particular attention to reheating and to the provision of feedheating, in order to achieve a high overall level of performance. A modular design approach has made available a family of machines suitable for the output range 600-1300 MW. The constructional features of the 630 MW Sizewell 'B' turbine generators from this range are described in detail. The importance of service experience with wet steam turbines and its influence on the design of modern turbines for pressurized water reactor (PWR) applications is discussed. (author)

  10. Cernavoda NPP Unit 1 - a plant of several generations

    International Nuclear Information System (INIS)

    Rotaru, I.; Metes, M.; Anghelescu, M.S.

    2000-01-01

    Cernavoda NPP Unit 1, the first nuclear power unit in Romania, has a long and tormented history. It represents a rather unique case in Eastern Europe. The project started well before 1989 (the construction phase lasted 17 years and generations were involved in its completion), but it is effectively based on western technology (Candu). Meanwhile, the national nuclear program underwent many changes, affecting the lives and careers of Romanian nuclear professionals. Finally, on December 2 nd 1996, the unit began its c ommercial operation , being operated at its nominal power rating of 706 MW e . It now provides a reliable source of electricity for Romanian economy, supplying to the national grid about 10% of the country's average annual demand. The paper reflects some aspects related to the shift of generations during the project's development, including the present stage. The operational performances achieved 'in service' by Cernavoda NPP Unit 1 up to the end of 1999 , are also presented. Reference to the electrical energy production, performance indicators, production costs, nuclear safety, radiation protection, radioactive wastes, nuclear fuel consumption and nuclear fuel performances are included. Comparisons are performed with similar indicators reported by other worldwide nuclear power plants, in order to assess our results. (authors)

  11. AECB staff annual assessment of the Point Lepreau Nuclear Generating Station

    International Nuclear Information System (INIS)

    1997-06-01

    The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Point Lepreau Generating Station in 1996. Point Lepreau operated safely but the worsening trends in NB Power's safety performance leads to the conclusion that urgent action is required. NB Power is required to report formally to the AECB on progress with measures to improve safety management every six months. Further licensing action will be taken on NB Power if it fails to make the improvements

  12. A study of wet deposition of atmospheric tritium releases at the Ontario Power Generation, Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Crooks, G.; DeWilde, J.; Yu, L.

    2001-01-01

    The Ontario Power Generation,Pickering Nuclear Generating Station (PNGS) has been investigating deposition of atmospheric releases of tritium on their site. This study has included numerical dispersion modelling studies conducted over the past three years, as well as an ongoing field monitoring study. The following paper will present results of the field monitoring study and make comparisons to the numerical modelling. The results of this study could be of potential use to nuclear stations in quantifying tritium deposition in near field regions where building wake effects dominate pollutant dispersion

  13. Socio-economic impacts of nuclear generating stations: summary report on the NRC post-licensing studies

    International Nuclear Information System (INIS)

    Chalmers, J.; Pijawka, D.; Branch, K.; Bergmann, P.; Flynn, J.; Flynn, C.

    1982-07-01

    Information is presented concerning the conceptual framework for the assessment of socioeconomic impacts; methodology for the post-licensing case studies; socioeconomic changes due to the construction and operation of nuclear generating stations; public response to the construction and operation of nuclear generating stations; socioeconomic consequences of the accident at Three Mile Island; the significance of socioeconomic change due to the construction and operation of nuclear generating stations; findings of the post-licensing studies relative to the nuclear station impact literature; and implications of the findings for projective assessments and planning studies

  14. Start up and commercial operation of Laguna Verde nuclear power plant. Unit 1

    International Nuclear Information System (INIS)

    Torres Ramirez, J.F.

    1991-01-01

    Prior to start up of Laguna Verde nuclear power plant preoperational tests and start tests were performed and they are described in its more eminent aspects. In relation to commercial operation of nuclear station a series of indicator were set to which allow the measurement of performance in unit 1, in areas of plant efficiency and personal safety. Antecedents. Laguna Verde station is located in Alto Lucero municipality in Veracruz state, 70 kilometers north-northeast from port of Veracruz and a 290 kilometers east-northeast from Mexico city. The station consist of two units manufactured by General Electric, with a nuclear system of vapor supply also called boiling water (BWR/5), and with a system turbine-generator manufactured by Mitsubishi. Each unit has a nominal power of 1931 MWt and a level design power of 675 Mwe and a net power of 654 Electric Megawatts

  15. Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1982-01-01

    Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

  16. Technical Specifications, Comanche Peak Steam Electric Station, Unit 1 (Docket No. 50-445)

    International Nuclear Information System (INIS)

    1990-04-01

    The Technical Specifications for Comanche Peak Steam Electric Station, Unit 1 were prepared by the US Nuclear Regulatory Commission. They set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility, as set forth in Section 50.36 of Title 10 of the Code of Federal Regulations Part 50, for the protection of the health and safety of the public

  17. The competitive economics of a middle aged multi unit nuclear generating station

    International Nuclear Information System (INIS)

    Talbot, K.H.

    1994-01-01

    In 1992 Ontario Hydro's 15 year old 4 x 850 MWe Candu, Bruce A Nuclear Generating Station was predicted to need considerable capital investment to replace pressure tubes, steam generators and other prematurely ageing equipment in order to restore the station to high performance. Over the subsequent two years the station has undergone 2 major economic assessment studies which have confirmed the economic viability of continued operation of the plant. Declining demand for electricity in Ontario combined with a excess of generating capacity and a need to stabilise electricity rates have however forced significant operational cost reductions and reduced capital availability for rehabilitation work, it's medium and long term future remains in question. This presentation offers a practical illustration of the need to maintain steady high performance from nuclear generating plant via the appropriate life management techniques. The avoidance of mid life infusion of capital is considered as essential if nuclear generation is to successfully survive major changes in economic conditions. 2 tabs., 7 figs

  18. Design and installation of a strategically placed algae mesh barrier at OPG Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Marttila, D.; Patrick, P.; Gregoris, C.

    2009-01-01

    Ontario Power Generation's Pickering Nuclear has experienced a number of events in which attached algae have become entrained in the water intake costing approximately $30M over the 1995-2005 period as a result of deratings, Unit shutdowns and other operational issues. In 2005-2006 OPG and Kinectrics worked collaboratively on evaluating different potential solutions to reduce the impact of algae on the station. One of the solutions developed by Kinectrics included a strategically placed barrier net designed to regulate algae flow into the station intake. In 2006, Kinectrics designed and installed the system, the first of its kind at a Nuclear Power Plant in Canada. The system was operational by May 2007. OPG completed an effectiveness study in 2007 and concluded the barrier system had a beneficial effect on reducing algae impact on the station. (author)

  19. AECB staff annual assessment of the Darlington Nuclear Generating Station for the year 1995

    International Nuclear Information System (INIS)

    1996-06-01

    This report is the Atomic ENergy Control Board staff assessment of safety at the Darlington Nuclear Generating Station for 1995. The report is based on observations made by our staff, and on information submitted to us by Ontario Hydro. Performance was satisfactory for all four special safety systems. In 1995, Ontario Hydro complied with the regulations made under the Atomic Energy Control Act, except for two instances of non-compliance with the Transport Packaging of Radioactive Materials Regulations. Radiation doses received by Ontario Hydro station staff were below the regulatory limits. In general Ontario Hydro's maintenance program was found satisfactory. 9 tabs

  20. Electrical system design and reliability at Ontario Hydro nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Royce, C. J. [Ontario Hydro, 700 University Avenue, Toronto, Ontario M5G 1X6 (Canada)

    1986-02-15

    This paper provides an overview of design practice and the predicted and actual reliability of electrical station service Systems at Ontario Nuclear Generating Stations. Operational experience and licensing changes have indicated the desirability of improving reliability in certain instances. For example, the requirement to start large emergency coolant injection pumps resulted in the turbine generator units in a multi-unit station being used as a back-up power supply. Results of reliability analyses are discussed. To mitigate the effects of common mode events Ontario Hydro adopted a 'two group' approach to the design of safety related Systems. This 'two group' approach is reviewed and a single fully environmentally qualified standby power supply is proposed for future use. (author)

  1. Final environmental statement related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414, Duke Power Company, et al

    International Nuclear Information System (INIS)

    1983-01-01

    This Final Environmental Statement contains the second assessment of the environmental impact associated with the operation of the Catawba Nuclear Station, Units 1 and 2, pursuant to the National Environmental Policy Act of 1969 (NEPA) and 10 CFR 51, as amended, of the NRC regulations. This statement examines: the affected environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs. Land use and terrestrial and aquatic-ecological impacts will be small. Operational impacts to historic and archeological sites will be negligible. The effects of routine operations, energy transmission, and periodic maintenance of rights-of-way and transmission facilities should not jeopardize any populations of endangered or threatened species. No significant impacts are anticipated from normal operational releases of radioactivity. The risk associated with accidental radiation exposure is very low. The net socioeconomic effects of the project will be beneficial

  2. MINAC radiography performed on susquehanna Steam Electric Station Unit 1

    International Nuclear Information System (INIS)

    Bognet, J.C.

    1986-01-01

    Ten welds were volumetrically examined with a manual and automated ultrasonic (UT) system during a Susquehanna Steam Electric Station (SES) Unit 1 preservice inspection. The automated system had been recently developed and several problems were encountered in this first field application. The ten welds examined had a Sweepolet-to-Risor weld configuration, which further complicated the examination effort. This weld configuration has corrosion-resistant cladding applied to the outside and inside circumference and, as a result of an installation/removal/reinstallation sequence during plant construction, is often referred to as the double weld. After several attempts to obtain interpretable UT data failed (e.g., repeatable data), the examination effort was terminated. PP and L opted to pursue using the Miniature Linear Accelerator (MINAC) to perform radiographic examination. The results were referenced in the Susquehanna SES Unit 1 outage summary report and submitted to the NRC. The total effort was viewed as a complete success with no impact to the overall outage duration. All welds previously attempted by automated and manual UT were successfully examined using the MINAC

  3. Aerial radiological survey of the San Onofre Nuclear Generating Station and surrounding area, San Clemente, California

    International Nuclear Information System (INIS)

    Hilton, L.K.

    1980-12-01

    An airborne radiological survey of an 11 km 2 area surrounding the San Onofre Nuclear Generating Station was made 9 to 17 January 1980. Count rates observed at 60 m altitude were converted to exposure rates at 1 m above the ground and are presented in the form of an isopleth map. Detected radioisotopes and their associated gamma ray exposure rates were consistent with that expected from normal background emitters, except directly over the plant

  4. IEEE Standard for qualification of Class 1E lead storage batteries for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This document describes qualification methods for Class 1E lead storage batteries and racks to be used in nuclear power generating stations outside of primary containment. Qualification required in ANSI/IEEE Std 279-1979 and IEEE Std 308-1978, can be demonstrated by using the procedures provided in this Standard in accordance with IEEE Std 323-1974. Battery sizing, maintenance, capacity testing, installation, charging equipment and consideration of other types batteries are beyond the scope of this Standard

  5. IEEE standard for qualification of class 1E lead storage batteries for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    IEEE Std 323-1974, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, was developed to provide guidance for demonstrating and documenting the adequacy of electrical equipment used in all Class 1E and interface systems. This standard, IEEE Std 535-1979, was developed to provide specific methods and type test procedures for lead storage batteries in reference to IEEE Std 323-1974

  6. IEEE guide for general principles of reliability analysis of nuclear power generating station protection systems

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Presented is the Institute of Electrical and Electronics Engineers, Inc. (IEEE) guide for general principles of reliability analysis of nuclear power generating station protection systems. The document has been prepared to provide the basic principles needed to conduct a reliability analysis of protection systems. Included is information on qualitative and quantitative analysis, guides for failure data acquisition and use, and guide for establishment of intervals

  7. IEEE standard for qualifying class IE equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    The Institute of Electrical and Electrical Engineers, Inc. (IEEE) standards for electrical equipment (Class IE) for nuclear power generating stations are given. The standards are to provide guidance for demonstrating and documenting the adequacy of electric equipment used in all Class IE and interface systems. Representative in containment design basis event conditions for the principal reactor types are included in the appendixes for guidance in enviromental simulation

  8. Guide to the collection and presentation of electrical, electronic, and sensing component reliability data for nuclear-power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    This Guide is intended to establish a method of collecting and presenting reliability data for quantitative systematic reliability analysis in nuclear power generating stations, as outlined in IEEE Std 351-1975. Appendix D, which is not a part of IEEE Std 500-1977 but which comprises the bulk of this publication, presents tables of reliability data for nuclear power generating stations, intended for use of nuclear systems reliability analysts or design engineers

  9. Purification and solidification of reactor wastes at a Canadian nuclear generating station

    International Nuclear Information System (INIS)

    Buckley, L.P.; Burt, D.A.

    1981-01-01

    The study aimed at development and demonstration of volume reduction and solidification of CANDU reactor wastes has been underway at Chalk River Nuclear Laboratories in the Province of Ontario, Canada. The study comprises membrane separation processes, evaporator appraisal and immobilization of concentrated wastes in bitumen. This paper discusses the development work with a wiped-film evaporator and the successful completion of demonstration tests at Douglas Point Nuclear Generating Station. Heavy water from the moderator system was purified and wastes arising from pump bowl decontamination were immobilized in bitumen with the wiped-film evaporator that was used in the development tests at Chalk River

  10. Analysis of Ling'ao nuclear power station unit 1 exciter No.11 bearing white metal damage and its operating temperature abnormally high

    International Nuclear Information System (INIS)

    Jia Kaili

    2005-01-01

    On the base of analyzing the type of exciter No.11 bearing white metal damage, the root cause and its solution are found. No damage was found on bearing white metal in the later time. On the base of analyzing the structure of the generator end bracket, it is pointed out that when the generator frame is full of pressed gas, the end bracket will deform, that result in the load on No.11 bearing increase, as a result causes the bearing temperature high. A proposal to this problem is presented. (author)

  11. TMI-1 restart: an evaluation of the licensee's management integrity as it affects restart of Three Mile Island Nuclear Station (Unit 1 Docket 50-289). Supplement 5

    International Nuclear Information System (INIS)

    1984-07-01

    Supplement 5 to the Safety Evaluation Report (SER) on TMI-1 Restart documents the review by the Nuclear Regulatory Commission (NRC) staff of nine investigations conducted by the NRC Office of Investigations into matters identified as relevant and material to an evaluation of the licensee's management integrity. The staff has included, as part of its evaluation, materials from its review of the GPU v. B and W lawsuit record (NUREG-1020LD, GPU, v. B and W Lawsuit Review and Its Effect on TMI-1) as well as other relevant materials developed since the close of the record in the TMI-1 Restart proceeding. In developing its position on General Public Utilities Nuclear Corporation's character (i.e., management integrity), the staff evaluated matters that cast doubt on the licensee's character, individually and collectively; considered the remedial actions taken by the licensee; and balanced past improper conduct of the licensee against its subsequent record of remedial actions and performance and record of current senior management of the licensee. The staff concluded that, while the past improper conduct was grave, the remedial actions taken, the subsequent record of performance, and the record of current senior management support a finding that GPUN can and will operate TMI-1 without undue risk to the health and safety of the public

  12. Qualifying commercial grade instruments for use in nuclear power generating stations

    International Nuclear Information System (INIS)

    Lamothe, R.J.; Scally, C.R.

    1983-01-01

    Nuclear environmental qualification of instrumentation has been successfully accomplished by many commercial grade equipment manufacturers. This paper was prepared as a guide to those manufacturers who want some insight into a qualification program. The areas addressed are the regulations and documents, the qualification program, and a case history of a chart recorder qualifications. The principal standards relating to a nuclear qualification program are IEEE Std. 323-1974 IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, IEEE Std. 344-1975 IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations and 10CFR50.49. Previously NUREG 0588 Interim Staff Position on Environmental Qualification of Safety-Related Equipment. These define the intent and purpose of the qualification. The qualification program itself consists of several distinct parts which require explanation, including the determination of qualified life, choice of test samples, selection of appropriate acceptance criteria, aging program, radiation testing, seismic testing, abnormal environment tests and others. The case history illustrates the qualification program and the thought processes involved

  13. Some environmental effects of emissions from CANDU nuclear generating stations and heavy water plants

    International Nuclear Information System (INIS)

    Effer, W.R.

    Non-radioactive releases during normal operation of Ontario Hydro's nuclear generating stations and heavy water plants are summarized and related to existing regulations and guidelines. Low-grade heat in the circulating cooling water discharge is the most important of the non-radioactive effluents. Some of the hydrological, biological and water quality aspects of thermal discharges are discussed in relation to the operation of Ontario Hydro's thermal generating stations on the Great Lakes. Chemical releases to air or water include chlorine, hydrogen sulphide, water treatment plant effluents, oily waste water and sewage lagoon effluents. The significance of the first two of these releases to the environment is reviewed, particularly in relation to Great Lakes water quality and biological concerns. (author)

  14. Emergency operating instruction improvements at San Onofre Nuclear Generating Station Units 2 and 3

    International Nuclear Information System (INIS)

    Trillo, M.W.; Smith, B.H.

    1989-01-01

    In late 1987, San Onofre nuclear generating station (SONGS) began an extensive upgrade of the units 2 and 3 emergency operating instructions (EOIs). The original intent of this program was to incorporate revised generic guidance and to correct problems that were identified by operators. While this program was in progress, the US Nuclear Regulatory Commission (NRC) conducted a series of audits of emergency operating procedure (EOP) development and maintenance programs as 16 commercial nuclear facilities in the United States. These audits included four stations with Combustion Engineering-designed nuclear steam supply systems. (One of these audits included a review of preupgrade SONGS units 2 and 3 EOIs.) Significant industrywide comments resulted from these audits. The NRC has stated its intent to continue the review and audit of EOIs and the associated maintenance programs at all US commercial nuclear facilities. The units 2 and 3 EOI upgrade program developed procedural improvements and procedural program maintenance improvements that address many of the existing audit comments that have been received by the industry. Other resulting improvements may be useful in minimizing NRC comments in future such audits. Specific improvements are discussed. The upgrade program resulted in benefits that were not originally anticipated. The results of this program can be of significant use by other utilities in addressing the industrywide concerns that have been raised in recent NRC audits of EOP development and maintenance programs

  15. Analysis of steam generator tube sections removed from Gentilly-2 nuclear generating station

    International Nuclear Information System (INIS)

    Semmler, J.; Lockley, A.J.; Doyon, D.

    2010-01-01

    deposition rate in the steam generators from 2001 to 2009 were estimated and compared to the estimated values from 1983 (station commissioning) to 2001. This paper presents a summary of steam generator tube characterization results and describes how the data on the oxide deposition rate were correlated to the changes in station operating practices. These data demonstrate the benefits that resulted from recent changes to station chemistry and operational practices. (author)

  16. IEEE Std 649-1980: IEEE standard for qualifying Class 1E motor control centers for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard describes the basic principles, requirements, and methods for qualifying Class 1E motor control centers for outside containment applications in nuclear power generating stations. Qualification of motor control centers located inside containment in a nuclear power generating station is beyond the scope of this standard. The purpose of this standard is (1) to define specific qualification requirements for Class 1E motor control centers in accordance with the more general qualification requirements of IEEE Std 323-1974, IEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations; (2) to provide guidance in establishing a qualification program for demonstrating the design adequacy of Class 1E motor control centers in nuclear power generating station applications

  17. 78 FR 32278 - Vogtle Electric Generating Station, Units 3 and 4; Southern Nuclear Operating Company; Change to...

    Science.gov (United States)

    2013-05-29

    ... Generating Station, Units 3 and 4; Southern Nuclear Operating Company; Change to Information in Tier 1, Table... Nuclear Operating Company, Inc., and Georgia Power Company, Oglethorpe Power Corporation, Municipal... Table 3.3-1, ``Definition of Wall Thicknesses for Nuclear Island Buildings, Turbine Buildings, and Annex...

  18. Steam generator replacement in Bruce A Unit 1 and Unit 2

    International Nuclear Information System (INIS)

    Hart, R.S.

    2006-01-01

    The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Units 2 that will extend the in-service life of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from the bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

  19. Cost comparison of 4x500 MW coal-fuelled and 4x850 MW CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Costa, M.

    1981-01-01

    The lifetime costs for a 4x850 MW CANDU generating station are compared to those for 4x500 MW bituminous coal-fuelled generating stations. Two types of coal-fuelled stations are considered; one burning U.S. coal which includes flue gas desulfurization and one burning Western Canadian coal. Current estimates for the capital costs, operation and maintenance costs, fuel costs, decommissioning costs and irradiated fuel management costs are shown. The results show: (1) The accumulated discounted costs of nuclear generation, although initially higher, are lower than coal-fuelled generation after two or three years. (2) Fuel costs provide the major contribution to the total lifetime costs for coal-fuelled stations whereas capital costs are the major item for the nuclear station. (3) The break even lifetime capacity factor between nuclear and U.S. coal-fuelled generation is projected to be 5%; that for nuclear and Canadian coal-fuelled generation is projected to be 9%. (4) Large variations in the costs are required before the cost advantage of nuclear generation is lost. (5) Comparison with previous results shows that the nuclear alternative has a greater cost advantage in the current assessment. (6) The total unit energy cost remains approximately constant throughout the station life for nuclear generation while that for coal-fuelled generation increases significantly due to escalating fuel costs. The 1978 and 1979 actual total unit energy cost to the consumer for several Ontario Hydro stations are detailed, and projected total unit energy costs for several Ontario Hydro stations are shown in terms of escalated dollars and in 1980 constant dollars

  20. Safety research of insulating materials of cable for nuclear power generating station

    Science.gov (United States)

    Lee, C. K.; Choi, J. H.; Kong, Y. K.; Chang, H. S.

    1988-01-01

    The polymers PE, EPR, PVC, Neoprene, CSP, CLPE, EP and other similar substances are frequently used as insulation and protective covering for cables used in nuclear power generating stations. In order to test these materials for flame retardation, environmental resistance, and cable specifications, they were given the cable normal test, flame test, chemical tests, and subjected to design analysis and loss of coolant accident tests. Material was collected on spark tests and actual experience standards were established through these contributions and technology was accumulated.

  1. Quality management for design engineering for San Onofre nuclear generating station

    International Nuclear Information System (INIS)

    Thompson, P.C.; Baker, R.L.

    1991-01-01

    Quality management, as applied to design engineering for the San Onofre nuclear generating station, provides a systematic process for data collection and analysis of performance indicators for quality, cost, and delivery of design modifications for the three operating units. Southern California Edison (SCE) and Bechtel Power Corporation (BPC) have collaborated to establish a performance baseline from nearly 2 years of data. This paper discusses how the baseline was developed and how it can be used to predict and assess future performance. It further discusses new insights to the engineering process and opportunities for improvements that have been identified

  2. AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1995

    International Nuclear Information System (INIS)

    1996-06-01

    The Atomic Energy Control Board conducts a staff assessment of safety at Bruce Nuclear Generating Station A for 1995. On-site Project Officers and Ottawa based specialists monitored the station throughout the year. Ontario Hydro operated Bruce A safely in 1995, maintaining the risk to workers and the public at an acceptably low level. Radiation doses to workers and releases to the environment were well below regulatory limits. However, Ontario Hydro must improve contamination control at Bruce A. Special safety system performance a Bruce A was less than adequate. The negative pressure containment system and units 4's shutdown system two exceeded unavailability targets in 1995. However, we are satisfied Ontario Hydro is taking appropriate action to correct this. 5 tabs., 5 figs

  3. Processing of sump sludges at the Commonwealth Edison Byron Nuclear Generating Station

    International Nuclear Information System (INIS)

    Herrmann, D.; Gardner, D.A.; Taylor, E.R. Jr.

    1990-01-01

    A basic criterion for the disposal of radioactive waste by shallow land burial is that the material must not contain free liquids. In addition burial sites' requirements regarding radioactive waste containing oils, even though solidified, are restrictive. At Commonwealth Edison Byron Nuclear Generating Station a methodology for processing treated waste sludges, originating form the turbine building's floor drains was developed and implemented. As a result of this effort, 322 drums of oil and water sludge were processed. A dry cake, i.e., no free liquids, was produced, packaged, and readied for disposal. The dry cake contained less than 2% oil. The liquid phases resulting from the processing of the treated waste sludge were oil (that was to be processed for disposal as non-radioactive) and filtrate containing less than 5 ppm total suspended solids (TSS) and oil/grease. The filtrate TSS was below the Station's National Pollution Discharge Elimination System (NPDES) permit release limits. 4 figs

  4. Pre-license team training at San Onofre Nuclear Generating Station

    International Nuclear Information System (INIS)

    Freers, S.M.; Hyman, M.

    1987-01-01

    Team Training at San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 has been developed to enhance the performance of station operations personnel. The FACT Training Program (Formality, Attention to Detail, Consistency and Team Effort) is the common denominator for operations team training. Compliance with good operating practices is enhanced by operators working as a team toward the same goal, using the same language, practicing the same operating and communication skills, possessing a clear understanding of individual roles and responsibilities of team members and practicing attention to detail in every task. These elements of effective teamwork are emphasized by the processes and criteria used in the Pre-License Operator Training Program at SONGS

  5. Surry Power Station, Units 1 and 2. Semiannual operating report, July--December 1974

    International Nuclear Information System (INIS)

    1974-01-01

    Net electric power generated by Surry Unit 1 was 6,930,353 MWH with the generator on line for 10,417.7 hours. Net electric power generated by Unit 2 was 5,699,299 MWH with the generator on line for 8,384.2 hours. Information is presented concerning operation, radioactive effluent releases, solid radioactive wastes, fuel shipments, occurrences in which temperature limitations on the condenser cooling water discharge were exceeded, changes in station organization, occupational personnel radiation exposure, nonradiological monitoring including thermal, physical, and biological programs, and the radiological environmental monitoring program. (U.S.)

  6. A novel feedwater system for the RETRAN model of the Palo Verde nuclear generating station

    International Nuclear Information System (INIS)

    Secker, P.A.; Webb, J.R.

    1988-01-01

    This paper presents a feedwater system model which supplies realistic boundary conditions to the RETRAN model of a Palo Verde Nuclear Generating Station reactor plant. The RETRAN thermal hydraulic code is used to analyze nuclear reactor system transients through a generalized thermal hydraulic volume/junction network. The feedwater system model is implemented using the control block modeling option available in the RETRAN code. The output of the control block model is coupled to the thermal hydraulic network by a fill junction. A forward Euler integration scheme is used by RETRAN for control block variables. The feedwater system model is formulated to allow implicit integration within the existing code framework. The potential need for small integration time steps is, therefore, alleviated. The model results are compared with test data

  7. Technical Specifications, Seabrook Station, Unit 1 (Docket No. 50-443): Appendix ''A'' to License No. NPF-67

    International Nuclear Information System (INIS)

    1989-05-01

    The Seabrook Station, Unit 1 Technical Specifications were prepared by the US Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  8. 77 FR 35079 - License Renewal Application for Seabrook Station, Unit 1 ; NextEra Energy Seabrook, LLC

    Science.gov (United States)

    2012-06-12

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-443; NRC-2010-0206] License Renewal Application for Seabrook Station, Unit 1 ; NextEra Energy Seabrook, LLC AGENCY: Nuclear Regulatory Commission. ACTION: License renewal application; intent to prepare supplement to draft [[Page 35080

  9. IEEE recommended practices for seismic qualification of Class 1E equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The IEEE has developed this document to provide direction for developing programs to seismically qualify Class 1E equipment for nuclear power generating stations. It supplements IEEE Std 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, which describes the basic requirements for equipment qualification. The Class 1E equipment to be qualified by produres or standards established by this document are of many forms, characteristics, and materials; therefore, the document presents many acceptable methods with the intent of permitting the user to make a judicious selection from among the various options. In making such a selection, the user should choose those that best meet a particular equipment's requirements. Further, in using this document as a specification for the purchase of equipment, the many options should also be recognized and the document invoked accordingly. It is recommended that the need for specific standards for the seismic qualifiction of particular kinds of equipment be evaluated by those responsible for such documents and that consideration be given to the application of particular methods from these documents which are most suitable

  10. Final environmental statement related to construction of Cherokee Nuclear Station, Units 1, 2, and 3: (Docket Nos. STN 50-491, STN 50-492, and STN 50-493)

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of a construction permit to the Duke Power Company for the construction of the Cherokee Nuclear Station (CNS) Units 1, 2, and 3 located in Cherokee County, South Carolina. A total of 2263 acres will be removed from public use for the CNS site. Construction-related activities on the site will disturb about 751 acres. Approximately 654 acres of land will be required for transmission line right-of-way, and a railroad spur will affect 83 acres. This constitutes a minor regional impact. No significant environmental impacts are anticipated from normal operational releases of radioactive materials. The total annual dose to the US population (total body plus thyroid) from operation of the plant is 210 man-rems which is less than the normal fluctuations in the background dose this population would receive. The occupational dose is approximately 1400 man-rems/year. The heat dissipation system will require a maximum water makeup of 55,814 gpm, of which 50,514 gpm will be consumed due to drift and evaporative losses. This amount represents 4.5% of the mean monthly flow and 23.8% of the low flow of the Broad River. The cooling tower blowdown and chemical effluents from the station will increase the dissolved solids concentration in the river by a maximum of 44 ppM. The thermal alterations and increases in total dissolved solids concentration will not significantly affect the aquatic productivity of the river. 114 refs., 25 figs., 46 tabs

  11. IAEA International Peer Review Mission on Mid-and-Long-Term Roadmap Towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station Units 1-4, Tokyo and Fukushima Prefecture, Japan, 15-22 April 2013. Mission Report

    International Nuclear Information System (INIS)

    2013-01-01

    Following the accident at TEPCO's Fukushima Daiichi Nuclear Power Station (NPS) on 11 March 2011, the ''Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station Units 1-4'' was adopted by the Government of Japan and TEPCO Council on Mid-to-Long-Term Response for Decommissioning in December 2011 and revised in July 2012. The Roadmap, which is scheduled for an additional update in June 2013, describes the main steps and activities to be implemented for the decommissioning of the Fukushima Daiichi NPS through the combined efforts of the Government of Japan and TEPCO. Within the framework of the IAEA Action Plan on Nuclear Safety, the Government of Japan invited the IAEA to conduct an independent peer review of the Roadmap with two main objectives: - To improve the decommissioning planning and the implementation of pre-decommissioning activities at TEPCO's Fukushima Daiichi NPS; and - To share with the international community the good practices and lessons learned by the review. The review has been organized in two steps, and the IAEA conducted the first part in Japan from 15 to 22 April 2013. The objective of the first mission was to undertake an initial review of the Roadmap, including assessments of decommissioning strategy, planning and timing of decommissioning phases and a review of several specific short-term issues and recent challenges. Specifically, it covered the assessment of current reactor conditions, assessment of management of radioactive releases and associated doses, control of radioactive exposure of employees and decontamination within the site for improvement of working environment, structural integrity of reactor buildings and other constructions. The incidents recently experienced at the site, related with failures of the power supply and leakages of water from the underground reservoirs, were also included in the review of the specific short-term issues. The Government of Japan and TEPCO have

  12. IEEE standard criteria for type tests of class 1E modules used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The Institute of Electrical and Electronics Engineers has generated this document to provide direction for type testing Class 1E modules and obtaining specific type test data. It supplements IEEE Std 323-1974, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, which describes the basic requirements for Class 1E equipment qualification. Adherence to this document alone may not suffice for assuring public health and safety because it is the integrated performance of the structures, the fluid systems, the electrical systems, the instrumentation systems of the station, and in particular, the plant protection system of which these modules are a part that prevents accidents or limits the consequences of accidents. Each applicant to the Nuclear Regulatory Commission for a license to operate a nuclear power generating station has the responsibility to assure himself and others that this document, if used, is pertinent to his application and that the integrated performance of his station is adequate

  13. Use of intelligent loop diagrams at San Onofre Nuclear Generation Station (SONGS)

    International Nuclear Information System (INIS)

    Groves, J.E.; Johnson, K.I.; Foulk, J.; Reinschmidt, K.F.; Tutos, N.C.

    1991-01-01

    The use of advanced information systems will result in five million dollars potential cost reduction and two years less time for producing over 2000 Instrumentation and Control Loop Diagrams for the three nuclear units at San Onofre Nuclear Generating Station (SONGS). This new information technology will also assist plant management at SONGS in generating even larger savings from reduction in operations and maintenance costs. The key element of the new solution is the use of plant drawings, the traditional primary source of plant information, for on-line access to all plant databases and information systems, by replacing paper drawings with intelligent electronic drawings. The implementation of this concept for the Instrumentation and Control Loop Diagrams, presently in progress, is part of the Integrated Nuclear Data Management Systems (INDAMS) program at SONGS, a joint effort which includes support from Stone and Webster Advanced Systems Development Services, International Business Machines Corporation (IBM), and Dassault Systems of France. The initial results have encouraged plant management to speed up the implementation process

  14. 75 FR 53985 - Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit 3; Temporary...

    Science.gov (United States)

    2010-09-02

    ... NUCLEAR REGULATORY COMMISSION [Docket No. STN 50-530; NRC-2010-0281] Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit 3; Temporary Exemption 1.0 Background Arizona Public Service Company (APS, the licensee) is the holder of Facility Operating License No. NPF-74, which...

  15. Purification and solidification of reactor wastes at a Canadian nuclear generating station

    International Nuclear Information System (INIS)

    Buckley, L.P.; Burt, D.A.

    1981-06-01

    Chalk River Nuclear Laboratories are developing methods to condition power reactor wastes and to immobilize their radionuclides. Evaporation alone and combined with bituminization has been an important part of the program. After testing at the laboratories a 0.5 m 2 wiped-film evaporator was sent to the Douglas Point Nuclear Generating Station (220 MWe) to demonstrate its suitability to handle typical reactor liquid wastes. Two specific tasks undertaken with the wiped-film evaporator were successfully completed. The first was purification of contaminated heavy water which had leaked from the moderator circuit. The heavy water is normally recovered, cleaned by filters and ion-exchange resin and then upgraded by electrolysis. Cleaning the heavy water with the wiped-film evaporator produced better quality water for upgrading than had been achieved by any previous method and at much lower operating cost. The second task was to concentrate and immobilize a decontamination waste. The waste was generated from the decontamination of pump bowls used in the primary heat transport circuit. The simultaneous addition of the liquid waste and bitumen emulsion to the wiped-film evaporator produced a solid containing 30 wt% waste solids in a bitumen matrix. The volume reduction achieved was 16:1 based on the volumes of initial liquid waste and the final product generated. The quantity sent to storage was 20 times less than had the waste been immobilized in a cement matrix. The successful demonstration has resulted in a proposal to install a wiped-film evaporator at the station to clean heavy water and immobilize decontamination wastes. (author)

  16. The structural aging assessment program: ranking methodology for CANDU nuclear generating station concrete components

    International Nuclear Information System (INIS)

    Philipose, K.E.; Muhkerjee, P.K.; McColm, E.J.

    1997-01-01

    Most of the major structural components in CANDU nuclear generating stations are constructed of reinforced concrete. Although passive in nature, these structures perform many critical safety functions in the operation of each facility. Aging can affect the structural capacity and integrity of structures. The reduction in capacity due to aging is not addressed in design codes. Thus a program is warranted to monitor the aging of safety-related CANDU plant structures and to prioritize those that require maintenance and repairs. Prioritization of monitoring efforts is best accomplished by focusing on those structures judged to be the most critical to plant performance and safety. The safety significance of each sub-element and its degradation with time can be evaluated using a numerical rating system. This will simplify the utility's efforts, thereby saving maintenance costs while providing a higher degree of assurance that performance is maintained. This paper describes the development of a rating system (ranking procedure) as part of the Plant Life Management of CANDU generating station concrete structures and illustrates its application to an operating plant. (author)

  17. Electric utilities deregulation and its impact on nuclear power generating stations

    International Nuclear Information System (INIS)

    Trehan, N.K.

    1998-01-01

    Under restructuring and deregulation, it is not clear as to who would have the responsibility, and what obligations the market participants would have to ensure that the electrical system reliability (stability) is maintained. Due to the dynamic nature of the electrical grid, especially with the implementation of restructuring and deregulation, vulnerabilities exist which may impact the reliability (stability) of the offsite electrical power system. In a nuclear power generating unit, an offsite electric power system and an onsite electric power system are required to permit the functioning of structures, systems, and components which are important to safety. The safety function for each system is to provide sufficient capacity and capability to assure that the containment integrity is maintained during power operation or in the event of a postulated accident. Analyses performed by the applicants must verify that the electrical grid remains stable in the event of a loss of the nuclear unit generator, the largest other unit on the grid or the most critical transmission line. The stability of the electric grid is assumed in the safety analyses and a change in it would impact those analyses. However, it may impact the availability of a stable electric power to the safety buses because of the limited number of available transmission lines. This paper discusses electrical power generation and demand, reserve margins, power transfer capability, development of new innovative technologies to compensate for lack of the construction of transmission lines, legislation for the formulation of a self regulation organization (SRO), grid disturbances that may lead to a voltage collapse, and the vulnerabilities which may impact the availability of a stable power to the nuclear power generating stations

  18. Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354)

    International Nuclear Information System (INIS)

    1984-10-01

    The Safety Evaluation Report for the application filed by Public Service Electric and Gas Company, as applicant, for a license to operate the Hope Creek Generating Station (Docket No. 50-354), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Salem County, New Jersey. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  19. Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement No. 1

    International Nuclear Information System (INIS)

    1985-03-01

    Supplement No. 1 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company as applicant for itself and Atlantic City Electric Company, as owners, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  20. Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement No. 4

    International Nuclear Information System (INIS)

    1985-12-01

    Supplement No. 4 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  1. Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354)

    International Nuclear Information System (INIS)

    1985-10-01

    Supplement No. 3 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report. 6 tabs

  2. Safety evaluation report related to the operation of Hope Creek Generation Station (Docket No. 50-354). Supplement No. 5

    International Nuclear Information System (INIS)

    1986-04-01

    Supplement No. 5 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of the publication of the Safety Evaluation Report

  3. Safety evaluation report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement 2

    International Nuclear Information System (INIS)

    1985-08-01

    Supplement No. 2 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company as applicant for itself and Atlantic City Electric Company, as owners, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  4. Some causes of vibrations recorded by in-service diagnostic systems in steam generators of units 1 and 2 of Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Sadilek, J.; Matal, O.

    1989-01-01

    A brief description is presented of the design of the steam generators of the first and second units of the Dukovany nuclear power plant. Attention is also given to the feed water systems and the diagnostic systems. The causes are analyzed of the irregularly occurring vibrations in the steam generators in service. It is demonstrated that the source of the vibrations transmitted to the steam generators are the valves in the feeding tract. The vibrations are induced by dynamic forces from the feed water. Reducing the water pressure at the delivery of the electric feed pumps by reducing the size of the rotor, etc., does not remove all vibrations. It is therefore recommended that valves be ins+alled with better regulating characteristics. (Z.M.). 6 figs., 1 tab., 3 refs

  5. Radioactive emission data from Canadian nuclear generating stations 1986 to 1995

    International Nuclear Information System (INIS)

    1997-05-01

    All nuclear generating station (NGSs) release small quantities of radioactivity in a controlled manner into both the atmosphere (as gaseous effluents) and adjoining water bodies (as liquid effluents). The purpose of this document is to report on the magnitude of these emissions for each operating NGS in Canada and to indicate how these emissions compare with the relevant limitations imposed by the AECB as part of its regulatory and licensing program. The data show that the levels of emissions of gaseous and liquid effluents from all currently operating NGSs are well below the values mandated by the AECB. In fact, since 1987 no emissions have exceeded 1% of those values. 3 tabs., 46 figs

  6. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Frescura, G.M.; Smith, A.J.; Lau, J.H.

    1991-01-01

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  7. The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

    1993-01-01

    The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

  8. Environmental radiological studies downstream from Rancho Seco Nuclear Power Generating Station

    International Nuclear Information System (INIS)

    Noshkin, V.E.; Wong, K.M.; Eagle, R.J.; Dawson, J.W.; Brunk, J.L.; Jokela, T.A.

    1985-01-01

    This report summarizes the information compiled in 1984 while assessing the environmental impact of radionuclides in aquatic releases from the Rancho Seco Nuclear Power Generating Station. Gamma-emitting radionuclides discharged since 1981 are found in many of the dietary components derived from the creeks receiving the effluent wastewater. Some soils and crops are found to contain radionuclides that originate from the contaminated water that was transferred to land during the irrigation season. 134 Cs and 137 Cs are the primary gamma-emitting radionuclides detected in the edible flesh of fish from the creeks. Concentrations in the flesh of fish decreased exponentially with distance from the plant. No significant differences in the 137 Cs activity were found between male and female fish of equal size, but concentrations may vary in fish of different size, with the season and diet. 21% of the total 137 Cs and 134 Cs discharged between 1981 and 1984 is associated with the creek sediments to a distance of 27 km from the plant. Fractions of the missing inventory have been transferred to land during the irrigation season or to downstream regions more distant than 27 km from the plant. The radiocesium content of the sediments in 1984 decreased significantly in a downstream direction, much in the same manner as concentrations decreased in fish. Radioactivity originating from the plant was not above detection limits in any terrestrial food item sampled beyond 6.5 km from the plant. Based on the usage factors provided by individuals interviewed in a 1984 survey, the fish and aquatic-organism ingestion pathway contributed the largest radiological dose to humans utilizing products contaminated with the radionuclides in the liquid wastes discharged from the Rancho Seco Nuclear Power Generating Station in 1984

  9. Measurements of tritium (HTO, TFWT, OBT) in environmental samples at varying distances from a nuclear generating station

    International Nuclear Information System (INIS)

    Kotzer, T.G.; Workman, W.J.G.

    1999-12-01

    Concentrations of tritium have been measured in environmental samples (vegetation, water, soil, air) from sites distal and proximal to a CANDU nuclear generating station in Southern Ontario (OPG-Pickering). Levels of tissue-free water tritium (TFWT) and organically bound tritium (OBT) in vegetation are as high as 24,000 TU immediately adjacent to the nuclear generating station and rapidly decrease to levels of tritium which are comparable to natural ambient concentrations for tritium in the environment (approximately ≤ 60 TU). Tritium concentrations (OBT, TFTW) have also been measured in samples of vegetation and tree rings growing substantial distances away from nuclear generating stations and are within a factor of 1 to 2 of the ambient levels of tritium measured in precipitation in several parts of Canada (approximately ≤30 TU). (author)

  10. Measurements of tritium (HTO, TFWT, OBT) in environmental samples at varying distances from a nuclear generating station

    Energy Technology Data Exchange (ETDEWEB)

    Kotzer, T.G.; Workman, W.J.G

    1999-12-01

    Concentrations of tritium have been measured in environmental samples (vegetation, water, soil, air) from sites distal and proximal to a CANDU nuclear generating station in Southern Ontario (OPG-Pickering). Levels of tissue-free water tritium (TFWT) and organically bound tritium (OBT) in vegetation are as high as 24,000 TU immediately adjacent to the nuclear generating station and rapidly decrease to levels of tritium which are comparable to natural ambient concentrations for tritium in the environment (approximately {<=} 60 TU). Tritium concentrations (OBT, TFTW) have also been measured in samples of vegetation and tree rings growing substantial distances away from nuclear generating stations and are within a factor of 1 to 2 of the ambient levels of tritium measured in precipitation in several parts of Canada (approximately {<=}30 TU). (author)

  11. IEEE standard requirements for reliability analysis in the design and operation of safety systems for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The purpose of this standard is to provide uniform, minimum acceptable requirements for the performance of reliability analyses for safety-related systems found in nuclear-power generating stations, but not to define the need for an analysis. The need for reliability analysis has been identified in other standards which expand the requirements of regulations (e.g., IEEE Std 379-1972 (ANSI N41.2-1972), ''Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection System,'' which describes the application of the single-failure criterion). IEEE Std 352-1975, ''Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems,'' provides guidance in the application and use of reliability techniques referred to in this standard

  12. 77 FR 69506 - Exelon Generation Company, LLC., Oyster Creek Nuclear Generating Station; Exemption

    Science.gov (United States)

    2012-11-19

    ... unit'' to ``The closed head automatic sprinkler system in the condenser bay area was designed... the time of design.'' Dated in Rockville, Maryland, this 9th day of November 2012. For the Nuclear...

  13. 77 FR 64834 - Exelon Generation Company, LLC; Oyster Creek Nuclear Generating Station; Exemption

    Science.gov (United States)

    2012-10-23

    ... oil unit,'' to ``The closed head automatic sprinkler system in the condenser bay area was designed... the time of design.'' Dated in Rockville, Maryland, this 15th day of October 2012. For the Nuclear...

  14. 77 FR 135 - Exelon Generation Company, LLC, Oyster Creek Nuclear Generating Station; Exemption

    Science.gov (United States)

    2012-01-03

    ... not have a significant effect on the quality of the human environment (76 FR 79227, December 21, 2011.... For The Nuclear Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor...

  15. IEEE Std 649-1991: IEEE standard for qualifying Class 1E motor control centers for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The basic principles, requirements, and methods for qualifying Class 1E motor control centers for both harsh and mild environment applications in nuclear power generating stations are described. In addition to defining specific qualification requirements for Class 1E motor control centers and their components in accordance with the more general qualification requirements of IEEE Std 323-1983, this standard is intended to provide guidance in establishing a qualification program for demonstrating the adequacy of Class 1E motor control centers in nuclear power generating station applications

  16. Type test of Class 1E electric cables, field splices, and connections for nuclear power generating stations - 1975

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class 1E Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices

  17. Nuclear Regulatory Commission issuances, March 1975

    International Nuclear Information System (INIS)

    1975-04-01

    Reactor licensing actions taken by the Nuclear Regulatory Commission, the Atomic Safety and Licensing Appeal Board and the Atomic Safety and Licensing Boards for March 1975 are presented. Action was included for the following reactors: Big Rock Point Nuclear Plant; West Valley Reprocessing Plant; Limerick Generating Station, Units 1 and 2; Midland Plants, Units 1 and 2; Wolf Creek Generating Station, Unit 1; Monticello Nuclear Generating Plant, Unit 1; Douglas Point Nuclear Generating Station, Units 1 and 2; Seabrook Station, Units 1 and 2; Vermont Yankee Nuclear Power Station; and WPPSS Hanford Units 1 and 4. (U.S.)

  18. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Apendices

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A; Lopez M, R [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1995-07-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and.analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  19. Different methods of informing the public on projected sites of nuclear generating stations. Results in France

    International Nuclear Information System (INIS)

    Taccoen, L.B.C.

    1977-01-01

    The situation in France of local public opinion concerning nuclear power stations is very different, according to the region in question. It seems that in industrialized regions, where the inhabitants are already familiar with nuclear energy, public opinion is much more open to the installation of new reactors. In regions which are only slightly industrialized, however, the prejudices of the population against nuclear energy can be very strong. This situation is quite serious because it could lead to an even greater widening of the development gap between various regions in France. Electricite de France has thus been led to develop a program of public information concerning the problems of nuclear energy. In practice, this program consists of: a policy of active presence in meeting and debates as well as at the actual site; eventual opening of information centers; and organization of visists of existing nuclear power stations [fr

  20. Environmental monitoring report for the Point Lepreau, New Brunswick nuclear generating station 1984

    International Nuclear Information System (INIS)

    Nelson, R.W.P.; Ellis, K.M.; Smith, J.N.

    1986-07-01

    The Point Lepreau Environmental Monitoring Program (PLEMP) has been established within the Department of Fisheries and Oceans to assess the environmental impact of radioactive, thermal, and chemical releases from the Point Lepreau, N.B. Nuclear Generating Station (NGS) located on the Bay of Fundy. This report contains the results for the second year of the operational phase of the monitoring program. Emphasis has been placed on those areas where effects from the NGS operation were expected to be observed. Further studies were carried out on the characterization of the thermal plume during several pre-arranged tritium releases. Attention was given to tritium levels at the outfall, in the atmosphere and in biological samples. An initial attempt was made to relate the distribution of tritium in the Point Lepreau area in various phases to local meteorological conditions. Radionuclide levels in lichen, including, for the first time, aerial lichen on a regular basis, an efficient accumulator of atmospheric particulates were monitored. In addition, radionuclide measurements were made on samples collected during the pre-operational phase from the major environmental reservoirs and these radioactivity levels were compared to previous measurements to assess the impact of the operation of the NGS. The purpose of this program is to provide government with a comprehensive scientific basis upon which to assess the environmental implications of the operation of nuclear reactors in coastal regions

  1. Palo Verde Nuclear Generating Station: an example of the state role in regional nuclear projects

    International Nuclear Information System (INIS)

    Pasternak, A.

    1980-10-01

    A nuclear power plant siting policy which confines new construction to existing sites will lead to the formation of large regional power centers, each involving many utilities from several states. The Palo Verde Nuclear Project in Arizona has been examined in terms of the role state regulation plays in large regional nuclear projects. State regulatory processes do not reflect the regional nature of large power centers. Decisions and actions by individual state regulatory commissions create risk and uncertainty for all the utility participants in regional projects. A climate and mechanism to encourage and facilitate interstate cooperation are needed to enhance the viability of the confined siting policy and the regional power center concept

  2. Steam generator replacement in Bruce A Unit 1 and Unit 2

    International Nuclear Information System (INIS)

    Hart, R.S.

    2007-01-01

    The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Unit 2 that will extend the in-service tile of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from he bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

  3. The vibrational behaviour of the generator support structure for Koeberg nuclear power station at high frequencies

    International Nuclear Information System (INIS)

    Lee, D.E.

    1988-06-01

    The vibrational behaviour of the generator support structure at Koeberg nuclear power station at frequencies primarily in the region of 80 Hz to 110 Hz was examined. The effect of soil-structure interaction and the change in stiffness of the foundation soil was investigated. Vibration tests were performed on the generator support structure and the results were compared with a theoretical finite element analysis of the structure. By varying the soil-cement foundation stiffness it was possible to demonstrate the change in dynamic behaviour of the structure in the higher frequency band 80 Hz to 110 Hz. Comment has been made on the design code DIN 4024 in view of the findings of this thesis. It was concluded that the empirical rules regarding the inclusion of the foundation in an analysis specified by the code do not cover all cases and greater cognisance of the effect of the foundation stiffness on the vibration behaviour of such machine foundations is necessary. Obvious machine frequencies higher than the operational frequencies should be analysed where it is considered necessary. 24 refs., 25 tabs., 83 figs

  4. A recommended approach for calculating degraded voltage relay setpoints for nuclear generating stations

    International Nuclear Information System (INIS)

    Jancauskas, J.R.

    1994-01-01

    The purpose of degrading voltage relays (DVRs) is to ensure that adequate voltage is available to operate all Class 1E loads at all voltage distribution levels. Should voltage drop below the setpoint of the DVRs, the Class 1E power system is disconnected from its supply and resequenced onto the diesel generators in order to restore system voltages to acceptable levels. These relays represent one of the two levels of voltage protection required for the onsite power system. Determining the proper setpoint for degraded voltage relays in nuclear generating stations is a complex task which requires a complete understanding of the Class 1E power distribution system. Despite the importance of degraded voltage relay setpoint calculations, most of the available references only give clues on how not to set these relays rather than provide guidance on how to determine the appropriate setpoint. This paper presents an approach for performing these calculations which attempts to ensure that all of the relevant design issues are addressed

  5. Nuclear Regulatory Commission issuances. Volume 17, No. 3

    International Nuclear Information System (INIS)

    1983-03-01

    This report contains the Issuances received during March 1983 from the Commission (CLI), the Atomic Safety and Licensing Appeal Boards (ALAB), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judge (ALJ), the Directors Decisions (DD), and the Denials of Petition for Rulemaking (DPRM). The Issuances concerned the following facilities: Three Mile Island Nuclear Station, Unit No. 1; Comanche Peak Steam Electric Station, Units 1 and 2; Vallecitos Nuclear Center; Floating Nuclear Power Plants; San Onofre Nuclear Generating Station, Units 2 and 3; Point Beach Nuclear Plant, Unit 1; Perry Nuclear Power Plant, Units 1 and 2; Shoreham Nuclear Power Station, Unit 1; Western New York Nuclear Service Center; Limerick Generating Station, Units 1 and 2; Seabrook Station, Units 1 and 2; Black Fox Station, Units 1 and 2; WmH Zimmer Nuclear Power Station, Unit 1; WPPSS Nuclear Project No. 1; Zion Nuclear Plant, Units 1 and 2; and South Texas Project, Units 1 and 2

  6. AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1995

    International Nuclear Information System (INIS)

    1996-06-01

    This report is the Atomic Energy Control Board (AECB) staff assessment of safety at the Pickering Nuclear Generating Station (PNGS-A and PNGS-B) for 1995. Our on-site Project Officers and Ottawa-based specialists monitored the stations throughout the year. In 1995, compliance with the Transportation Packaging of Radioactive Materials Regulations and the Cost Recovery Fees Regulations was satisfactory. The performance of the special safety systems was good. Releases of radioactive materials from the station were low and well below the legal limits for public safety. 10 tabs., 7 figs

  7. Nuclear power stations licensing

    International Nuclear Information System (INIS)

    Solito, J.

    1978-04-01

    The judicial aspects of nuclear stations licensing are presented. The licensing systems of the United States, Spain, France and Federal Republic of Germany are focused. The decree n 0 60.824 from July 7 sup(th), 1967 and the following legislation which define the systematic and area of competence in nuclear stations licensing are analysed [pt

  8. Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station from August 20--30, 1992

    International Nuclear Information System (INIS)

    Hebdon, F.J.

    1993-03-01

    On August 24, 1992, Hurricane Andrew, a Category 4 hurricane, struck the Turkey Point Electrical Generating Station with sustained winds of 145 mph (233 km/h). This is the report of the team that the US Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO) jointly sponsored (1) to review the damage that the hurricane caused the nuclear units and the utility's actions to prepare for the storm and recover from it, and (2) to compile lessons that might benefit other nuclear reactor facilities

  9. Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station from August 20--30, 1992. [Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hebdon, F.J. [Institute of Nuclear Power Operations, Atlanta, GA (United States)

    1993-03-01

    On August 24, 1992, Hurricane Andrew, a Category 4 hurricane, struck the Turkey Point Electrical Generating Station with sustained winds of 145 mph (233 km/h). This is the report of the team that the US Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO) jointly sponsored (1) to review the damage that the hurricane caused the nuclear units and the utility`s actions to prepare for the storm and recover from it, and (2) to compile lessons that might benefit other nuclear reactor facilities.

  10. Aging of nuclear station diesel generators: Evaluation of operating and expert experience: Workshop

    International Nuclear Information System (INIS)

    Hoopingarner, K.R.; Vause, J.W.

    1987-08-01

    Pacific Northwest Laboratory (PNL) evaluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear service. The research, sponsored by the US Nuclear Regulatory Commission (NRC), identified and characterized the contribution of aging to emergency diesel generator failures. This report, Volume II, reports the results of an industry-wide workshop held on May 28 and 29, 1986, to discuss the technical issues associated with aging of nuclear service emergency diesel generators. The technical issues discussed most extensively were: man/machine interfaces, component interfaces, thermal gradients of startup and cooldown and the need for an accurate industry database for trend analysis of the diesel generator system

  11. Control room habitability survey of licensed commercial nuclear power generating stations

    International Nuclear Information System (INIS)

    Driscoll, J.W.

    1988-10-01

    This document presents the results of a survey of control room habitability systems at twelve commercial nuclear generating stations. The survey, conducted by Argonne National Laboratory (ANL), is part of an NRC program initiated in response to concerns and recommendations of the Advisory Committee on Reactor Safeguards (ACRS). The major conclusion of the report is that the numerous types of potentially significant discrepancies found among the surveyed plants may be indicative of similar discrepancies throughout the industry. The report provides plant-specific and generalized findings regarding safety functions with respect to the consistency of the design, construction, operation and testing of control room habitability systems and corresponding Technical Specifications compared with descriptions provided in the license basis documentation including assumptions in the operator toxic gas concentration and radiation dose calculations. Calculations of operator toxic gas concentrations and radiation doses were provided in the license basis documentation and were not performed by the ANL survey team. Recommendation for improvements are provided in the report

  12. Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

    International Nuclear Information System (INIS)

    Serdula, K.J.

    1988-01-01

    Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

  13. IEEE standard for design qualification of safety systems equipment used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This standard is written to serve as a general standard for qualification of all types of safety systems equipment, mechanical and instrumentation as well as electrical. It also establishes principles and procedures to be followed in preparing specific safety systems equipment standards. Guidance for qualifying specific safety systems equipment may be found in various specific equipment qualification standards that are now available or are being prepared. It is required that safety systems equipment in nuclear power generating stations meet or exceed its performance requirements throughout its installed life. This is accomplished by a disciplined program of design qualification and quality assurance of design, production, installation, maintenance and surveillance. This standard is for the design qualification section of the program only. Design qualification is intended to demonstrate the capability of the equipment design to perform its safety function(s) over the expected range of normal, abnormal, design basis event, post design basis event, and in-service test conditions. Inherent to design qualification is the requirement for demonstration, within limitations afforded by established technical state-of-the-art, that in-service aging throughout the qualified life established for the equipment will not degrade safety systems equipment from its original design condition to the point where it cannot perform its required safety function(s), upon demand. The above requirement reflects the primary role of design qualification to provide reasonable assurance that design- and age-related common failure modes will not occur during performance of safety function(s) under postulated service conditions

  14. Spectral analysis of coolant activity from a commercial nuclear generating station

    International Nuclear Information System (INIS)

    Swann, J.D.; Lewis, B.J.; Ip, M.

    2008-01-01

    In support of the development of a real-time on-line fuel failure monitoring system for the CANDU reactor, actual gamma spectroscopy data files from the gaseous fission product (GFP) monitoring system were acquired from almost four years of operation at a commercial Nuclear Generating Station (NGS). Several spectral analysis techniques were used to process the data files. Radioisotopic activity from the plant information (PI) system was compared to an in-house C++ code that was used to determine the photopeak area and to a separate analysis with commercial software from Canberra-Aptec. These various techniques provided for a calculation of the coolant activity concentration of the noble gas and iodine species in the primary heat transport system. These data were then used to benchmark the Visual DETECT code, a user friendly software tool which can be used to characterize the defective fuel state based on a coolant activity analysis. Acceptable agreement was found with the spectral techniques when compared to the known defective bundle history at the commercial reactor. A more generalized method of assessing the fission product release data was also considered with the development of a pre-processor to evaluate the radioisotopic release rate from mass balance considerations. The release rate provided a more efficient means to characterize the occurrence of a defect and was consistent with the actual defect situation at the power plant as determined from in-bay examination of discharged fuel bundles. (author)

  15. Bradwell Nuclear Power Station

    International Nuclear Information System (INIS)

    1987-01-01

    When built, the Magnox reactors were expected to have operating lifetimes of 20-25 years. In order to satisfy the licensing authorities of their continued safety, long term safety reviews (LTSRs) are being carried out as the reactors reach 20 years of operation. This is the Nuclear Installations Inspectorate's (NII) summary report on Bradwell nuclear power station. The objectives of the LTSR are stated. A description of the plant is followed by an explanation of the statutory position on licensing. The responsibilities of the Central Electricity Generating Board (CEGB) and the NII are defined. From the examination of the CEGB's LTSR it is concluded that this generally confirms the validity of the existing safety case for present operation. However, some recommendations are made as to work required for reactor operation up to 1992. A summary of the NII findings is presented. This includes the reactor pressure circuit integrity, effects of ageing and in-service wear and radiation doses. (U.K.)

  16. Evaluation of Perry Nuclear Power Plant Unit 1 technical specifications

    International Nuclear Information System (INIS)

    Baxter, D.E.; Bruske, S.J.

    1985-11-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Perry Nuclear Power Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Perry T/S. Several discrepancies were identified and subsequently resolved through telephone conversations with the staff reviewer and the utility representative. Pending completion of the resolutions noted in Parts 3 and 4 of this report, the Perry Nuclear Power Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

  17. Evaluation of Watts Bar Nuclear Plant Unit 1 Technical Specifications

    International Nuclear Information System (INIS)

    Baxter, D.E.; Bruske, S.J.

    1985-08-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Watts Bar Nuclear Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumption of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Watts Bar T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Watts Bar Nuclear Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

  18. Work-in-progress report on maintenance good practices for motors in nuclear power generating stations - Part 2

    International Nuclear Information System (INIS)

    Eckstein, E.R.; Hall, A.R.; Kasper, S.; Smith, P.R.; Subudhi, M.

    1988-01-01

    Motor maintenance is addressed by a variety of industry standards, utility procedures and manufacture's instructions, as well as by recently published assessments of nuclear and non-nuclear operating experience and motor reliability data. This experience supports the premise that motor failure rates are in large part related to application stress factors such as operating time, environmental factors and vibration. Due to the large number of critical motors installed in a nuclear generating station, additional considerations are necessary to effectively schedule their preventive maintenance activities. Therefore, in this part, a method for prioritizing motor maintenance based on application stresses experienced by the motors is presented

  19. Tritium releases from the Pickering Nuclear Generating Station and birth defects and infant mortality in nearby communities 1971-1988

    International Nuclear Information System (INIS)

    Johnson, K.C.; Rouleau, J.

    1991-10-01

    This study was commissioned to examine whether there were elevated rates of stillbirth, birth defects, or death in the first year of life between 1971 and 1988 among offspring of residents of communities within a 25-kilometre radius of the Pickering Nuclear Generating Station. The study was also to investigate whether there were any statistical associations between the monthly airborne or waterborne tritium emissions from the Pickering Nuclear Generating Station and the rates of these reproductive outcomes. Overall analysis did not support a hypothesis of increased rates of stillbirths, neonatal mortality or infant mortality near the Pickering Nuclear Generating Station, or a hypothesis of increased birth prevalence of birth defects for 21 of 22 diagnostic categories. The prevalence of Down Syndrome was elevated in both Pickering and Ajax; however, there was no consistent pattern between tritium release levels and Down Syndrome prevalence, chance could not be ruled out for the associations between Down Syndrome and tritium releases or ground-monitored concentrations, the association was detected in an analysis where multiple testing was done which may turn up significant associations by change, and maternal residence at birth and early in pregnancy needs to be verified. The association between Down Syndrome and low-level radiation remains indeterminate when existing evidence from epidemiological studies is summed. The estimated radiation exposure from the nuclear plant for residents of Pickering and Ajax is lower by a factor of 100 than the normal natural background radiation. Further study is recommended. (21 tabs., 29 figs., 5 maps, 37 refs.)

  20. AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1996

    International Nuclear Information System (INIS)

    1997-06-01

    The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Pickering A and B Generating Stations for 1996. PNGS-A and PNGS-B operated safely during 1996. Although the risk to the workers and the public is low, major safety related changes are necessary at the stations and the sustainability of those changes needs to be demonstrated. Improvement is needed by Ontario Hydro in meeting the time limits for reporting reportable events. Ontario Hydro's follow up to events and causal factor analyses continue to need improvements. Improvements are needed to operational safety and reactor maintenance at both A and B. There are signs of improvement through Ontario Hydro's plan for recovery, and in station management changes. There also appears to be commitment to safety expressed at the highest level of the utility

  1. Aging of nuclear station diesel generators: Evaluation of operating and expert experience: Phase 1, Study

    International Nuclear Information System (INIS)

    Hoopingarner, K.R.; Vause, J.W.; Dingee, D.A.; Nesbitt, J.F.

    1987-08-01

    Pacific Northwest Laboratory evaluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear service. The research, sponsored by the US Nuclear Regulatory Commission (NRC), identified and characterized the contribution of aging to emergency diesel generator failures. This report, Volume I, reviews diesel-generator experience to identify the systems and components most subject to aging degradation and isolates the major causes of failure that may affect future operational readiness. Evaluations show that as plants age, the percent of aging-related failures increases and failure modes change. A compilation is presented of recommended corrective actions for the failures identified. This study also includes a review of current, relevant industry programs, research, and standards. Volume II reports the results of an industry-wide workshop held on May 28 and 29, 1986 to discuss the technical issues associated with aging of nuclear service emergency diesel generators

  2. Ecological studies of wood-boring bivalves in the vicinity of the Oyster Creek Nuclear Generating Station. Progress report Sep-Nov 81

    International Nuclear Information System (INIS)

    Hoagland, K.E.; Crocket, L.

    1982-06-01

    The species composition, distribution, and population dynamics of wood-boring bivalves are being studied in the vicinity of the Oyster Creek Nuclear Generating Station, Barnegat Bay, New Jersey. Untreated wood test panels are used to collect organisms at 12 stations. Physiological tolerances of 3 species are also under investigation in the laboratory. Competition among the species is being analyzed. In the fall of 1981, Teredo bartschi remained in Oyster Creek despite continuous prolonged outages of the Oyster Creek Nuclear Generating Station

  3. Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

    2014-01-01

    The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

  4. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  5. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2

  6. Safety Evaluation Report related to the operation of Hope Creek Generation Station (Docket No. 50-354). Supplement No. 6

    International Nuclear Information System (INIS)

    1986-07-01

    Supplement No. 6 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that has not been resolved at the time of the publication of the Safety Evaluation Report. This supplement supports the issuance of a full-power license to operate Hope Creek Generating Station

  7. IEEE Std 317-1972: IEEE standard for electric penetration assemblies in containment structures for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard prescribes the mechanical, electrical, and test requirements for the design, construction, and installation of electric penetration assemblies in containment structures for stationary nuclear power generating stations. The electric conductor and insulation characteristics of external circuits which connect to penetration assemblies are beyond the scope of these criteria. If there should be any conflict between this standard and those documents referenced herein, this standard shall take precedence over the referenced documents

  8. Confirmatory Survey Results for the Reactor Building Dome Upper Structural Surfaces, Rancho Saco Nuclear Generating Station, Herald, California

    International Nuclear Information System (INIS)

    Wade C. Adams

    2006-01-01

    Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006

  9. Steam generator leak detection at Bruce A Unit 1

    International Nuclear Information System (INIS)

    Maynard, K.J.; McInnes, D.E.; Singh, V.P.

    1997-01-01

    A new steam generator leak detection system was recently developed and utilized at Bruce A. The equipment is based on standard helium leak detection, with the addition of moisture detection and several other capability improvements. All but 1% of the Unit 1 Boiler 03 tubesheet was inspected, using a sniffer probe which inspected tubes seven at a time and followed by individual tube inspections. The leak search period was completed in approximately 24 hours, following a prerequisite period of several days. No helium leak indications were found anywhere on the boiler. A single water leak indication was found, which was subsequently confirmed as a through-wall defect by eddy current inspection. (author)

  10. The feature of emergency diesel generator relaying protection in Tianwan nuclear power station

    International Nuclear Information System (INIS)

    Jiang Xiaopeng; Shi Yan; Li Cong

    2014-01-01

    This paper mainly introduces the function and feature of emergency diesel generator in nuclear power plant, which plays an important role in nuclear accident. It minutely tells about the feature and configuration of relay protection and discusses the rationality of protection scheme, which shows that it can be completely contented all kinds of operation states. It is an analysis and argument about the principle of relay protection in detail, that would operate correctly when emergency diesel generator be in abnormal operating and serious fault conditions, such as cut off emergency diesel generator in order to avoid more harm to emergency diesel generator. It analyzes how the relay responses quickly and locks up the protection action under perturbations in the external power, so it can avoid unnecessary resection of emergency diesel generator to emergency power supply loss and effect of nuclear safety. It also analyzes the flexible use of protection setting of the protective relay to meet various operating status. It elaborates the particularity of relay protection which is due to the particularity of nuclear safety. It analyses the possibility of relay protection which has to be applied to other equipment and the protection setting that was provided by design institute, and puts forward the author's viewpoints. (authors)

  11. The nuclear power station

    International Nuclear Information System (INIS)

    Plettner, B.

    1987-04-01

    The processes taking place in a nuclear power plant and the dangers arising from a nuclear power station are described. The means and methods of controlling, monitoring, and protecting the plant and things that can go wrong are presented. There is also a short discourse on the research carried out in the USA and Germany, aimed at assessing the risks of utilising nuclear energy by means of the incident tree analysis and probability calculations. (DG) [de

  12. AECB staff annual assessment of the Point Lepreau Nuclear Generating Station for the year 1994

    International Nuclear Information System (INIS)

    1995-06-01

    This report is the Atomic Energy Control Board assessment of safety at Point Lepreau Generating Station for 1994. AECB on-site Project Officers, and Ottawa based specialists monitored the station throughout the year. The station operated safely during 1994. Performance of the special safety systems was very good, and NB Power made important progress improving the safety management of these systems. NB Power staff failed to comply with the conditions of our Operating Licence seven times during 1994. This is an improvement from last year, but the number of compliance problems is still too high. NB Power introduced safety culture training for its staff last year. AECB is optimistic that this will help to improve compliance. The station fully met the radiation safety requirements during the year. Releases of radioactive material to the environment are well below the limits set. Radiation exposure to workers are being well controlled, and AECB believe that NB Power is maintaining these exposures 'As Low As Reasonably Achievable' for this reactor type

  13. AECB staff annual assessment of the Point Lepreau Nuclear Generating Station for the year 1995

    International Nuclear Information System (INIS)

    1996-06-01

    The Atomic Energy Control Board conducts a staff assessment of safety at Point Lepreau Generating Station for 1995. On-site Project Officers and Ottawa based specialists monitored the station throughout the year. Point Lepreau operated safely during 1995 which was an unusual year for the station. The station was shut down for a planned five month outage to allow NB Power to complete maintenance on the reactor's pressure tubes. NB Power failed to comply with the terms of the Operating Licence we issue on fourteen occasions in 1995. In addition, NB Power reported an unusually large number of events. None of the events themselves directly affected public safety, if this level of performance continues unchecked, it might result in increased risk from operation in the future. Human error was an important feature of these problems. NB Power had already introduced safety culture training for their staff, but they will need to undertake further work urgently to resolve the problems. Three of Point Lepreau's special safety systems failed to meet their availability targets during 1995. NB Power's safety analysis program progress slowed during 1995, due to the additional workload from the outage. The Point Lepreau's training, quality assurance and safeguards related activities continued to function satisfactorily. They have assigned additional resources to emergency preparedness planning, which should enable them to take a more pro-active approach in this area. 7 tabs., 5 figs

  14. Socio-economic impacts of nuclear generating stations: Nine Mile Point and Fitzpatrick case study

    International Nuclear Information System (INIS)

    Branch, K.; Cochran, R.; Meale, R.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Nine Mile Point and Fitzpatrick nuclear power stations. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  15. Socio-economic impacts of nuclear generating stations: Crystal River Unit 3 case study

    International Nuclear Information System (INIS)

    Bergmann, P.A.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Crystal River Unit 3 nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  16. Socio-economic impacts of nuclear generating stations: Surry case study

    International Nuclear Information System (INIS)

    Flynn, J.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Surry nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  17. Socio-economic impacts of nuclear generating stations: Calvert Cliffs case study

    International Nuclear Information System (INIS)

    Flynn, J.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Calvert Cliffs nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  18. Socio-economic impacts of nuclear generating stations: Three Mile Island case study

    International Nuclear Information System (INIS)

    Flynn, C.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Three Mile Island nuclear power stations. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  19. Socio-economic impacts of nuclear generating stations: Rancho Seco case study

    International Nuclear Information System (INIS)

    Bergmann, P.A.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Rancho Seco nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  20. Socio-economic impacts of nuclear generating stations: Peach Bottom case study

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Peach Bottom nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  1. Socio-economic impacts of nuclear generating stations: Oconee case study

    International Nuclear Information System (INIS)

    Flynn, J.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the Oconee nuclear power stations. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of the construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  2. Socio-economic impacts of nuclear generating stations: D. C. Cook case study

    International Nuclear Information System (INIS)

    Branch, K.

    1982-07-01

    This report documents a case study of the socio-economic impacts of the construction and operation of the D. C. Cook nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  3. Socio-economic impacts of nuclear generating stations: Diablo Canyon case study

    International Nuclear Information System (INIS)

    Pijawka, K.D.; Yaquinto, G.

    1982-07-01

    This report documents a case study of the socioeconomic impacts of the construction and operation of the Diablo Canyon nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  4. Surry Power Station, Units 1 and 2. Semiannual operating report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Net electric power generated by Unit 1 was 2,315,124 MWH(e) and Unit 2 generated 2,062,954 MWH(e) with Unit 1 generator on line for 3,157.8 hrs and Unit 2 on line for 2,881.2 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, abnormal occurrences, and environmental monitoring. (FS)

  5. Point Lepreau generating station

    International Nuclear Information System (INIS)

    Ganong, G.H.D.; Strang, A.E.; Gunter, G.E.; Thompson, T.S.

    Point Lepreau-1 reactor is a 600 MWe generating station expected to be in service by October 1979. New Brunswick is suffering a 'catch up' phenomenon in load growth and needs to decrease dependence on foreign oil. The site is on salt water and extensive study has gone into corrosion control. Project management, financing and scheduling have unique aspects. (E.C.B.)

  6. Draft Environmental Statement related to the operation of Vogtle Electric Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50-425)

    International Nuclear Information System (INIS)

    1984-10-01

    This Draft Environmental Statement contains an assessment of the environmental impact associated with the operation of the Vogtle Electric Generating Plant, Units 1 and 2, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, part 51 (10 CFR 51), as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environmental impacts, environmental consequences and mitigating actions, and environmental and economic benefits and costs associated with station operation

  7. AECB staff annual assessment of the Bruce B Nuclear Generating Station for the year 1994

    International Nuclear Information System (INIS)

    1995-06-01

    AECB staff believes Ontario Hydro operated Bruce B safely in 1994. The Bruce B reactors will remain limited to 88% full power until Ontario Hydro is able to demonstrate that it is safe to operate at higher powers. Ontario Hydro's compliance with AECB regulations and the Operating Licence was satisfactory. AECB found no major violations. The station performance was similar to previous years. Radiation doses to workers and the public were well below the legal limits and also remained below Ontario Hydro's internal targets. Worker radiation doses increased slightly but were comparable to previous years. Inspection of pressure tubes and steam generator tubes by Ontario Hydro showed continuing tube degradation. However, we believe that Ontario Hydro made progress in correcting and managing these problems. Ontario Hydro carried out a full-scale fire drill at Bruce B in 1994. AECB witnessed the drill and were pleased to observe a significant improvement in the station's fire-fighting capability. 7 tabs., 4 figs

  8. Optimizing the use of operating experience at Ontario Hydro's Bruce Nuclear Generating Station 'A'

    International Nuclear Information System (INIS)

    Williams, E.L.

    1991-01-01

    One of the most significant lessons learned from the Three Mile Island event (March 1979), and again with the Chernobyl disaster - (April 1986) was the ongoing requirement to learn from our mistakes and near misses, and those of our fellow utilities around the world: so that as an industry we do not repeat the same mistakes. The very future of our industry will depend on how well each one of us accomplishes this important ask. This paper describes in detail the challenges encountered by one station when incorporating a comprehensive 'Operating Program'. It begins with the Corporate Office's directives to its stations for such a program; and follows up with the details of the actual station implementation of the program, and day to day operating experiences. The paper describes in detail the following Operating Experience programs: - Root Cause Determination process. The Institute of Nuclear Power Operations, Human Performance Enhancement System (HPES) as an integral component of the Root Cause process. Finding solutions for our station for problems identified elsewhere is covered herein; - Significant Event Recommendation Tracking System: - Commitment Tracking System; - Operating Experience (Sharing Lessons Learned) System. The paper will show all the above processes tie closely together and complement each other. The paper discusses the staff required for such processes and their training requirements. It recommends process time lines, reporting mechanisms, and sign off requirements. It will describe the equipment utilized to carry out this work effectively, and with a minimum of staff. One unique feature of the Bruce 'A' system is an 'Effectiveness Follow-Up', usually three to six months after the event recommendations have been completed. By rechecking the finished actions and reviewing them with the personnel involved with the originating event we ensure that the real root causes have been identified and resolved. (author)

  9. Optimizing the use of operating experience at Ontario Hydro's Bruce Nuclear Generating Station 'A'

    Energy Technology Data Exchange (ETDEWEB)

    Williams, E L [Operating Experience Reactor Safety, Bruce Nuclear Generating Station ' A' , Ontario Hydro, Tiverton, Ontario (Canada)

    1991-04-01

    One of the most significant lessons learned from the Three Mile Island event (March 1979), and again with the Chernobyl disaster - (April 1986) was the ongoing requirement to learn from our mistakes and near misses, and those of our fellow utilities around the world: so that as an industry we do not repeat the same mistakes. The very future of our industry will depend on how well each one of us accomplishes this important ask. This paper describes in detail the challenges encountered by one station when incorporating a comprehensive 'Operating Program'. It begins with the Corporate Office's directives to its stations for such a program; and follows up with the details of the actual station implementation of the program, and day to day operating experiences. The paper describes in detail the following Operating Experience programs: - Root Cause Determination process. The Institute of Nuclear Power Operations, Human Performance Enhancement System (HPES) as an integral component of the Root Cause process. Finding solutions for our station for problems identified elsewhere is covered herein; - Significant Event Recommendation Tracking System: - Commitment Tracking System; - Operating Experience (Sharing Lessons Learned) System. The paper will show all the above processes tie closely together and complement each other. The paper discusses the staff required for such processes and their training requirements. It recommends process time lines, reporting mechanisms, and sign off requirements. It will describe the equipment utilized to carry out this work effectively, and with a minimum of staff. One unique feature of the Bruce 'A' system is an 'Effectiveness Follow-Up', usually three to six months after the event recommendations have been completed. By rechecking the finished actions and reviewing them with the personnel involved with the originating event we ensure that the real root causes have been identified and resolved. (author)

  10. Planning for decommissioning of Ignalina Nuclear Power Plant Unit-1

    International Nuclear Information System (INIS)

    Poskas, P.; Poskas, R.; Zujus, R.

    2002-01-01

    In accordance to Ignalina NPP Unit 1 Closure Law, the Government of Lithuania approved the Ignalina NPP Unit 1 Decommissioning Program until 2005. For enforcement of this program, the plan of measures for implementation of the program was prepared and approved by the Minister of Economy. The plan consists of two parts, namely technical- environmental and social-economic. Technical-environmental measures are mostly oriented to the safe management of spent nuclear fuel and operational radioactive waste stored at the plant and preparation of licensing documents for Unit 1 decommissioning. Social-economic measures are oriented to mitigate the negative social and economic impact on Lithuania, inhabitants of the region, and, particularly, on the staff of Ignalina NPP by means of creating favorable conditions for a balanced social and economic development of the region. In this paper analysis of planned radioactive waste management technologies, licensing documents for decommissioning, other technical-environmental and also social-economic measures is presented. Specific conditions in Lithuania important for defining the decommissioning strategy are highlighted. (author)

  11. Design and construction features of steam generators at a nuclear power station

    International Nuclear Information System (INIS)

    Chakrabarti, A.K.; Gupta, K.N.; Bapat, C.N.; Sharma, V.K.

    1996-01-01

    The Indian nuclear power programme is based on Pressurised Heavy Water Reactors (PHWRs) using natural uranium as fuel and heavy water as reactor coolant as well as moderator. The nuclear heat is generated in the fuel located in the pressure tubes. Pressurised heavy water in the primary heat transport (PHT) system is circulated through the tubes which picks up the heat from the fuel and transfers it to ordinary water in steam generators (SGs) to produce steam. The steam is used for providing power to the turbine. The steam generator is a critical equipment in the nuclear steam supply system (NSSS) of a nuclear reactor. SG tube surface area constitute about 80% of total primary circuit surface area. A typical value in a 220 MWe reactor is 9000 m 2 which can release considerable amount of corrosion products unless very low corrosion rates are achieved by proper design, material selection and water chemistry control. Design and construction features of SGs are given. 1 tab

  12. Maintenance and life assessment of steam generators at Embalse Nuclear Station

    International Nuclear Information System (INIS)

    Luna, P.; Diaz, G.; Sveruga, H.; Sainz, R.

    2006-01-01

    The Embalse Nuclear Generating Station (ENGS) has four vertical I-800 U-tubes Steam Generators (SGs) manufactured by Babcock and Wilcox (B and W). They are one of the most important components from the point of view of safety and cost-related elements for potential life extensions in case of a replacement thereof. A Life Management program has been started covering the entire plant and starting with the Life Assessment (LA) of this component which consists in a systematic way to evaluate aging mechanisms focused on the plant refurbishment and life extension. Because of this, maintenance-based ageing assessment from beginning of operation is analyzed and current LA-frame maintenance and inspections programs are carried out in order to maintain a high availability of the SGs then to enable the planning for the plant life extension. The most important taken actions have been the Eddy Current (EC) In Service Inspection program which performs 100% of the tubes of two SG every 1.5 years started in 1992, the mechanical cleaning by blasting of the internal tube surface, the sludge removal from the secondary side tubesheet, the divider plate replacement, the installation of antivibration bars (AVB's), installation of TSP inspection ports and an exhaustive inspection of the secondary internals as a preliminary result of the Life Assessment started during early 2000. The most relevant aging mechanism up to 2004 was the Flow Accelerated Corrosion (FAC) of U-bend supports and consequent fretting of tubes. The eddy current inspections allowed the fretting degradation to be detected and mitigated by installing AVB's. Currently, efficiency of this mitigating action is being performed by vibration measurements and visual inspections. However, other degradation mechanism that could have origin due to the U-bend FAC like loose part damage (LPD) is being to be analyzed since could be an issue in the future. At present, FAC degradation on the cold leg side and sludge deposition on the

  13. Examination of failed studs from No. 2 steam generator at the Maine Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Czajkowski, C.

    1983-02-01

    Three studs removed from service on the primary manway cover from steam generator No. 2 of the Maine Yankee station were sent to Brookhaven National Laboratory (BNL) for examination. The examination consisted of visual/dye penetrant examination, optical metallography and Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) evaluation. One bolt was through cracked and its fracture face was generally transgranular in nature with numerous secondary intergranular cracks. The report concludes that the environmenally assisted cracking of the stud was due to the interaction of the various lubricants used with steam leaks associated with this manway cover

  14. Quad-Cities Station, Units 1 and 2. Annual report, 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated by Unit 1 was 2,246,757 MWh(e) with the generator on line 4,287.5 hrs while Unit 2 generated 1,729,147 MWh(e) with the generator on line 3,056.21 hrs. Information is presented concerning operations, power generation, shutdowns, maintenance, changes, tests, and experiments

  15. Quad-Cities Station, Units 1 and 2. Annual operating report for 1976

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated by Unit 1 was 3,393,062 MWH with the generator on line 5,703.0 hrs. Unit 2 generated 4,304,684 MWH with the generator on line 7,145.3 hrs. Information is presented concerning modifications, maintenance, power generation, shutdowns, occupational radiation exposures, and organization

  16. Use of mock-up training to reduce personnel exposure at the North Anna Unit 1 Steam Generator Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Henry, H.G. [Virginia Power, Mineral, VA (United States); Reilly, B.P. [Bechtel Power Corp., Gaithersburg, MD (United States)

    1995-03-01

    The North Anna Power Station is located on the southern shore of Lake Anna in Louisa County, approximately forty miles northwest of Richmond, Virginia. The two 910 Mw nuclear units located on this site are owned by Virginia Electric and Power Company (Virginia Power) and Old Dominion Electric Cooperative and operated by Virginia Power. Fuel was loaded into Unit 1 in December 1977, and it began commercial operation in June 1978. Fuel was loaded into Unit 2 in April 1980 and began commercial operation in December 1980. Each nuclear unit includes a three-coolant-loop pressurized light water reactor nuclear steam supply system that was furnished by Westinghouse Electric Corporation. Included within each system were three Westinghouse Model 51 steam generators with alloy 600, mill-annealed tubing material. Over the years of operation of Unit 1, various corrosion-related phenomena had occurred that affected the steam generators tubing and degraded their ability to fulfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators tubing and degraded their ability to fullfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators would not last their design life and must be repaired. To this end Virginia Power determined that a steam generator replacement (SGR) program was necessary to remove the old steam generator tube bundles and lower shell sections, including the channel heads (collectively called the lower assemblies), and replace them with new lower assemblies incorporating design features that will prevent the degradation problems that the old steam generators had experienced.

  17. ATWS analysis for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    Dallman, R.J.; Jouse, W.C.

    1985-01-01

    Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of the primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4 0 F/min

  18. Field testing of behavioral barriers for cooling water intake structures -test site 1 - Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Patrick, P.H.; McKinley, R.S.; Micheletti, W.C.

    1988-01-01

    A multi-year research program was developed by the Electric Power Research Institute to evaluate the effectiveness of selected behavioral systems for fish exclusion at sites representative of different aquatic environments. The first test site was the Pickering Nuclear Generating Station (NGS) located on Lake Ontario which represented the Great Lakes environment. A single pneumatic popper, a low frequency, high amplitude sound deterrent, was found to effectively exclude adult alewife, the principal species impinged at Pickering NGS. An air bubble curtain, used either alone or combined with strobe lights, was not a consistent deterrent. Effectiveness of air bubbles was only enhanced when used in association with a popper. Strobe lights were the least effective of the three devices tested. Operation of all three devices together did not surpass the effectiveness of the popper when used alone. Sound deterrents show promise for fish exclusion at generating stations located on the Great Lakes

  19. IEEE standard for type test of class 1E electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    The Institute of Electrical and Electronics Engineers has generated this document to provide guidance for developing a program to type test cables, field splices, and connections and obtain specific type test data. It supplements IEEE Std 323-1974 Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations, which describes basic requirements for equipment qualification. It is the integrated performance of the structures, fluid systems, the electrical systems, the instrumentation systems of the station, and in particular, the plant protection system, that limits the consequences of accidents. Seismic effects on installed cable systems are not within the scope of this document. Section 2 of this guide is an example of type tests. It is the purpose of this guide to deal with cable and connections; however, at the time of issue, detailed examples of tests for connections were not available

  20. Steam generator materials and secondary side water chemistry in nuclear power stations

    International Nuclear Information System (INIS)

    Rudelli, M.D.

    1979-04-01

    The main purpose of this work is to summarize the European and North American experiences regarding the materials used for the construction of the steam generators and their relative corrosion resistance considering the water chemestry control method. Reasons underlying decision for the adoption of Incoloy 800 as the material for the secondary steam generator system for Atucha I Nuclear Power Plant (Atucha Reactor) and Embalse de Rio III Nuclear Power Plant (Cordoba Reactor) are pointed out. Backup information taken into consideration for the decision of utilizing the All Volatil Treatment for the water chemistry control of the Cordoba Reactor is detailed. Also all the reasonswhich justify to continue with the congruent fosfatic method for the Atucha Reactor are analyzed. Some investigation objectives which would eventually permit the revision of the decisions taken on these subjects are proposed. (E.A.C.) [es

  1. Aerial radiological survey of the creeks and tributaries near the Rancho Seco Nuclear Generation Station, Clay Station, California. Date of survey: December 1984

    International Nuclear Information System (INIS)

    1986-06-01

    Radiological contamination due to man-made radionuclides was detected using hand-held instruments in the summer of 1984 in the creeks and tributaries near the Rancho Seco Nuclear Generating Station at Clay Station, California. To help determine the extent of the contamination an aerial radiological survey centered over the creeks and tributaries and including the Rancho Seco facility was conducted during the period 3 to 15 December 1984. Radiological contaminants were detected along a 9-mile segment of the system of creeks in the area. These contaminants included cesium-134, cesium-137, and cobalt-60. Radiation measurements away from the contaminated areas were the same as those made during the aerial radiological survey conducted in 1980

  2. Final environmental statement related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2: (Docket Nos. 50-445 and 50-446)

    International Nuclear Information System (INIS)

    1981-09-01

    The proposed action is the issuance of operating licenses to the Texas Utilities Generating Company for the startup and operation of Units 1 and 2 of the Comanche Peak Steam Electric Station located on Squaw Creek Reservoir in Somervell County, Texas, about 7 km north-northeast of Glen Rose, Texas, and about 65 km southwest of Fort Worth in north-central Texas. The information in this environmental statement represents the second assessment of the environmental impact associated with the Comanche Peak Steam Electric Station pursuant to the guidelines of the National Environmental Policy Act of 1969 (NEPA) and 10 CFR Part 51 of the Commission's Regulations. After receiving an application to construct this station, the staff carried out a review of impact that would occur during its construction and operation. This evaluation was issued as a Final Environmental Statement -- Construction Phase. After this environmental review, a safety review, an evaluation by the Advisory Committee on Reactor Safeguards, and public hearings in Glen Rose, Texas, the US Atomic Energy Commission (now US Nuclear Regulatory Commission) issued construction permits for the construction of Units 1 and 2 of the Comanche Peak Steam Electric Station. 16 figs., 34 tabs

  3. New Jersey's experience with implementing Protective Action Guides during the 1988 Salem ingestion pathway exercise

    International Nuclear Information System (INIS)

    White, Duncan

    1989-01-01

    On November 30 and December 1, 1988, the New Jersey Department of Environmental Protection (DEP) and three other State agencies (Health, Agriculture and State Police) participated in the ingestion pathway portion of the 1988 Salem Nuclear Generating Station Emergency Exercise. The purpose of this phase of the exercise was to demonstrate the ingestion pathway components of the State's Radiological Emergency Response Plan (RERP) to the Federal Emergency Management Agency (FEMA). The intent of this paper is to provide a summary of difficulties and some lessons learned in implementing the DEP's ingestion pathway Protective Action Guides (PAGs) during the exercise as well as during the preparation of a total population dose estimate (TPDE)

  4. New Jersey's experience with implementing Protective Action Guides during the 1988 Salem ingestion pathway exercise

    Energy Technology Data Exchange (ETDEWEB)

    White, Duncan [New Jersey Department of Environmental Protection, Trenton, NJ (United States)

    1989-09-01

    On November 30 and December 1, 1988, the New Jersey Department of Environmental Protection (DEP) and three other State agencies (Health, Agriculture and State Police) participated in the ingestion pathway portion of the 1988 Salem Nuclear Generating Station Emergency Exercise. The purpose of this phase of the exercise was to demonstrate the ingestion pathway components of the State's Radiological Emergency Response Plan (RERP) to the Federal Emergency Management Agency (FEMA). The intent of this paper is to provide a summary of difficulties and some lessons learned in implementing the DEP's ingestion pathway Protective Action Guides (PAGs) during the exercise as well as during the preparation of a total population dose estimate (TPDE)

  5. Quad-Cities Station, Units 1 and 2. Semiannual operating report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Unit 1 generated 2,024,125 net electrical MWH and the generator was on line 3162.6 hours. Unit 2 generated 746,184 net electrical MWH and was on line 1475.3 hrs. Data is included concerning operations, power generation, shutdowns, maintenance, changes, and tests. (FS)

  6. IEEE Std 650-1990: IEEE standard for qualification of Class 1E static battery chargers and inverters for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Methods for qualifying static battery chargers and inverters for Class 1E installations in a mild environment outside containment in nuclear power generating stations are described. The qualification methods set forth employ a combination of type testing and analysis, the latter including a justification of methods, theories, and assumptions used. These procedures meet the requirements of IEEE Std 323-1983, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations

  7. Socioeconomic impacts of nuclear generating stations: Arkansas Nuclear One station case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Arkansas Nuclear One nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  8. Remote multiplexing system application to a nuclear generating station--An update

    International Nuclear Information System (INIS)

    Mazzoni, O.S.; Cava, A.L.; Bijoor, G.D.; Wiitala, M.F.

    1979-01-01

    Washington Public Power Supply System's Nuclear Power Plant WNP-2 is a 1200 MW, boiling water reactor (BWR) nuclear generating facility now under construction at Hanford, Washington. An early study indicated that substantial savings in cost could be obtained by the substitution of the Remote Multiplexing System (RMS) for the control and signal wiring between the Control Building and the outside facilities, which required long cable runs. It is a high speed, data system providing time-sharing digital communication. The purpose of this paper is to give the details of the RMS used in the WNP-2 project. It discusses the qualification requirements for Class IE systems and includes the description of the tests on the system

  9. The costs of generating electricity in nuclear and coal fired power stations

    International Nuclear Information System (INIS)

    Petroll, M.

    1984-01-01

    An ad-hoc group of experts for international comparison of electricity generation cost was established in the OECD more than two years ago. This group of experts submitted their report of results in English at the end of last year. This paper publishes an abbreviated version making use of original quotations exclusively in order to keep true to the content of the study as much as possible. The study arrives at the following conclusion: ''There is no uniform set of input data for nuclear and coalfired power plants and assumptions concerning the base parameters of the reactor differ from country to country. Despite these differences, the outcome is that, nuclear energy is cheaper than coal in all countries concerned with the exception of some parts of the United States and Canada.'' (orig./UA) [de

  10. Design and verification of computer-based reactor control system modification at Bruce-A candu nuclear generating station

    International Nuclear Information System (INIS)

    Basu, S.; Webb, N.

    1995-01-01

    The Reactor Control System at Bruce-A Nuclear Generating Station is going through some design modifications, which involve a rigorous design process including independent verification and validation. The design modification includes changes to the control logic, alarms and annunciation, hardware and software. The design (and verification) process includes design plan, design requirements, hardware and software specifications, hardware and software design, testing, technical review, safety evaluation, reliability analysis, failure mode and effect analysis, environmental qualification, seismic qualification, software quality assurance, system validation, documentation update, configuration management, and final acceptance. (7 figs.)

  11. Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro's Bruce Nuclear Generating Station 'A'

    International Nuclear Information System (INIS)

    Day, J.E.; Baker, R.L.

    1995-01-01

    Ontario Hydro at the Bruce Nuclear Generating Station 'A' has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

  12. Projected costs of generating electricity from nuclear and coal-fired power stations for commissioning in 1995

    International Nuclear Information System (INIS)

    1986-01-01

    This report updates and extends the previous NEA study, ''The Costs of Generating Electricity in Nuclear and Coal-fired Power Stations'', published by the OECD in late 1983. Despite the changed expectations concerning coal prices and the considerable movements in exchange rates since the first study was completed, the conclusions remain essentially the same. Nuclear Power is projected to be economically superior by a significant margin to coal-fired plants for base load electricity production in Europe, Japan and some regions of North America. In areas of North America in close proximity to supplies of cheap coal, this would be the more economic fuel, unless future nuclear investment costs can be reduced to match the best US and Canadian experience. In all regions considered, the economic advantage of both coal and nuclear over oil and gas-fired plants for commissioning in the mid-1990s is expected to be substantial. These conclusions are based on an analysis of cost projections for 900 MWe to 1400 MWe Light Water Reactors to be commissioned in 1995, operating at a levelised load factor of about 72 per cent over an assumed 25 years economic life and calculated with a 5 per cent (real) discount rate. This parallels the reference reactor selected for the NEA report ''The Economics of the Nuclear Fuel Cycle'', which was published by the OECD in June 1985, though it deviates somewhat from the reference conditions of the previous generation cost study. Contemporary coal-fired stations ranging in capacity from 330 MWe to 700 MWe with the same assumed economic life and load factor provide the basis for comparison. Some data are included on CANDU Pressurised Heavy Water Reactors, and a brief comment is annexed on the relevance of the comparisons for the smaller plants that may be of interest to countries with smaller electricity networks or where special circumstances apply

  13. Aseismatic design and safety of nuclear power generation facilities. On aseismatic capability of commercial nuclear power stations

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1995-01-01

    In view of the great Hanshin earthquake, the aseismatic safety of the important facilities in nuclear power stations is ensured by the location direct on base rocks, the design with the earthquake force at least three times as large as that in the building standard, and the consideration of the earthquakes due to active faults as design earthquake. The basic policy of the aseismatic design of nuclear power stations is described. The determination of the earthquake motions due to strongest earthquake and utmost limit earthquake for design, the survey of the geological features and ground of the sites and so on are explained. In the aseismatic design of buildings and structures, structural planning, the modeling for the aseismatic analysis of buildings, the analysis of time historical response and so on are carried out. In the aseismatic design of equipment and piping systems, the planning of aseismatic support structures, the aseismatic design and the analysis of time historical response, the spectral modal analysis for other systems such as multiple material point system and so on are described. The tests and researches related to the aseismatic design are reported. (K.I.)

  14. The Paks Nuclear Power Station

    International Nuclear Information System (INIS)

    Erdosi, N.; Szabo, L.

    1978-01-01

    As the first stage in the construction of the Paks Nuclear Power Station, two units of 440 MW(e) each will be built. They are operated with two coolant loops each. The reactor units are VVER 440 type water-moderated PWR type heterogeneous power reactors designed in the Soviet Union and manufactured in Czechoslovakia. Each unit operates two Soviet-made K-220-44 steam turbines and Hungarian-made generators of an effective output of 220 MW. The output of the transformer units - also of Hungarian made - is 270 MVA. The radiation protection system of the nuclear power station is described. Protection against system failures is accomplished by specially designed equipment and security measures especially within the primary circuit. Some data on the power station under construction are given. (R.P.)

  15. Tube structural integrity evaluation of Palo Verde Unit 1 steam generators for axial upper-bundle cracking

    International Nuclear Information System (INIS)

    Woodman, B.W.; Begley, J.A.; Brown, S.D.; Sweeney, K.; Radspinner, M.; Melton, M.

    1995-01-01

    The analysis of the issue of upper bundle axial ODSCC as it apples to steam generator tube structural integrity in Unit 1 at the Palo Verde Nuclear generating Station is presented in this study. Based on past inspection results for Units 2 and 3 at Palo Verde, the detection of secondary side stress corrosion cracks in the upper bundle region of Unit 1 may occur at some future date. The following discussion provides a description and analysis of the probability of axial ODSCC in Unit 1 leading to the exceedance of Regulatory Guide 1.121 structural limits. The probabilities of structural limit exceedance are estimated as function of run time using a conservative approach. The chosen approach models the historical development of cracks, crack growth, detection of cracks and subsequent removal from service and the initiation and growth of new cracks during a given cycle of operation. Past performance of all Palo Verde Units as well as the historical performance of other steam generators was considered in the development of cracking statistics for application to Unit 1. Data in the literature and Unit 2 pulled tube examination results were used to construct probability of detection curves for the detection of axial IGSCC/IGA using an MRPC (multi-frequency rotating panake coil) eddy current probe. Crack growth rates were estimated from Unit 2 eddy current inspection data combined with pulled tube examination results and data in the literature. A Monte-Carlo probabilistic model is developed to provide an overall assessment of the risk of Regulatory Guide exceedance during plant operation

  16. Environmental assessment, proposed generating station for Darlington

    International Nuclear Information System (INIS)

    1975-04-01

    This document indicates the intention of Ontario Hydro to seek approval from the Provincial Government for its plan to construct and operate a 3400 MWe nuclear generating station at the Darlington site, west of Bowmanville. This preliminary proposal also contains the environmental assessment. The environmental section of this proposal describes and assesses the existing environment and the environmental influences which would occur due to the construction and operation of a nuclear generating station, consisting of four 850 MW units, at the Darlington site. This proposed station is similar to the Bruce GS A station presently under construction. (author)

  17. Eddy current magnetic bias x-probe qualification and inspection of steam generator Monel 400 tubing in Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Lepine, B.A.; Van Langen, J.; Obrutsky, L.

    2006-01-01

    This paper presents an overview of the x-probe MB 350 eddy current inspection array probe, for detection of open OD axial crack-like flaws in Monel 400 tubes at Pickering Nuclear Generating Station. This report contains a selection of inspection results from the field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, specially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential out diameter (OD) axially oriented crack. Post-inspection tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability μ r Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

  18. Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

    International Nuclear Information System (INIS)

    Khalil, Y.F.

    1993-01-01

    Generic Letter 88-20, Supplement No. 1 (Ref. 1), issued by the Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits for nuclear power reactor facilities to perform Individual Plant Examinations (IPE) of their plant(s) for severe accident vulnerabilities and to submit the results to the Commission. This paper summarizes the major Front-End (Level-1 PRA) and Back-End (Level-2 PRA) insights gained from the Millstone Unit 1 (MP-1) IPE study. No major plant vulnerabilities have been identified from a Front-End perspective. The Back-End analysis, however, has identified two potential containment vulnerabilities during postulated events that progress beyond the Design Basis Accidents (DBAs), namely, (1) MP-1 is dominated by early source term releases that would occur within a six-hour time frame from time of accident initiation, or reactor trip, and (2) MP-1 containment is somewhat vulnerable to leak-type failure through the drywell head. As a result of the second finding, a recommendation currently under evaluation, has been made to increase the drywell head bolt's preload from 54 Kips to resist the containment design pressure value (62 psig)

  19. Decommissioning of nuclear power stations

    International Nuclear Information System (INIS)

    Gregory, A.R.

    1988-01-01

    In the United Kingdom the Electricity Boards, the United Kingdom Atomic Energy Authority (UKAEA) and BNFL cooperate on all matters relating to the decommissioning of nuclear plant. The Central Electricity Generating Board's (CEGB) policy endorses the continuing need for nuclear power, the principle of reusing existing sites where possible and the building up of sufficient funds during the operating life of a nuclear power station to meet the cost of its complete clearance in the future. The safety of the plant is the responsibility of the licensee even in the decommissioning phase. The CEGB has carried out decommissioning studies on Magnox stations in general and Bradwell and Berkeley in particular. It has also been involved in the UKAEA Windscale AGR decommissioning programme. The options as to which stage to decommission to are considered. Methods, costs and waste management are also considered. (U.K.)

  20. Floating nuclear power station of APWS-80 type for electricity generation and fresh water production

    International Nuclear Information System (INIS)

    Zverev, K.V.; Polunichev, V.I.; Sergeev, Yu.A.

    1997-01-01

    To solve the problem of seawater desalination and electric energy generation, the designing organizations of Russia have developed two variants of floating nuclear desalination plant. The KLT-40 type reactors, with maximum 160 MW thermal power, is used as the power source for such plant. Depending on the customer requirement one or two power unit could be installed in the floating desalination plant. There are APWS-80 with two reactors, producing 80,000 m 3 desalinated water per day and APWS-40 with one reactor, producing 40,000 m 3 desalinated water per day. The advantages of floating desalination plants are the possibility to build and test them at the ship-build plant of the supplier country and to hand them over on turnkey base. (author). 5 figs

  1. Black Fox Station, Units 1 and 2. Application for construction permits and operating licenses

    International Nuclear Information System (INIS)

    1975-01-01

    An application to construct and operate Black Fox Station, Units 1 and 2, is presented. The two BWR type reactors will have a rated core thermal power of 3579 MW(t) and a net electrical power of approximately 1150 MW(e). The facility will be located in Inola Township, 23 miles east of Tulsa on the east side of the Verdigris River in Rogers County, Oklahoma

  2. Environmental radiological studies downstream from the Rancho Seco Nuclear Power Generating Station, 1985

    International Nuclear Information System (INIS)

    Noshkin, V.E.; Wong, K.M.; Eagle, R.J.; Brunk, J.L.; Jokela, T.A.

    1986-01-01

    Information compiled in 1985 while assessing the environmental impact of radionuclides previously discharged with aqueous releases from the Rancho Seco Nuclear Power Generating Plant is presented. In October 1984, the quantities of gamma-emitting radionuclides in water discharged to Clay Creek from the plant were reduced below operationally defined detection limits for liquid effluents. However, radionuclides previously discharged persist in the downstream environment and are found in many aquatic dietary components. 134 Cs and 137 Cs are the primary gamma-emitting radionuclides detected in the edible flesh of different fish, crayfish, and frogs. Coefficients for exponential equations are generated, from a least square analysis, that relate the change in concentration of 137 Cs in fish to distance downstream and time between March and October 1985. Concentrations of 137 Cs in surface creek sediments also decreased in the downstream direction much in the same manner as concentrations decreased in fish. However, there was no significant difference in the radiocesium concentrations in surface sediements collected from comparable locations during both 1984 and 1985

  3. Development of the MARS input model for Kori nuclear units 1 transient analyzer

    International Nuclear Information System (INIS)

    Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.

    2004-11-01

    KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1

  4. Training of system engineers for Salem and Hope Creek nuclear power plant personnel

    International Nuclear Information System (INIS)

    Ketcham, S.

    1988-01-01

    To establish and maintain a high level of plant reliability at a nuclear power station, a conscientious, integrated, day-to-day effort is required from operations and maintenance personnel, as well as the engineering support groups. The catalyst and focus to ensure that this occurs within the Public Service Electric and Gas Company system is the station system engineer. These engineers have total responsibility for their designated systems. For these individuals to communicate with the operations and maintenance departments effectively, they must have an understanding of the design and operation of each and every system and the effect they may have on the overall plant. (author)

  5. Local society and nuclear power stations

    International Nuclear Information System (INIS)

    1984-02-01

    This report was made by the expert committee on region investigation, Japan Atomic Industrial Forum Inc., in fiscal years 1981 and 1982 in order to grasp the social economic influence exerted on regions by the location of nuclear power stations and the actual state of the change due to it, and to search for the way the promotion of local community should be. The influence and the effect were measured in the regions around the Fukushima No. 1 Nuclear Power Station of Tokyo Electric Power Co., Inc., the Mihama Power Station of Kansai Electric Power Co., Inc., and the Genkai Nuclear Power Station of Kyushu Electric Power Co., Inc. The fundamental recognition in this discussion, the policy of locating nuclear power stations and the management of regions, the viewpoint and way of thinking in the investigation of the regions where nuclear power stations are located, the actual state of social economic impact due to the location of nuclear power stations, the connected mechanism accompanying the location of nuclear power stations, and the location of nuclear power stations and the acceleration of planning for regional promotion are reported. In order to economically generate electric power, the rationalization in the location of nuclear power stations is necessary, and the concrete concept of building up local community must be decided. (Kako, I.)

  6. Pumps for nuclear power stations

    International Nuclear Information System (INIS)

    Ogura, Shiro

    1979-01-01

    16 nuclear power plants are in commercial operation in Japan, and nuclear power generation holds the most important position among various substitute energies. Hereafter also, it is expected that the construction of nuclear power stations will continue because other advantageous energy sources are not found. In this paper, the outline of the pumps used for BWR plants is described. Nuclear power stations tend to be large scale to reduce the construction cost per unit power output, therefore the pumps used are those of large capacity. The conditions to be taken in consideration are high temperature, high pressure, radioactive fluids, high reliability, hydrodynamic performances, aseismatic design, relevant laws and regulations, and quality assurance. Pumps are used for reactor recirculation system, control rod driving hydraulic system, boric acid solution injecting system, reactor coolant purifying system, fuel pool cooling and purifying system, residual heat removing system, low pressure and high pressure core spraying systems, and reactor isolation cooling system, for condensate, feed water, drain and circulating water systems of turbines, for fresh water, sea water, make-up water and fire fighting services, and for radioactive waste treating system. The problems of the pumps used for nuclear power stations are described, for example, the requirement of high reliability, the measures to radioactivity and the aseismatic design. (Kako, I.)

  7. US central station nuclear electric generating units: significant milestones (status as of January 1, 1979)

    International Nuclear Information System (INIS)

    1979-03-01

    Construction and operational milestones are tabulated for US nuclear power plants. Data are presented on nuclear steam supply system orders. A schedule of commercial operation and projected capital costs through 1990 is given

  8. Reload Startup Physics Tests for Tianwan Nuclear Power station

    International Nuclear Information System (INIS)

    Yang Xiaoqiang; Li Wenshuang; Li Youyi; Yao Jinguo; Li Zaipeng Jiangsu

    2010-01-01

    This paper briefly describes the test purposes, test items, test schedules and test equipment's for reload startup physics test's on Unit 1 and 2 of Tianwan Nuclear Power station. Then, an overview of the previous thrice tests and evaluations on the tests results are presented. In the end, the paper shows the development and work direction of optimization project for reload startup physics tests on Unit 1 and 2 of Tianwan Nuclear Power station. (Authors)

  9. Impact of offshore nuclear generating stations on recreational behavior at adjacent coastal sites. Technical report

    International Nuclear Information System (INIS)

    Baker, E.J.; Moss, D.J.; West, S.G.; Weyant, J.K.

    1977-12-01

    A multi-faceted investigation was undertaken to project the impact of offshore nuclear power plants on beach visitation at adjacent beaches. 1. Related literature was reviewed concerning human adjustment to natural hazards, risk-taking behavior, and public attitudes toward nuclear power. 2. Approximately 2400 people were interviewed at beaches in three states with respect to: (a) intended avoidance of beaches near a hypothetical floating nuclear plant (FNP), (b) relative importance of proximity to a FNP, when compared to other beach attributes, (c) onshore-offshore preference for coastal nuclear plant location, (d) behavioral impact of NRC licensing of FNP's, (e) relative tourism impact of coastal nuclear plant compared to coastal coal-fired plant, (f) public concerns about nuclear safety, (g) public attitudes toward alternative energy sources, (h) public confidence in sources of information about nuclear power, (i) visual impact of a FNP, and (j) knowledge about nuclear power. 3. Four beach areas near currently operating coastal nuclear power plants were studied to assess impacts on tourism resulting from plant construction. Data suggest that proximity of a FNP is less important than other beach attributes in determining beach attractiveness, probably no more than (and perhaps less than) 5% to 10% of current beach patrons would avoid a beach after FNP siting three miles directly offshore, and impact of a FNP would decrease exponentially as distance away increased

  10. Impact of offshore nuclear generating stations on recreational behavior at adjacent coastal sites

    International Nuclear Information System (INIS)

    Baker, E.J.; West, S.G.; Moss, D.J.; Weyant, J.K.

    1977-10-01

    A multi-faceted investigation was undertaken to project the impact of offshore nuclear power plants on beach visitation at adjacent beaches. Related literature was reviewed concerning human adjustment to natural hazards, risk-taking behavior, and public attitudes toward nuclear power. Approximately 2400 people were interviewed at beaches in three states with respect to: intended avoidance of beaches near a hypothetical floating nuclear plant (FNP), relative importance of proximity to a FNP, when compared to other beach attributes, onshore-offshore preference for coastal nuclear plant location, behavioral impact of NRC licensing of FNPs, relative tourism impact of coastal nuclear plant compared to coastal coal-fired plant, public concerns about nuclear safety, public attitudes toward alternative energy sources, public confidence in sources of information about nuclear power, visual impact of a FNP, and knowledge about nuclear power. Four beach areas near currently operating coastal nuclear power plants were studied to assess impacts on tourism resulting from plant construction. Data suggest that proximity of a FNP is less important than other beach attributes in determining beach attractiveness, probably no more than (and perhaps less than) 5% to 10% of current beach patrons would avoid a beach after FNP siting three miles directly offshore, and impact of a FNP would decrease exponentially as distance away increased

  11. US central station nuclear electric generating units: significant milestones (status as of July 1, 1980)

    International Nuclear Information System (INIS)

    1980-09-01

    The status of 189 US nuclear power plants is reported in a table which gives the name, owner, capacity, type, NSSS architect and contractor and data of public announcement, NSSS order, licensing, and initial operation. The plants are also indexed according to state, region, utility, and alphabetical name. The utility nuclear steam supply system orders are also listed

  12. Science, Society, and America's Nuclear Waste: Nuclear Waste, Unit 1. Teacher Guide. Second Edition.

    Science.gov (United States)

    Department of Energy, Washington, DC. Office of Civilian Radioactive Waste Management, Washington, DC.

    This guide is Unit 1 of the four-part series Science, Society, and America's Nuclear Waste produced by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management. The goal of this unit is to help students establish the relevance of the topic of nuclear waste to their everyday lives and activities. Particular attention is…

  13. IEEE Std 323-1983: IEEE standard for qualifying Class 1E equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard describes the basic requirements for qualifying Class 1E equipment with interfaces that are to be used in nuclear power generating stations. The requirements presented include the principles, procedures, and methods of qualification. These qualification requirements, when met, will confirm the adequacy of the equipment design under normal, abnormal, design basis event, post design basis event, and in-service test conditions for the performance of safety function(s). The purpose of this standard is to identify requirements for the qualification of Class 1E equipment, including those interfaces whose failure could adversely affect the performance of Class 1E equipment and systems. The methods described shall be used for qualifying equipment, extending qualification, and updating qualification if the equipment is modified. Other issued IEEE standards which present qualification methods for specific equipment or components, or both, and those that deal with parts of the qualification program, may be used to supplement this standard, as applicable

  14. Estimation of the costs of electricity generation in nuclear and other types of power stations

    International Nuclear Information System (INIS)

    Col Buti, P.; Tapia Fernandez, C.

    2002-01-01

    (Study of the Nuclear Power Sector in the Context of the Energy Study in Catalonia to the 2020 Horizon) prepared by the same authors, which analyses the characteristics of the nuclear sector in Catalonia. This an other studies has provided the basis for the preparation of the Pla de l'Energia a Catalunya a l'Horitzo de l'Any 2010 (Catalonia Energy Plan to the 2010 Horizon). (Author)

  15. Considerations for increasing unit 1 spent fuel pool capacity at the Laguna Verde station

    International Nuclear Information System (INIS)

    Vera, A.

    1992-01-01

    To increase the spent fuel storage capacity at the Laguna Verde Station in a safe and economical manner and assure a continuous operation of the first Mexican Nuclear Plant, Comision Federal de Electricidad (CFE), the Nation's Utility, seeked alternatives considering the overall world situation, the safety and licensing aspects, as well as the economics and the extent of the nuclear program of Mexico. This paper describes the alternatives considered, their evaluation and how the decision taken by CFE in this field, provides the Laguna Verde Station with a maximum of 37 years storage capacity plus full core reserve

  16. AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1994

    International Nuclear Information System (INIS)

    1995-06-01

    AECB believe that Ontario Hydro operated Bruce A in a safe manner during 1994, and that the risk to workers and the public has been maintained at an acceptably low level. Radiation doses to workers and releases to the environment were well below regulatory limits. All special safety systems met availability targets. We noted improvements in operation and maintenance but some further improvements are still required. This is particularly true of the station's compliance with the Operating Licence. AECB believe that the station continues to be well managed, with a high priority placed on safety. However, there is a need for increased capability in the area of safety analysis and assessment. 4 tabs., 4 figs

  17. Discharges from nuclear power stations

    International Nuclear Information System (INIS)

    1991-02-01

    HM Inspectorate of Pollution commissioned, with authorising responsibilities in England and Wales, a study into the discharges of radioactive effluents from Nuclear Power Stations. The study considered arisings from nuclear power stations in Europe and the USA and the technologies to treat and control the radioactive discharges. This report contains details of the technologies used at many nuclear power stations to treat and control radioactive discharges and gives, where information was available, details of discharges and authorised discharge limits. (author)

  18. Indian Point Station, Unit 1 and 2. Semiannual operating report No. 24, July--December 1974

    International Nuclear Information System (INIS)

    1975-01-01

    Net electrical power generated by Unit 1 was 519,130 MWH with the reactor critical for 2,400.39 hours and the generator on line for 2,316.14 hours. Unit 2 generated 2,427,828 MWH electrical power, was critical for 3,590.31 hours and the generator was on line for 3,485.41 hours. Operations and maintenance are summarized. Information is presented concerning radioactive effluent releases, occupational personnel radiation protection, primary coolant chemistry, changes, tests, and experiments. Environmental radioactivity is discussed. (U.S.)

  19. EMI environment EMC considerations concerning equipment upgrades at a nuclear utility power generating station a case history

    International Nuclear Information System (INIS)

    Metcalf, M.J.

    1993-01-01

    As equipment is upgraded during scheduled power outages in nuclear power generating stations, more and more utilities will be faced with the problem of dealing with Electromagnetic Compatibility (EMC) issues that have evolved with the trend of manufacturers designing equipment based around state-of-the-art high technology logic devices. This paper lists Commonwealth Edison Company (CECO) equipment that was scrutinized by National Technical Systems (NTS) for its EMI impact. The test requirements and test procedures for assessing EMI and EMC are outlined. Although on-site mapping data was gathered to assist CECO for the upgrade described here of the Auxiliary Electric Equipment Room of a Westinghouse Eagle 21 Reactor Protection System at the Zion Unit No.1 Station, blanket mapping of every location for every upgrade is a short term, expensive solution to the EMI problem. It is concluded that the primary problem of lack of system and component level EMI specifications must be addressed by a governing body. 7 refs., 2 figs

  20. IEEE Std 383-1974: IEEE standard for type test of Class IE electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices. This guide does not cover cables for service within the reactor vessel

  1. 1300MVA steam-turbine generators for Kansai Electric Power's Oi Nuclear Power Station

    Energy Technology Data Exchange (ETDEWEB)

    Oishi, N; Amagasa, N; Ito, H; Yagi, K [Mitsubishi Electric Corp., Kobe (Japan). Kobe Works

    1977-06-01

    Mitsubishi Electric has completed two 1300 MVA generators, equipped with 5500kW brushless exciters, that will be the No. 1 and No. 2 generators of the Oi plant. They are among the largest anywhere, and incorporate such technological innovations as water cooling of the stator coil and asymmetrical arrangement of the rotor slots. The article discusses generator specifications and construction, the brushless exciter, and the results of factory tests.

  2. Services for nuclear power stations

    International Nuclear Information System (INIS)

    Fremann, M.; Ryckelynck

    1987-01-01

    This article gives an information as complete as possible about the activities of the french nuclear industry on the export-market. It describes the equipment and services available in the field of services for nuclear power stations [fr

  3. Revalidation program for nuclear standby diesel generators

    International Nuclear Information System (INIS)

    Muschick, R.P.

    1985-01-01

    This paper describes the program which Duke Power Company carried out to revalidate the diesel engines used in diesel generators for nuclear standby service at Unit 1 of the Catawba Nuclear Station. The diesels operated satisfactorily during the tests, and only relatively minor conditions were noted during the test and inspections, with one exception. This exception was that cracks were detected in the piston skirts. The piston skirts have been replaced with improved design skirts. The diesels have been fully revalidated for their intended service, and have been declared operable

  4. Wet steam turbines for nuclear generating stations -design and operating experience

    International Nuclear Information System (INIS)

    Usher, J.

    1977-01-01

    Lecture to the Institution of Nuclear Engineers, 11 Jan. 1977. The object of this lecture was to give an account of some design features of large wet steam turbines and to show by describing some recent operational experience how their design concepts were fulfilled. Headings are as follows: effects of wet steam cycle on turbine layout and operation (H.P. turbine, L.P. turbine); turbine control and operation; water separators; and steam reheaters. (U.K.)

  5. Torness: proposed nuclear power station

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The need for and desirability of nuclear power, and in particular the proposed nuclear power station at Torness in Scotland, are questioned. Questions are asked, and answered, on the following topics: position, appearance and cost of the proposed Torness plant, and whether necessary; present availability of electricity, and forecast of future needs, in Scotland; energy conservation and alternative energy sources; radiation hazards from nuclear power stations (outside, inside, and in case of an accident); transport of spent fuel from Torness to Windscale; radioactive waste management; possibility of terrorists making a bomb with radioactive fuel from a nuclear power station; cost of electricity from nuclear power; how to stop Torness. (U.K.)

  6. Heat-shrinkable splicing materials for Class 1E wire and cable systems in nuclear power generating stations

    International Nuclear Information System (INIS)

    Handa, Katsue; Maruyama, Masahiro; Kanno, Mikio; Ohya, Shingo; Nagakawa, Seiji; Sugimori, Mikihiro

    1987-01-01

    This report describes the shapes of heat-shrinkable splicing materials (cable sleeve and breakout, and round end cap) made of polyolefine resin, their application to cable splicing, and the properties of the materials as well as of the splice using them. Particularly, the report features introduction of their properties as determined by tests under the same conditions as used in Japan in qualifying tests on wires and cables for nuclear power generating stations. The heat-shrinkable splicing materials proved to be equal in properties to flame-retardant cables for nuclear power plants when tested for oxygen index and subjected to a vertical flame test on ''insulated wire'' and a vertical tray flame test on the cable splice. It was also confirmed that Class 1E cable using these splicing materials could stand the most rigorous environmental test in Japan. Therefore they can be used for splicing Class 1E wires and cables and the splice formed with them can be regarded as Class 1E specified in IEEE Std. 383. (author)

  7. Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455). Supplement No. 7

    International Nuclear Information System (INIS)

    1986-11-01

    Supplement No. 7 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power operation for Unit 2

  8. Safety Evaluation Report, related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

    International Nuclear Information System (INIS)

    1983-11-01

    Supplement No. 3 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  9. Safety evaluation report related to the operation of Byron Station, Units 1 and 2. Docket Nos. STN 50-454 and STN 50-455

    International Nuclear Information System (INIS)

    1983-01-01

    Supplement No. 2 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  10. Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

    International Nuclear Information System (INIS)

    1984-05-01

    Supplement No. 4 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  11. The Spanish Nuclear Safety Council and nuclear power stations in operation in Spain

    International Nuclear Information System (INIS)

    Perello, M.

    1984-01-01

    On 20 April 1980 the Spanish Congress of Deputies passed an Act setting up the Nuclear Safety Council (CSN) as the sole organization responsible for nuclear safety and radiation protection. In this paper it is stated that that date marked the beginning of a new nuclear safety policy in Spain. As one of its objectives, this policy is aimed at the monitoring and testing of operating nuclear installations. A detailed description is given of the Operating Nuclear Installation Service (SINE), including its basic structure, its functions and the technical and manpower resources available to it. The maintenance of close relations with other organs of the CSN is considered of paramount importance in order for the tasks allotted to SINE to be fulfilled. International co-operation and outside contracting greatly assist importing countries which have limited manpower resources. A description is then given of the present state of the nuclear power stations in operation in Spain together with an account of the most important initiatives which have been taken so far. The year 1968 saw the beginning of commercial operation of the Jose Cabrera nuclear power station, which has the only single-loop PWR reactor in the world. At present, it is being subjected to the Systematic Evaluation Programme (SEP). The Santa Maria de Garona nuclear power station has been operating for over twelve years and is also being subjected to the SEP although design modifications derived from operating experience have already been introduced. The Vandellos I station was the last of the first generation and has also benefited from the operating experience of similar French plants. Unit 1 of the Almaraz power station opens the door to the second generation and the generic problem which has occurred with the steam generators is in process of being solved. Lastly, some general conclusions are presented about the organization of and experience acquired with operating nuclear power stations. (author)

  12. The new generation of nuclear power stations. A new trend in atomic power?

    International Nuclear Information System (INIS)

    Hohlefelder, W.

    2006-01-01

    According to the author, all options for future power supply should be followed, including atomic power provided that it can be made technically safe and treated with a maximum safety culture. On the one hand, power supply is an elementary human need, deciding on public welfare, economic development and technical progress. On the other hand, there is an impending shortage of power owing to depletion of resources and the emergence of new industrialized nations especially in south east Asia. For this reason, all options should be considered, from renewable energy sources to coal and nuclear power. (orig.)

  13. Three Mile Island nuclear generating station accident of March 28, 1979

    International Nuclear Information System (INIS)

    Aitken, J.H.; Johnson, A.C.; Kelly, R.J.; Wong, K.Y.

    1979-04-01

    The government of Ontario dispatched a Scientific Assessment Team to the Three Mile Island nuclear power plant to assess the consequences of an accident which occurred at unit 2 on 1979 March 28. The team's objectives were to acquire up-to-the-minute information concerning the accident, study the potential for environmental impacts on Ontario, observe the outcome of off-site emergency procedures, and offer assisance from Ontario should it appear of value. The findings, observations, and impressions of the team are summarized. (OST)

  14. Experiences from Loviisa Nuclear Power Station concerning the decontamination of steam generators and primary system components

    International Nuclear Information System (INIS)

    Jaernstroem, R.

    1989-01-01

    Loviisa 1 and 2 are 465 MWe PWR units of the Soviet type VVER-440. Loviisa 1 has been in commercial operation since spring 1977 and Loviisa 2 from the beginning of 1980. Decontamination of primary circuit components - even big ones as steam generators - can be performed in an efficient and quick way with good results and resonable expences. Total costs for decontamination of the two steam generators including planning, construction, documentation, operation, chemicals etc. did not rise above 100,000.00 dollars. (author) 6 figs., 2 tabs

  15. Nuclear power generating station operability assurance reliability, availability, and maintainability application for maintenance management

    International Nuclear Information System (INIS)

    Cleveland, J.W.; Regenie, T.R.; Wilson, R.J.

    1985-01-01

    Environmental qualification and equipment warrantee insurance stipulations should be supplemented with a reliable maintainability program structured to identify and control fast failing subcomponents within critical equipment. Anticipation of equipment subcomponent failures can control unnecessary plant off-line occurrences. Incorporation of reliability, availability, and maintainability considerations into plant maintenance policies on power generation and safety related items have positive cost benefit advantages

  16. Fort St. Vrain Nuclear Generating Station. Semiannual operations report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    The reactor was in start-up testing during this period. Information is presented concerning fuel and equipment performance; changes in facility design; procedures; testing; staff; licensed operator training; surveillance; power generation; maintenance; shutdowns, primary coolant chemistry; radiation exposure; fission product release; solid wastes; and radioactive effluents

  17. Islands for nuclear power stations

    International Nuclear Information System (INIS)

    Usher, E.F.F.W.; Fraser, A.P.

    1981-01-01

    The safety principles, design criteria and types of artificial island for an offshore nuclear power station are discussed with particular reference to siting adjacent to an industrial island. The paper concludes that the engineering problems are soluble and that offshore nuclear power stations will eventually be built but that much fundamental work is still required. (author)

  18. Aerial radiological survey of the Palo Verde Nuclear Generating Station and surrounding area, Wintersburg, Arizona. Date of survey: November 1982

    International Nuclear Information System (INIS)

    Semmler, R.A.

    1984-01-01

    An aerial survey of terrestrial gamma radiation was performed during the period 4 November through 15 November 1982 over a 16-kilometer by 16-kilometer (10-mile by 10-mile) area approximately centered on the Palo Verde Nuclear Generating Station near Wintersburg, Arizona. Gamma radiation spectral data were collected while flying a helicopter over a regular pattern of parallel lines spaced 150 meters (500 feet) apart at an altitude of 90 meters (300 feet). All radiation measurements taken at the nominal flight altitude were corrected for altitude variations, cosmic radiation, and helicopter background to generate exposure rates from terrestrial sources extrapolated to 1 meter above ground level. The data are presented as isoradiation contour maps. The average terrestrial radiation levels fall between 8 and 14 microrentgens per hour (μR/h). All gamma radiation detected within the survey area was associated with naturally occurring radionuclides. Direct ground-based measurements at 1 meter height were also taken at four scattered sites within the survey area. These values agree with the contour intervals determined from the aerial measurements and differ from the mean value of adjacent contours by no more than 10%. 6 references, 5 figures, 2 tables

  19. Gas and water permeability tests of 25 year old concrete from the NPD Nuclear Generating Station

    International Nuclear Information System (INIS)

    Mills, R.H.

    1990-05-01

    Permeability tests on cores recovered from concrete which had been in service for 25 years in the Nuclear Power Demonstration (NPD) reactor showed rates of mass transfer of gas and water which were greater than young fresh concrete of the same proportions and that reported in previous AECB reports. This transparency of the concrete was also 2 orders of magnitude greater than that of comparable concrete which had been stored in the laboratory atmosphere for 19 years. Analysis of the effluent in water permeability tests revealed the presence of unusual amounts of soluble materials, mainly Na and K but little Ca, in the reactor concrete. This suggested service-related deterioration of the concrete rather than the release of soluble Ca by continuing hydration of cement

  20. Large-scale demonstration of reliability centered maintenance at two nuclear generating stations

    International Nuclear Information System (INIS)

    Gaertner, J.P.; Edgar, C.; Rodin, M.E.

    1989-01-01

    This paper reports that after successful single-system pilot applications of Reliability Centered Maintenance (RCM) at various utilities, EPRI with Rochester Gas and Electric and Southern California Edison is undertaking multiple-system applications of RCM at their respective nuclear plants. The objective is to demonstrate the feasibility and cost-effectiveness of large-scale RCM application. In addition, each utility has plant-specific objectives to improve maintenance and plant availability. Each project has selected a prioritized list of some 15-20 systems on which to perform RCM. Each project is employing somewhat different RCM analysis methods, both of which conform to a global RCM definition applicable to all EPRI RCM work to date. Each project has developed important insights for improving cost and value of future analyses. Both projects will have applied the RCM process, including implementation, on several plant systems by April 1989

  1. Technical Specifications, Seabrook Station, Unit 1 (Docket No. 50-443). Appendix ''A'' to License No. NPF-56

    International Nuclear Information System (INIS)

    1986-10-01

    This report provides specifications for the Seabrook Station Unit 1 reactor concerning: safety limits and limiting safety settings; limiting conditions for operation and surveillance requirements; design features; and administrative controls

  2. Construction work management for nuclear power stations

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1982-01-01

    Nuclear power generation is positioned as the nucleus of petroleum substitution. In the Kansai Electric Power Co., efforts have been made constantly to operate its nuclear power plants in high stability and safety. At present, Kansai Electric Power Co. is constructing Units 3 and 4 in the Takahama Nuclear Power Station in Fukui Prefecture. Under the application of the management of construction works described here, both the nuclear power plants will start operation in 1985. The activities of Kansai Electric Power Co. in the area of this management are described: an outline of the construction works for nuclear power stations, the management of the construction works in nuclear power stations (the stages of design, manufacturing, installation and test operation, respectively), quality assurance activities for the construction works of nuclear power plants, important points in the construction work management (including the aspects of quality control). (J.P.N.)

  3. Indian Point Station, Units 1, 2, and 3. Annual operating report for 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Unit 1 remained in a shutdown condition pending a decision by the Company on the installation of an ECCS as required by NRC. Net electrical power generated by Unit 2 was 2,267,654 MWH with the unit on line 3,056.45 hrs. Unit 3 generated 1,872,947 MWH and was on line 2,286.01 hrs. Information is presented concerning operations, reportable events, corrective maintenance, fuel performance, radioactivity releases, shutdowns, primary coolant chemistry, and occupational radiation exposures

  4. IEEE guide for planning of pre-operational testing programs for class 1E power systems for nuclear-power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The Institute of Electrical and Electronics Engineers (IEEE) guide for pre-operational testing of Class 1E power systems for nuclear-power generating stations is presented. The guidelines apply to power systems both ac and dc supplies but not to the equipment which utilizes the ac and dc power. The pre-operational tests are performed after appropriate construction tests

  5. The nuclear power generation

    International Nuclear Information System (INIS)

    Serres, R.

    1999-01-01

    The French nuclear generating industry is highly competitive. The installations have an average age of fifteen years and are half way through their expected life. Nuclear power accounts for 70% of the profits of the French generating company, EDF. Nuclear generation has a minimal effect on the atmosphere and France has a level of CO 2 emissions, thought to be the main cause of the greenhouse effect, half that of Europe as a whole. The air in France is purer than in neighbouring countries, mainly because 75% of all electrical power is generated in nuclear plants and 15% in hydroelectric stations. The operations and maintenance of French nuclear power plants in the service and distribution companies out of a total of 100 000 employees in all, 90 % of whom are based in mainland France. (authors)

  6. Application of best estimate and uncertainty safety analysis methodology to loss of flow events at Ontario's Power Generation's Darlington Nuclear Generating Station

    International Nuclear Information System (INIS)

    Huget, R.G.; Lau, D.K.; Luxat, J.C.

    2001-01-01

    Ontario Power Generation (OPG) is currently developing a new safety analysis methodology based on best estimate and uncertainty (BEAU) analysis. The framework and elements of the new safety analysis methodology are defined. The evolution of safety analysis technology at OPG has been thoroughly documented. Over the years, the use of conservative limiting assumptions in OPG safety analyses has led to gradual erosion of predicted safety margins. The main purpose of the new methodology is to provide a more realistic quantification of safety margins within a probabilistic framework, using best estimate results, with an integrated accounting of the underlying uncertainties. Another objective of the new methodology is to provide a cost-effective means for on-going safety analysis support of OPG's nuclear generating stations. Discovery issues and plant aging effects require that the safety analyses be periodically revised and, in the past, the cost of reanalysis at OPG has been significant. As OPG enters the new competitive marketplace for electricity, there is a strong need to conduct safety analysis in a less cumbersome manner. This paper presents the results of the first licensing application of the new methodology in support of planned design modifications to the shutdown systems (SDSs) at Darlington Nuclear Generating Station (NGS). The design modifications restore dual trip parameter coverage over the full range of reactor power for certain postulated loss-of-flow (LOF) events. The application of BEAU analysis to the single heat transport pump trip event provides a realistic estimation of the safety margins for the primary and backup trip parameters. These margins are significantly larger than those predicted by conventional limit of the operating envelope (LOE) analysis techniques. (author)

  7. Relationship between IEEE Std. 7-4.3.2-1993 and ASME NQA-1, parts I and II revisions and the impact on nuclear power generating stations

    International Nuclear Information System (INIS)

    Blauw, R.J.

    1996-01-01

    Clear understanding of software related design control requirements is key to growth in the use of computers in nuclear power generating stations. Inconsistent terminology within the nuclear and software standards arena has impacted the ability of both nuclear station system engineers (i.e., the domain expert) to clearly communicate with the software/computer hardware experts. In order for computer development to occur both groups need to have a common terminology basis. Without this commonality, inappropriate application of requirements could result. This paper will present a overview of ongoing efforts within the Institute of Electrical and Electronics Engineers Nuclear Power Engineering Committee (IEEE NPEC) and the American Society of Mechanical Engineers Nuclear Quality Assurance (ASME NQA) Committee to develop this commonality

  8. Safety evaluation report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

    International Nuclear Information System (INIS)

    1987-03-01

    Supplement No. 8 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by th Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement provides recent information regarding resolution of the license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that the Byron Station, Unit 2 can be operated by the licensee at power levels greater than 5% without endangering the health and safety of the public

  9. Nuclear Regulatory Commission Issuances, September 1981

    International Nuclear Information System (INIS)

    1981-01-01

    Contents include: Issuances of the Nuclear Regulatory Commission--Commonwealth Edison Company (Dresden Nuclear Power Station, Unit 1), Consolidated Edison Company of New York (Indian Point, Unit 2), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Station, Unit 1), Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Power Authority of the State of New York (Indian Point, Unit 3), Texas Utilities Generating Company, et al. (Comanche Peak Steam Electric Station, Units 1 and 2); Issuances of Atomic Safety and Licensing Appeal Boards--Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Philadelphia Electric Company, et al. (Peach Bottom Atomic Power Statin, Units 2 and 3), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Statin, Unit No. 2), Public Service Electric and Gas Company (Hope Creek Generating Station, Units 1 and 2), The Toledo Edison Company, et al. (Davis-Besse Nuclear Power Station, Units 2 and 3); Issuances of the Atomic Safety Licensing Boards--Cleveland Electric Illuminating Company, et al. (Perry Nuclear Power Plant, Units 1 and 2), Commonwealth Edison Company (Dresden Station, Units 2 and 3), Houston Lighting and Power Company (Allens Creek Nuclear Generating Station, Unit 1), Southern California Edison Company, et al. (San Onofre Nuclear Generating Station, Units 2 and 3), Texas Utilities Generating Company, et al. (Comanche Peak Steam Electric Station, Units 1 and 2), Texas Utilities Generating Company, et al

  10. Subsidence analysis Forsmark nuclear power plant - unit 1

    International Nuclear Information System (INIS)

    Bono, Nancy; Fredriksson, Anders; Maersk Hansen, Lars

    2010-12-01

    On behalf of SKB, Golder Associates Ltd carried out a risk analysis of subsidence during Forsmark nuclear power plant in the construction of the final repository for spent nuclear fuel near and below existing reactors. Specifically, the effect of horizontal cracks have been studied

  11. Nuclear power generating costs

    International Nuclear Information System (INIS)

    Srinivasan, M.R.; Kati, S.L.; Raman, R.; Nanjundeswaran, K.; Nadkarny, G.V.; Verma, R.S.; Mahadeva Rao, K.V.

    1983-01-01

    Indian experience pertaining to investment and generation costs of nuclear power stations is reviewed. The causes of investment cost increases are analysed and the increases are apportioned to escalation, design improvements and safety related adders. The paper brings out the fact that PHWR investment costs in India compare favourably with those experienced in developed countries in spite of the fact that the programme and the unit size are relatively much smaller in India. It brings out that in India at current prices a nuclear power station located over 800 km from coal reserves and operating at 75% capacity factor is competitive with thermal power at 60% capacity factor. (author)

  12. Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

    International Nuclear Information System (INIS)

    Abucay, R.C.; Mahil, K.S.; Goszczynski, J.J.

    1995-01-01

    The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

  13. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  14. Field testing of a 1,300MVA turbine generator for the Oi nuclear-power station

    International Nuclear Information System (INIS)

    Inoue, Toshiaki; Tajiri, Yoshiaki; Ito, Hiroyuki; Fukuda, Mitsuo.

    1980-01-01

    The first Mitsubishi 1,300MVA turbine generator for this power station was put into commercial operation in March 1979, and the second unit in December of that year. The turbine generators use new technology in a variety of areas, including the cooling system, to achieve great increases in capacity over previously designed generators, and are destined to become the worldwide standard for large-scale generators of this type. Valuable experience was gained in the installation and testing of the generators. The outline of the tests performed on the generators with respect to heating and vibration are described in the article. (author)

  15. Analogue to digital upgrade project-boiler feedwater control system for Bruce Power nuclear units 1 & 2

    International Nuclear Information System (INIS)

    Long, R.

    2012-01-01

    Bruce Power Nuclear Generating Station A, “Bruce A” is in the final stages of its Restart Project. This capital project will see a large scale rehabilitation of Units 1 and 2 resulting in addition of 1500MW of safe, reliable, clean electricity to the Ontario grid. Restart Project Scope 375, Boiler Feedwater Controls Upgrade was sanctioned to replace obsolete analog devices with a modern digital control system. This project replaced the existing Foxboro H Line analog controls which comprised of 81 individual control modules and support instrumentation. The replacement system was a Triconex Triple Modular Redundant PLC which interfaces with two redundant touch screen monitors. The upgraded digital system incorporates the following controls: 1. Boiler Level Control Loops 2. Dearator Level Control Loops 3. Dearator Pressure Control Loops 4. Boiler Feedwater Recirculation Flow Control Loops A number of technical challenges were addressed when installing a new digital system within the existing plant configuration. Interfaces to new, old and refurbished field devices must be understood as well as implications of connecting to the plant’s Digital Control Computers (DCC’s) and newly installed Steam Generators. The overall project involved many stakeholders to address various requirements from conceptual / design stage through procurement, construction, commissioning and return to service. In addition, the project highlighted the unique requirements found in Nuclear Industry with respect to Human Factors and Software Quality Assurance. (author)

  16. Application of ecological risk assessment to establish nonhuman environmental protection at nuclear generating stations in Ontario, Canada

    International Nuclear Information System (INIS)

    Wismer, D.A.

    2004-01-01

    A screening ecological risk assessment was performed for regulatory compliance at three Ontario nuclear generating station sites to establish design requirements for a routine contaminant monitoring program and to address the need for non-contaminant stressor management. Site specific assessments went beyond traditional contaminant risk assessment to include stressors associated with land-use change, cooling water systems and site storm water runoff. Valued terrestrial and aquatic ecosystem components were selected from species lists after stakeholder consultation, and contaminants of concern were selected based on their relative loadings, and with respect to regulatory and literature-based benchmarks. Predictive modeling was used to estimate chemical and radionuclide exposures and likelihood of effects. Adverse effects on individual biota were predicted for aqueous emissions of chlorine and storm water but not for radionuclides. Retrospective analyses of past field monitoring were used to determine likelihood of effects from non-contaminant stressors. Individual-level adverse effects were observed for fish losses from cooling water withdrawal. Depending on the site and the biological species, either beneficial or adverse effects from thermal discharge and land-use change were observed. Followup studies include monitoring, laboratory study, computer modeling and mitigation. Field monitoring will generate more precise species-level estimates of intake fish losses, magnitude of fish response to thermal discharge and chlorine concentrations in near-field discharge waters. Laboratory study is determining the effectiveness of intake fish loss mitigation technology. Computer fish population models are being used to design field studies and interpret individual level effects. Mitigation includes storm water controls and habitat biodiversity management projects to offset past losses from site development and construction. Routine contaminant monitoring is planned to

  17. Studying dynamics of indicators of nuclear power stations exploitation (the case of US nuclear power stations)

    OpenAIRE

    Varshavsky, Leonid

    2013-01-01

    Analysis of external and internal factors influencing significant improvement of economic indicators of US nuclear power stations in the 1990s is carried out. Approaches to modeling dynamics of capacity factors of nuclear power stations are proposed. Comparative analysis of dynamics of capacity factors and occupational radiation exposure for various generations of US nuclear power plants is carried out. Dynamical characteristics of «learning by doing» effects for analyzed indicators are measu...

  18. Integrated plant safety assessment: systematic evaluation program. Oyster Creek nuclear generating station. GPU Nuclear Corporation and Jersey Central Power and Light Company. Docket No. 50-219

    International Nuclear Information System (INIS)

    1982-09-01

    The Systematic Evaluation Program was initiated in February 1978 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Oyster Creek Nuclear Generating Station (located in Ocean County, New Jersey), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

  19. Decommissioning of units 1 - 4 at Kozloduy nuclear power plant in Bulgaria

    International Nuclear Information System (INIS)

    Dishkova, Denitsa

    2014-01-01

    Nuclear safety and security are absolute priorities for the European Union countries and this applies not only to nuclear power plants in operation but also to decommissioning. In terms of my technical background and my working experience in the field of licensing and environmental impact assessment during the decommissioning of Units 1 to 4 at Kozloduy Nuclear Power Plant (KNPP) in Bulgaria, I decided to present the strategy for decommissioning of Units 1 to 4 at KNPP which was selected and followed to achieve safe and effective decommissioning process. The selected strategy in each case must meet the legislative framework, to ensure safe management of spent fuel and radioactive waste, to provide adequate funding and to lead to positive socio-economic impact. The activities during the decommissioning generate large volume of waste. In order to minimize their costs and environmental impact it should be given a serious consideration to the choice, the development and the implementation of the most adequate process for treatment and the most appropriate measurement techniques. The licensing process of the decommissioning activities is extremely important and need to cope with all safety concerns and ensure optimal waste management. (authors)

  20. IEEE Std 600: IEEE trial-use standard requirements for organizations that conduct qualification testing of safety systems equipment for use in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The purpose of this standard is to provide requirements for establishing a program for conducting qualification tests of safety systems equipment used in nuclear power generating stations. Compliance with the requirements of this standard does not assure the adequacy of the qualification tests performed. This standard applies to organizations that conduct qualification tests on equipment that has a definable safety function and is an identifiable part of a safety system for use in nuclear power generating stations. It requires a technical program, a quality assurance program, and a demonstrated ability to meet specified technical requirements. It does not apply to materials tests, production tests, normal performance testing, qualification by analysis, qualification by operating experience, or reliability tests such as diesel-generator multiple start tests. The intent of this standard is to achieve greater consistency, reliability, and reproducibility of test results and to provide adequate control of qualification testing of safety systems equipment

  1. Nuclear power generation

    International Nuclear Information System (INIS)

    Hirao, Katumi; Sato, Akira; Kaimori, Kimihiro; Kumano, Tetsuji

    2001-01-01

    Nuclear power generation for commercial use in Japan has passed 35 years since beginning of operation in the Tokai Nuclear Power Station in 1966, and has 51 machines of reactor and about 44.92 MW of total output of equipment scale in the 21st century. However, an environment around nuclear energy becomes severer at present, and then so many subjects to be overcome are remained such as increased unreliability of the public on nuclear energy at a chance of critical accident of the JCO uranium processing facility, delay of pull-thermal plan, requirement for power generation cost down against liberalization of electric power, highly aging countermeasure of power plant begun its operation as its Genesis, and so on. Under such conditions, in order that nuclear power generation in Japan survives as one of basic electric source in future, it is necessary not only to pursue safety and reliability of the plant reliable to the public, but also to intend to upgrade its operation and maintenance by positively adopting good examples on operational management method on abroad and to endeavor further upgrading of application ratio of equipments and reduction of generation cost. Here were outlined on operation conditions of nuclear power stations in Japan, and introduced on upgrading of their operational management and maintenance management. (G.K.)

  2. 78 FR 44603 - Byron Nuclear Station, Units 1 and 2, and Braidwood Nuclear Station, Units 1 and 2; Exelon...

    Science.gov (United States)

    2013-07-24

    ... Documents Access and Management System (ADAMS): You may access publicly available documents online in the... order to serve documents through the Electronic Information Exchange System, users will be required to... of acceptance for docketing and notice of opportunity for hearing regarding the renewal of operating...

  3. Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

    Science.gov (United States)

    Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

    2013-09-01

    Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

  4. Periodical inspection in nuclear power stations

    International Nuclear Information System (INIS)

    1986-01-01

    Periodical inspection is presently being made of eight nuclear power plants in nuclear power stations. Up to the present time, in three of them, failures as follows have been observed. (1) Unit 3 (PWR) of the Mihama Power Station in The Kansai Electric Power Co., Inc. Nineteen heat-transfer tubes of the steam generators were plugged up due to failure. A fuel assembly with a failed spring fixture and in another the control-rod cluster with a failed control rod fixture were replaced. (2) Unit 2 (PWR) of the Oi Power Station in The Kansai Electric Power Co., Inc. Eight heat-transfer tubes of the heat exchangers were plugged up due to failure. (3) Unit 6 (BWR) of the Fukushima Nuclear Power Station I in The Tokyo Electric Power Co., Inc. A fuel assembly with leakage was replaced. (Mori, K.)

  5. Efforts to perform safe and efficient decommissioning for Tsuruga Power Station Unit 1

    International Nuclear Information System (INIS)

    Saito, Shiro; Yamauchi, Toyoaki; Austin, Colin R.

    2017-01-01

    Tsuruga Power Station Unit-1 (Tsuruga-1) started commercial operation in March 1970, and the decision to terminate operation was made in 2015. In April 2016, JAPC signed an agreement with Energy Solutions (ES) on strategic cooperation for domestic D and D projects for introduction of successful international experiences. As a first step in this cooperation, D and D know-how developed by ES in the US is being applied to Tsuruga-1 with verifying its applicability to domestic D and D projects. One of the efforts is human resource development. JAPC has also started introduction of ES's project management method to the Tsuruga-1 project for solid project management and the base line is currently being prepared. Regarding the waste disposal paths, application document of approval for measurement and evaluation of clearance material was submitted in September 2016. However the disposal paths for waste are not established in Japan. It is necessary to cooperate with the government, utilities and local stakeholders to establish waste disposal paths. Because it is also important to obtain the understanding from local communities, JAPC and ES will try positively to utilize local companies for D and D works. JAPC and ES believe that their relationship will ensure success of the Tsuruga-1 NPP decommissioning project. (author)

  6. Nuclear safety inspection in treatment process for SG heat exchange tubes deficiency of unit 1, TNPS

    International Nuclear Information System (INIS)

    Zhang Chunming; Song Chenxiu; Zhao Pengyu; Hou Wei

    2006-01-01

    This paper describes treatment process for SG heat exchange tubes deficiency of Unit 1, TNPS, nuclear safety inspection of Northern Regional Office during treatment process for deficiency and further inspection after deficiency had been treated. (authors)

  7. Balakovo nuclear power station

    International Nuclear Information System (INIS)

    1996-01-01

    A key means of improving the safety and reliability of nuclear power plants is through effective training of plant personnel. The goal of this paper is to show the progress of the training at the Balakovo Nuclear Power Plant, and the important role that international cooperation programs have played in that progress

  8. IEEE C37.105-1987: IEEE standard for qualifying Class 1E protective relays and auxiliaries for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard describes the basic principles, requirements, and methods for qualifying Class 1E protective relays and auxiliaries such as test and control switches, terminal blocks, and indicating lamps for applications in nuclear power generating stations. When properly employed it can be used to demonstrate the design adequacy of such equipment under normal, abnormal, design basis event and post design basis event conditions in accordance with ANSI/IEEE Std 323-1983. When protective relays and auxiliaries are located in areas not subject to harsh environments, environmental qualification is not required. Protective relays and auxiliaries located inside primary containment in a nuclear power generating station present special conditions beyond the scope of this document. The qualification procedure presented is generic in nature. Other methods may be used at the discretion of the qualifier, provided the basic precepts of ANSI/IEEE Std 32301983 are satisfied

  9. IEEE No. 323, IEEE trial-use standard: General guide for qualifying Class I electric equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This document describes the basic requirements for the qualification of Class I electric equipment. This is equipment which is essential to the safe shutdown and isolation of the reactor or whose failure or damage could result in significant release of radioactive material. The purpose of this document is to provide guidance for demonstrating the qualifications of electrical equipment as required in the IEEE Std 279 -- Criteria for Nuclear Power Generating Station Protection Systems, and IEEE Std 308 -- Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations. The qualification methods described may be used in conjunction with the Guides for qualifying specific types of equipment, (see Foreword), for updating qualification following modifications or for qualifying equipment for which no applicable Guide exists

  10. IEEE Std 535-1979: IEEE standard for qualification of Class 1E lead storage batteries for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This document describes qualification methods for Class 1E lead storage batteries and racks to be used in nuclear power generating stations outside of primary containment. Qualification required in ANSI/IEEE Std 279-1971 and IEE Std 308-1978, can be demonstrated by using the procedures provided in this standard in accordance with IEEE Std 323-1974. Battery sizing, maintenance, capacity testing, installation, charging equipment and consideration of other type batteries are beyond the scope of this standard

  11. IEEE Std 535-1986: IEEE standard for qualification of Class 1E lead storage batteries for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This document describes qualification methods for Class 1E lead storage batteries and racks to be used in nuclear power generating stations outside of primary containment. Qualification required in ANSI/IEEE Std 308-1980 can be demonstrated by using the procedures provided in this standard in accordance with ANSI/IEEE Std 323-1983. Battery sizing, maintenance, capacity testing, installation, charging equipment, and consideration of other type batteries are beyond the scope of this standard

  12. Principles of nuclear power station control

    International Nuclear Information System (INIS)

    Knowles, J.B.

    1975-12-01

    This memorandum represents lecture notes first distributed as part of a UKAEA introductory course on Reactor Technology held during November 1975. A nuclear power station is only one element of a dispersed interconnected arrangement of other nuclear and fossil-fired units which together constitute the national 'grid'. Thus the control of any one station must relate to the objectives of the grid network as a whole. A precise control of the supply frequency of the grid is achieved by regulating the output power of individual stations, and it is necessary for each station to be stable when operating in isolation with a variable load. As regards individual stations, several special control problems concerned with individual plant items are discussed, such as: controlled reactivity insertions, temperature reactivity time constants and flow instability. A simplified analysis establishes a fundamental relationship between the stored thermal energy of a boiler unit (a function of mechanical construction) and the flexibility of the heat source (nuclear or fossil-fired) if the station is to cope satisfactorily with demands arising from unscheduled losses of other generating sets or transmission capacity. Two basic control schemes for power station operation are described known as 'coupled' and 'decoupled control'. Each of the control modes has its own merits, which depend on the proposed station operating strategy (base load or load following) and the nature of the heat source. (U.K.)

  13. Extension of life of nuclear power stations

    International Nuclear Information System (INIS)

    Takahashi, Hideaki

    1991-01-01

    At the time of designing nuclear power stations, as their service life, generally 40 years are taken, and the basic design specifications of machinery and equipment are determined. In USA where atomic energy has been developed, the new construction of nuclear power stations is cased for a while, however, if this situation continues as it is, since old power stations reach the service life of 40 years and are retired in near future, it is feared that the circumstance of the total amount of power generation becoming short will occur. As one of the countermeasures to this, the research on the extension of life of nuclear power stations has been carried out in many fields in USA, and it is expected that the application for extending the life for the power stations constructed in the initial period of development is submitted in 1991. The researches that have been carried out for solving the technical problems in this extension of life and the situation in Japan are reported. The NEC of USA decided that the operation period of nuclear power stations in USA, which is considered to be 40 years so far, can be extended up to the limit of 20 years. The background and circumstances of this problem in USA, Nuclear Plant Aging Research Program, Plant Life Extension Program and so on are reported. (K.I.)

  14. Hardware-oriented reliability centered maintenance for the diesel generators of Wolsong unit 1

    International Nuclear Information System (INIS)

    Bae, Sang Min; Park, Jin Hee; Kim, Tae Woon; Lee, Yoon Kee; Song, Jin Bae

    1997-01-01

    The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shut down the nuclear reactor safely in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defects, its failure modes were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless maintenance, the preventive maintenance, the database of maintenance information, and the predictive maintenance

  15. Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

    International Nuclear Information System (INIS)

    Bain, R.A.; Bayer, R.K.

    1993-01-01

    In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements

  16. Nuclear power station achievement 1968-1980

    International Nuclear Information System (INIS)

    Howles, L.R.

    This report reviews and gives an analysis of the achievement of operating nuclear power stations in the Western world on three relevant bases: (1) both annual and cumulative achievement of all nuclear power stations at a particular time; (2) cumulative achievement of all nuclear power stations at the end of the first and subsequent years of their lives to show trends with age; (3) achievement based on refuelling period considerations. Nowhere in the report are any operating details ignored, omitted or eliminated in the method of analysis. Summarising the results of the reviews shows: an improvement with time from initial electricity generation on all bases; that initially, larger sizes of reactor/turbine operate less well than smaller sizes (except for PHWR's); that after an initial number of years, the largest size units operate as well as the intermediate and smaller sizes, or better in the PHWR case; that a 75 per cent cumulative load factor achievement in the middle years of a reactor/turbines life can be expected on the refuelling period considerations base; that at June 1980, 35 nuclear power stations achieved an annual load factor over 75 per cent; that the above achievement was possible despite the repercussions following the Three Mile Island 'accident' and the shutdowns in the USA for piping system seismic adequacy checks required by the Nuclear Regulatory Commission for five nuclear power stations; and that even when reactors/turbines are reaching towards the end of their design life, there is no rapid deterioration in their achievements. (author)

  17. Simulators of nuclear power stations

    International Nuclear Information System (INIS)

    Zanobetti, D.

    1984-01-01

    The report deals with the simulators of nuclear power stations used for the training of operators and for the analysis of operations. It reviews the development of analogical, hybrid and digital simulators up to the present, indicating the impact resulting from the TMI-2 accident. It indicates, the components of simulators and the present accepted terminology for a classification of the various types of simulators. It reviews the present state of the art of the technology: how a basic mathematical model of a nuclear power system is worked out and what are the technical problems associated with more accurate models. Examples of elaborate models are given: for a PWR pressurizer, for an AGR steam generator. It also discusses certain problems of hardware technology. Characteristics of present replica simulators are given with certain details: simulated transient evolutions and malfunctions, accuracy of simulation. The work concerning the assessment of the validity of certain simulators is reported. A list of simulator manufacturers and a survey of the principal simulators in operation in the countries of the European Community, in the United States, and in certain other countries are presented. Problem associated with the use of simulators as training facilities, and their use as operational devices are discussed. Studies and research in progress for the expected future development of simulators are reviewed

  18. Final Environmental Statement related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446)

    International Nuclear Information System (INIS)

    1989-10-01

    In September 1981, the staff of the Nuclear Regulatory Commission (NRC) issued its Final Environmental Statement (NUREG-0775) related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446), located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. The NRC has prepared this supplement to NUREG-0775 to present its evaluation of the alternative of operating Comanche Peak with the installation of further severe-accident-mitigation design features. The NRC has discovered no substantial changes in the proposed action as previously evaluated in the Final Environmental Statement that are relevant to environmental concerns and bearing on the licensing of Comanche Peak Steam Electric Station, Units 1 and 2. 6 refs., 3 tabs

  19. Low-level radioactive waste from nuclear power generating stations: Characterization, classification and assessment of activated metals and waste streams

    International Nuclear Information System (INIS)

    Thomas, V.W.; Robertson, D.E.; Thomas, C.W.

    1993-02-01

    Since the enactment of 10 CFR Part 61, additional difficult-to-measure long-lived radionuclides, not specified in Tables 1 2 of Part 61, have been identified (e.g., 108m Ag, 93 Mo, 36 Cl, 10 Be, 113m Cd, 121m Sn, 126 Sn, 93m Nb) that may be of concern in certain types of waste. These nuclides are primarily associated with activated metal and perhaps other nuclear power low-level waste (LLW) being sent to disposal facilities. The concentration of a radionuclide in waste materials is normally determined by direct measurement or by indirect calculational methods, such as using a scaling factor to relate inferred concentration of a difficult-to-measure radionuclide to another that is easily measured. The total disposal site inventory of certain difficult-to-measure radionuclides (e.g., 14 C, 129 I, and 99 Tc) often control the total quantities of radioactive waste permitted in LLW burial facilities. Overly conservative scaling factors based on lower limits of detection (LLD), often used in the nuclear power industry to estimate these controlling nuclides, could lead to premature closure of a disposal facility. Samples of LLW (Class B and C activated metals [AM] and other waste streams) are being collected from operating nuclear power stations and analyzed for radionuclides covered in 10 CFR Part 61 and the additional difficult-to-measure radionuclides. This analysis will enhance the NRC's understanding of the distribution and projected quantities of radionuclides within AM and LLW streams from commercial nuclear power stations. This research will also provide radiological characterization of AM specimens for others to use in leach-rate and lysimeter experiments to determine nuclide releases and subsequent movement in natural soil environments

  20. Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Dockets Nos. STN 50-454 and STN 50-455)

    International Nuclear Information System (INIS)

    1984-10-01

    Supplement No. 5 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report. Because of the favorable resolution of the items discussed in this report, the staff concludes that there is reasonable assurance that the facility can be operated by the applicant without endangering the health and safety of the public

  1. Safety evaluation report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457)

    International Nuclear Information System (INIS)

    1983-11-01

    The Safety Evaluation Report for the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Reed Township, Will County, Illinois. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  2. Search and Retrieval of Foreign Objects for the Steam Generator of Wolsung NPP Unit 1

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Kyung-Ho

    2016-01-01

    We developed a foreign object search and retrieval (FOSAR) system for Wolsung NPP unit 1 steam generators. The steam generators of Wolsung NPP unit 1 have one 2.5 inch hand hole and two 4 inch hand holes. The FOSAR system was designed to be installed through 4 inch hand holes. Using permanent magnet, the FOSAR system was firmly attached to the vertical annulus wall of the steam generator. We successfully developed the FOSAR system for Wolsung NPP unit 1. Using the developed FOSAR system, technicians successfully found and removed various foreign objects. Most of the foreign objects, we found, were made of carbon steel sheet, therefore magnet tool was the most useful to remove it. Alligator tool was sometimes used. Based on the experience during the FOSAR activities, we are developing a lancing system for Wolsung NPP unit 1. It will be designed and manufactured until November 2016

  3. Search and Retrieval of Foreign Objects for the Steam Generator of Wolsung NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Kyung-Ho [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    We developed a foreign object search and retrieval (FOSAR) system for Wolsung NPP unit 1 steam generators. The steam generators of Wolsung NPP unit 1 have one 2.5 inch hand hole and two 4 inch hand holes. The FOSAR system was designed to be installed through 4 inch hand holes. Using permanent magnet, the FOSAR system was firmly attached to the vertical annulus wall of the steam generator. We successfully developed the FOSAR system for Wolsung NPP unit 1. Using the developed FOSAR system, technicians successfully found and removed various foreign objects. Most of the foreign objects, we found, were made of carbon steel sheet, therefore magnet tool was the most useful to remove it. Alligator tool was sometimes used. Based on the experience during the FOSAR activities, we are developing a lancing system for Wolsung NPP unit 1. It will be designed and manufactured until November 2016.

  4. Socioeconomic impacts of nuclear generating stations: Peach Bottom case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Peach Bottom nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  5. Socioeconomic impacts of nuclear generating stations: St. lucie case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Weisiger, M.L.; Pijawka, K.D.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the St. Lucie nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study are during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  6. Socioeconomic impacts of nuclear generating stations: Oconee case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Flynn, J.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Oconee nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  7. Socioeconomic impacts of nuclear generating stations: Rancho Seco case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Bergmann, P.A.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Rancho Seco nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  8. Socioeconomic impacts of nuclear generating stations: Three Mile Island case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Flynn, C.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Three Mile Island nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  9. Socioeconomic impacts of nuclear generating stations - Diablo Canyon case study. Technical report for 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Pijawka, K.D.; Yaquinto, G.

    1982-07-01

    This report documents a case study of the socioeconomic impacts of the construction and operation of the Diablo Canyon nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlemeent patterns and housing, local government and public services, social structure, and public response in the study during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  10. Socioeconomic impacts of nuclear generating stations: D. C. Cook case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Branch, K.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the D. C. Cook nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  11. Socioeconomic impacts of nuclear generating stations: Surry case study. Technical report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Flynn, J.

    1982-07-01

    The report documents a case study of the socioeconomic impacts of the construction and operation of the Surry nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  12. Interim reliability evaluation program: analysis of the Arkansas Nuclear One. Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kolb, G.J.; Kunsman, D.M.; Bell, B.J.

    1982-06-01

    This report represents the results of the analysis of Arkansas Nuclear One (ANO) Unit 1 nuclear power plant which was performed as part of the Interim Reliability Evaluation Program (IREP). The IREP has several objectives, two of which are achieved by the analysis presented in this report. These objectives are: (1) the identification, in a preliminary way, of those accident sequences which are expected to dominate the public health and safety risks; and (2) the development of state-of-the-art plant system models which can be used as a foundation for subsequent, more intensive applications of probabilistic risk assessment. The primary methodological tools used in the analysis were event trees and fault trees. These tools were used to study core melt accidents initiated by loss of coolant accidents (LOCAs) of six different break size ranges and eight different types of transients

  13. Nuclear safeguards control in nuclear power stations

    International Nuclear Information System (INIS)

    Boedege, R.; Braatz, U.; Heger, H.

    1976-01-01

    The execution of the Non-Proliferation Treaty (NPT) has initiated a third phase in the efforts taken to ensure peace by limiting the number of atomic powers. In this phase it is important, above all, to turn into workable systems the conditions imposed upon technology by the different provisions of the Verification Agreement of the NPT. This is achieved mainly by elaborating annexes to the Agreement specifically geared to certain model plants, typical representatives selected for LWR power stations being the plants at Garigliano, Italy (BWR), and Stade, Federal Republic of Germany (PWR). The surveillance measures taken to prevent any diversion of special nuclear material for purposes of nuclear weapons manufacture must be effective in achieving their specific objective and must not impede the circumspect management of operations of the plants concerned. A VDEW working party has studied the technical details of the planned surveillance measures in nuclear power stations in the Federal Republic of Germany and now presents a concept of material balancing by units which meets the conditions imposed by the inspection authority and could also be accepted by the operators of nuclear power stations. The concept provides for uninterrupted control of the material balance areas of the nuclear power stations concerned, allows continuous control of the whole nuclear fuel cycle, is based exclusively on existing methods and facilities, and can be implemented at low cost. (orig.) [de

  14. 75 FR 18572 - Supplemental Environmental Impact Statement for Sequoyah Nuclear Plant Units 1 and 2 License...

    Science.gov (United States)

    2010-04-12

    ... TENNESSEE VALLEY AUTHORITY Supplemental Environmental Impact Statement for Sequoyah Nuclear Plant... National Environmental Policy Act. TVA will prepare a supplemental environmental impact statement (SEIS) to update information in the 1974 Final Environmental Statement for Sequoyah Nuclear Plant Units 1 and 2...

  15. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1994-01-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  16. Wolf Creek Generating Station containment model

    International Nuclear Information System (INIS)

    Nguyen, D.H.; Neises, G.J.; Howard, M.L.

    1995-01-01

    This paper presents a CONTEMPT-LT/28 containment model that has been developed by Wolf Creek Nuclear Operating Corporation (WCNOC) to predict containment pressure and temperature behavior during the postulated events at Wolf Creek Generating Station (WCGS). The model has been validated using data provided in the WCGS Updated Safety Analysis Report (USAR). CONTEMPT-LT/28 model has been used extensively at WCGS to support plant operations, and recently, to support its 4.5% thermal power uprate project

  17. Monitoring of nuclear power stations

    International Nuclear Information System (INIS)

    Ull, E.; Labudda, H.J.

    1987-01-01

    The purpose of the invention is to create a process for undelayed automated detection and monitoring of accidents in the operation of nuclear power stations. According to the invention, this problem is solved by the relevant local measurements, such as radiation dose, components and type of radiation and additional relevant meteorological parameters being collected by means of wellknown data collection platforms, these being transmitted via transmission channels by means of satellites to suitable worldwide situated receiving stations on the ground, being processed there and being evaluated to recognise accidents. The local data collection platforms are used in the immediate vicinity of the nuclear power station. The use of aircraft, ships and balloons as data collection systems is also intended. (HWJ)

  18. Reviewing nuclear power station achievement

    International Nuclear Information System (INIS)

    Howles, L.R.

    1976-01-01

    For measurement of nuclear power station achievement against original purchase the usual gross output figures are of little value since the term loosely covers many different definitions. An authentically designed output figure has been established which relates to net design output plus house load at full load. Based on these figures both cumulative and moving annual load factors are measured, the latter measuring the achievement over the last year, thus showing trends with time. Calculations have been carried out for all nuclear stations in the Western World with 150 MW(e) gross design output and above. From these are shown: moving annual load factor indicating relative station achievements for all the plants; cumulative load factors from which return of investment can be calculated; average moving annual load factors for the four types of system Magnox, PWR, HWR, and BWR; and a relative comparison of achievement by country in a few cases. (U.K.)

  19. Vibration Spectrum Analysis for Indicating Damage on Turbine and Steam Generator Amurang Unit 1

    Directory of Open Access Journals (Sweden)

    Beny Cahyono

    2017-12-01

    Full Text Available Maintenance on machines is a mandatory asset management activity to maintain asset reliability in order to reduce losses due to failure. 89% of defects have random failure mode, the proper maintenance method is predictive maintenance. Predictive maintenance object in this research is Steam Generator Amurang Unit 1, which is predictive maintenance is done through condition monitoring in the form of vibration analysis. The conducting vibration analysis on Amurang Unit 1 Steam Generator is because vibration analysis is very effective on rotating objects. Vibration analysis is predicting the damage based on the vibration spectrum, where the vibration spectrum is the result of separating time-based vibrations and simplifying them into vibrations based on their frequency domain. The transformation of time-domain-wave into frequency-domain-wave is using the application of FFT, namely AMS Machinery. The measurement of vibration value is done on turbine bearings and steam generator of Unit 1 Amurang using Turbine Supervisory Instrument and CSI 2600 instrument. The result of this research indicates that vibration spectrum from Unit 1 Amurang Power Plant indicating that there is rotating looseness, even though the vibration value does not require the Unit 1 Amurang Power Plant to stop operating (shut down. This rotating looseness, at some point, can produce some indications that similar with the unbalance. In order to avoid more severe vibrations, it is necessary to do inspection on the bearings in the Amurang Unit 1 Power Plant.

  20. The J-resistance curve Leak-before-Break test program on material for the Darlington Nuclear Generating Station

    International Nuclear Information System (INIS)

    Mukherjee, B.

    1988-01-01

    The Darlington Leak-Before-Break (DLBB) approach has been developed for large diameter (21, 22, 24 inch) SA106B heat transport (HT) piping and SA105 fittings as a design alternative to pipewhip restraints and in recognition of the questionable benefits of providing such restraints. Ontario Hydro's DLBB approach is based on the elastic plastic fracture mechanics method. In this test program, J-resistance curves were determined from actual pipe heats that were used in the construction of the Darlington heat transport systems (Units 1 and 2). Test blocks were prepared using four different welding procedures for nuclear Class I piping. The test program was designed to take into account the effect of various factors such as test temperature, crack plane orientation, welding effects, etc., which have influence on fracture properties. A total of 91 tests were conducted. An acceptable lower bound J-resistance curve for the piping steels was obtained by machining maximum thickness specimens from the pipes and by testing side grooved compact tension specimens. Test results showed that all pipes, welds and heat-affected zone materials within the scope of the DLBB program exhibited uppershelf toughness behaviour. All specimens showed high crack initiation toughness Jsub(lc), rising J-resistance curve and stable and ductile crack extension. Toughness of product forms depended on the direction of crack extension (circumferential versus axial crack orientation). Toughness of DLBB welds and parent materials at 250 0 C was lower than that at 20 0 C. (author)

  1. Physical degradation assessment of generator station cables

    International Nuclear Information System (INIS)

    Stonkus, D.J.

    1988-01-01

    Preliminary studies of fossil-fired and nuclear generator station cables indicate that the low voltage PVC insulated cables are in relatively good condition. The insulation is flexible and in the case of nuclear cables can withstand a design basis event after nearly 15 years of service. Cables insulated with styrene butadiene rubber have been found embrittled and cables insulated with SBR should be closely inspected in any plant assurance program. Thermal analysis using oxidative induction technique shows promise to indicate cable insulation degradation. Long term reliability assurance and plant life extension studies are being actively pursued at Ontario Hydro. A major study is currently underway to extend the life of the oldest operating fossil-fuel station, the 8-unit, 2400 MW Lakeview TGS in operation since the 1960s. Plant life assurance programs have been initiated at the 2000 MW Lambton TGS in operation since 1969, and for the oldest operating nuclear plant, Pickering NGS A in operation since the early 1970s. As cables are considered one of the critical components in a generator station due to the extreme difficulty and cost of cable replacement, test programs have been initiated to evaluate the physical degradation of the cables and relate the results to electrical diagnostic tests and to chemical changes. The decommissioning of two small nuclear stations, the 20 MW Nuclear Power Demonstration (NPD) and the 200 MW Douglas Point NGS, which were placed in service in 1962 and 1967 respectively, will provide an opportunity to perform destructive electrical and physical evaluation on field aged cables

  2. Managing nuclear power stations for success

    International Nuclear Information System (INIS)

    Smith, G.

    2006-01-01

    Ontario Power Generation's (OPG) top operational priority is to manage its nuclear assets to ensure they operate as safely, efficiently and cost effectively as possible. In meeting these objectives, the company is focused on continuously improving its nuclear performance and benchmarking that performance against the best in North America. This presentation explores how OPG is improving its nuclear performance and the steps it is taking to sustain performance success going forward. Topics to be discussed include the measures OPG is taking to enhance human performance and station reliability as well as the company's preparations to determine if a business case exists for extending the lives of the Pickering B and eventually the Darlington nuclear stations. (author)

  3. Sociological impacts of nuclear generating stations - summary report on the NRC post-licensing studies. Final report 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Chalmers, J.; Pijawka, D.; Branch, K.; Bergmann, P.; Flynn, J.

    1982-07-01

    The Post-Licensing Studies had four objectives. The first was to identify the socioeconomic effects resulting from the construction and operation of each of twelve nuclear power stations. The socioeconomic variables examined included: economic, demographic, housing, government, public response, and social organization characteristics. The second objective was to determine the way in which the identified effects were evaluated by study area groups. The third objective was to identify the determinants of the project-related effects. This task required knowledge of what combination of site, project, or other determinants was responsible for the project-related effects and for the evaluation of the effects. The fourth objective was to make recommendations with respect to assessment methodologies that could best be used to project the socioeconomic effects of the construction and operation of proposed nuclear generating stations. The objectives of the Post-Licensing Studies are met by the twelve individual case studies and by the Summary Report. The case studies identified the nuclear power stations and describe the evaluation of the effects by area residents. The Summary Report describes the collective findings of the individual case studies, compares the findings across sites to identify possible determinants of the effects, and examines the implication of the findings for future siting decisions and for the methodology most appropriate for projective assessments

  4. Nuclear design report for Ulchin nuclear power plant unit 1, cycle 7

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Rae; Park, Yong soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-04-01

    This report presents nuclear design calculations for Cycle 7 of Ulchin Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 56 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s 36 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 7 amounts to 355 EFPD corresponding to a cycle burnup of 14280 MWD/MTU. (Author) 8 refs., 55 figs., 21 tabs.

  5. Nuclear design report for Kori nuclear power plant unit 1, cycle 13

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Moon, Bok Ja; Cho, Byeong Ho; Jung, Yil Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-04-01

    This report presents nuclear design calculations for cycle 13 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 44 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 16 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 13 amounts to 355 EFPD corresponding to a cycle burnup of 13240 MWD/MTU. (Author) 8 refs., 55 figs., 16 tabs.

  6. Nuclear design report for Yonggwang nuclear power plant unit 1 cycle 9

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young chul; Kim, Jae Hak; Song, Jae Woong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-03-01

    This report presents nuclear design calculations for Cycle 6 of Yonggwng Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s, 60 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 9 amounts to 434 EFPD corresponding to a cycle burnup of 17470 MWD/MTU. (Author) 8 refs., 55 figs., 19 tabs.

  7. Nuclear design report for Ulchin nuclear power plant unit 1, cycle 6

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Kim, Yong Rae; Park, Yong Soo; Cho, Byeong Ho; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    This report presents nuclear design calculations for cycle 6 of Ulchin unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 64 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 32 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 6 amounts to 369 EFPD corresponding to a cycle burnup of 14850 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

  8. Nuclear design report for Yonggwang nuclear power plant unit 1, cycle 8

    International Nuclear Information System (INIS)

    Cho, Young Chul; Kim, Jae Hak; Park, Sang Yoon; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan

    1993-10-01

    This report presents nuclear design calculations for cycle 8 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA's enriched by nominally 3.70 w/o U 235 . Among the KOFA's, 56 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 447 EFPD corresponding to a cycle burnup of 18020 MWD/MTU. (Author) 8 refs., 39 figs., 17 tabs

  9. Nuclear design report for Yonggwang nuclear power plant unit 1, cycle 8

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Chul; Kim, Jae Hak; Park, Sang Yoon; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-10-01

    This report presents nuclear design calculations for cycle 8 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 56 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 447 EFPD corresponding to a cycle burnup of 18020 MWD/MTU. (Author) 8 refs., 39 figs., 17 tabs.

  10. Moderator inlet line hanger replacement for Pickering nuclear power station

    International Nuclear Information System (INIS)

    Kirkpatrick, R.A.; Bowman, J.M.; Symmons, W.R.; El-Nesr, S.

    1988-01-01

    Ontario Hydro's Pickering Nuclear Generating Station (PNGS), Units 1 and 2 were shutdown for large scale fuel channel replacement. Other nonroutine inspection and maintenance activities were performed to determine the overall condition of the units and it was seen that a moderator inlet line hanger (identified as HR-29) had failed in both units. Subsequent inspections during planned maintenance outages of Pickering NGS Units 3 and 4 revealed that hanger HR-29 had failed and required replacement. A research program was conducted to find a suitable technique. These problems included accessing tooling through small inspection ports, manipulating tooling from a significant distance and the high radiation fields within the vault. This paper describes the program undertaken to replace hanger HR-29. (author)

  11. Procedure for qualify class 1E equipment for nuclear power generating stations; Kvalifikacija elektroopreme za stjecanje klase 1E u nuklearnim elektranama

    Energy Technology Data Exchange (ETDEWEB)

    Smiljanic, D [' Rade Koncar' Institut, Zagreb (Yugoslavia)

    1983-07-01

    This paper describes the basic requirements for qualifying Class 1E equipment and systems that are to be used in nuclear power generating stations. The requirements presented include the principles, procedures and methods of qualification. The qualification requirements will confirm the adequacy of the equipment design under normal, abnormal, design basis event, post design basis event and containment test conditions for the performance of Class 1 E functions. The presented methods of qualification are in accordance with IEEE Standards and other standards (IEC, JUS, DIN, etc,), and have the primary role to provide guidance for demonstrating the qualification of Class 1E equipment produced by 'Rade Koncar'.(author)

  12. IEEE Std 1205-1993: IEEE guide for assessing, monitoring, and mitigating aging effects on Class 1E equipment used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The guidelines are provided for assessing, monitoring, and mitigating degradation of Class 1E equipment used in nuclear power generating stations due to aging. The methods described can be used to identify the performance capability of Class 1E equipment beyond its qualified life. A discussion of stressors and aging mechanisms is included. If aging considerations have been satisfactorily addressed through other means (e.g., equipment qualification), then use of this guide may not be warranted. For some equipment, only partial application of this guide may be warranted

  13. Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457)

    International Nuclear Information System (INIS)

    1987-05-01

    In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986; the second supplement to NUREG-1002 was issued in October 1986. This third supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 2 was published. The facility is located in Reed Township, Will County, Illinois

  14. Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). Supplement No. 2

    International Nuclear Information System (INIS)

    1986-10-01

    In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986. This second supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 1 was published. The facility is located in Reed Township, Will County, Illinois

  15. Safety evaluation report related to the operation of Susquehanna Steam Electric Station, Units 1 and 2 (Docket Nos. 50-387 and 50-388). Suppl.6

    International Nuclear Information System (INIS)

    1984-03-01

    In April 1981, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0776) regarding the application of the Pennsylvania Power and Light Company (the applicant and/or licensee) and the Allegheny Electric Cooperative, Inc. (co-applicant) for licenses to operate the Susquehanna Steam Electric Station, Units 1 and 2, located on a site in Luzerne County, Pennsylvania. This supplement to NUREG-0776 addresses the remaining issues that required resolution before licensing operation of Unit 2 and closes them out

  16. Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). Supplement No. 1

    International Nuclear Information System (INIS)

    1986-09-01

    In November 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457). The facility is located in Reed Township, Will County, Illinois. This first supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time the Safety Evaluation Report was published

  17. Summary revaluation of energetic start-up of the unit 1 of nuclear power plant Mochovce

    International Nuclear Information System (INIS)

    Sarvaic, I.; Miskolci, M.

    1998-01-01

    The document contents stage revaluation of energetic start-up of the unit 1 of nuclear power plant Mochovce. Test results of the stage of energetic start-up are summarized in the document, valuation of important systems and block devices as well as fulfilling the operation limits and conditions has been performed. On that base conclusions and recommendations for start-up the unit 2 and for commercial operation of the unit 1 are elaborated. The valuation has been elaborated by a scientific management for start-up nuclear power plant Mochovce of nuclear safety of nuclear power facilities. Scientific management for start-up of nuclear power plant Mochovce performed continuous valuation of individual power levels after ending of each individual level and it gave its valuation to energy power level with recommendations and conditions for further start-up process and operation. Scientific management finished its activity at the unit 1 of nuclear power plant Mochovce according to a statute of scientific management for start-up after successful completion of conclusive block run. Scientific management group was founded in February 1998 at nuclear power plant Mochovce. Its members are experts from Slovak, Czech, Russian and French organizations which are participating in power plant completion. Members are listed in a supplement No. 2

  18. Insurance of nuclear power stations

    International Nuclear Information System (INIS)

    Debaets, M.

    1992-01-01

    Electrical utility companies have invested large sums in the establishment of nuclear facilities. For this reason it is normal for these companies to attempt to protect their investments as much as possible. One of the methods of protection is recourse to insurance. For a variety of reasons traditional insurance markets are unable to function normally for a number of reasons including, the insufficient number of risks, an absence of meaningful accident statistics, the enormous sums involved and a lack of familiarity with nuclear risks on the part of insurers, resulting in a reluctance or even refusal to accept such risks. Insurers have, in response to requests for coverage from nuclear power station operators, established an alternative system of coverage - insurance through a system of insurance pools. Insurers in every country unite in a pool, providing a net capacity for every risk which is a capacity covered by their own funds, and consequently without reinsurance. All pools exchange capacity. The inconvenience of this system, for the operators in particular, is that it involves a monopolistic system in which there are consequently few possibilities for the negotiation of premiums and conditions of coverage. The system does not permit the establishment of reserves which could, over time, reduce the need for insurance on the part of nuclear power station operators. Thus the cost of nuclear insurance remains high. Alternatives to the poor system of insurance are explored in this article. (author)

  19. Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro's Bruce nuclear generating station open-quotes Aclose quotes

    International Nuclear Information System (INIS)

    Day, J.E.; Baker, R.L.

    1994-01-01

    Ontario Hydro at the Bruce Nuclear Generating Station open-quotes Aclose quotes has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

  20. Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457)

    International Nuclear Information System (INIS)

    1988-06-01

    In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986; the second supplement was issued in October 1986; the third supplement was issued in May 1987; the fourth supplement was issued in July 1987 in support of the full power license for Unit 1; the fifth supplement was issued in December 1987 in support of the low power license for Unit 2. This sixth supplement to NUREG-1002 is in support of the full-power license for Unit 2 and provides the status of items that remained unresolved at the time Supplement 5 was published. The facility is located in Reed Township, Will County, Illinois