97-19931. Rochester Gas and Electric Corporation; R. E. Ginna Nuclear Power Plant; Environmental Assessment and Finding of No Significant Impact  

  • [Federal Register Volume 62, Number 145 (Tuesday, July 29, 1997)]
    [Notices]
    [Pages 40554-40555]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: 97-19931]
    
    
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    NUCLEAR REGULATORY COMMISSION
    
    [Docket No. 50-244]
    
    
    Rochester Gas and Electric Corporation; R. E. Ginna Nuclear Power 
    Plant; Environmental Assessment and Finding of No Significant Impact
    
        The U.S. Nuclear Regulatory Commission (the Commission) is 
    considering issuance of an exemption from certain requirements of its 
    regulations for Facility Operating License No. DRP-18 issued to 
    Rochester Gas and Electric Corporation (the licensee), for operation of 
    the R. E. Ginna Nuclear Power Plant located in Wayne County, New York.
    
    Environmental Assessment
    
    Identification of Proposed Action
    
        The proposed action would allow the licensee to utilize the 
    American Society of Mechanical Engineers Boiler and Pressure Vessel 
    Code (ASME Code) Case N-514, ``Low Temperature Overpressure 
    Protection'' to determine its low temperature overpressure protection 
    (LTOP) setpoints and is in accordance with the licensee's application 
    for exemption dated June 12, 1997. The proposed action requests an 
    exemption from certain requirements of 10 CFR 50.60, ``Acceptance 
    Criteria for Fracture Prevention Measures for Lightwater Nuclear Power 
    Reactors for Normal Operation,'' to allow application of an alternate 
    methodology to determine the LTOP setpoints for the R. E. Ginna Nuclear 
    Power Plant. The proposed alternate methodology is consistent with 
    guidelines developed by the ASME Working Group on Operating Plant 
    Criteria (WGOPC) to define pressure limits during LTOP events that 
    avoid certain unnecessary operational restrictions, provide adequate 
    margins against failure of the reactor pressure vessel, and reduce the 
    potential for unnecessary activation of pressure relieving devices used 
    for LTOP. These guidelines have been incorporated into Code Case N-514, 
    ``Low Temperature Overpressure Protection,'' which has been 
    incorporated into Appendix G of Section XI of the ASME Code and 
    published in the 1993 Addenda to Section XI. However, 10 CFR 50.55a, 
    ``Codes and Standards,'' and Regulatory Guide 1.147, ``Inservice 
    Inspection Code Case Acceptability'' have not been updated to reflect 
    the acceptability of Code Case N-514.
        The philosophy used to develop Code Case N-514 guidelines is to 
    ensure that the LTOP limits are still below the pressure/temperature 
    (P/T) limits for normal operation, but allow the pressure that may 
    occur with activation of pressure relieving devices to exceed the P/T 
    limits, provided acceptable margins are maintained during these events. 
    This philosophy protects the pressure vessel from LTOP events, and 
    still maintains the Technical Specifications P/T limits applicable for 
    normal heatup and cooldown in accordance with 10 CFR part 50, Appendix 
    G and Sections III and XI of the ASME Code.
    
    The Need for the Proposed Action
    
        Pursuant to 10 CFR 50.60, all lightwater nuclear power reactors 
    must meet the fracture toughness requirements for the reactor coolant 
    pressure boundary as set forth in 10 CFR part 50, Appendix G. Appendix 
    G of 10 CFR part 50 defines P/T limits during any condition of normal 
    operation including anticipated operational occurrences and system 
    hydrostatic tests, to which the pressure boundary may be subjected over 
    its service lifetime. It is specified in 10 CFR 50.60(b) that 
    alternatives to the described requirements in 10 CFR part 50, Appendix 
    G, may be used when an exemption is granted by the Commission under 10 
    CFR 50.12.
        To prevent transients that would produce excursions exceeding the 
    10 CFR part 50, Appendix G, P/T limits while the reactor is operating 
    at low temperatures, the licensee installed an LTOP system. The LTOP 
    system includes pressure relieving devices in the form of power-
    operated relief valves (PORVs) that are set at a pressure below the 
    LTOP enabling temperature that would prevent the pressure in the 
    reactor vessel from exceeding the P/T limits of 10 CFR Part 50, 
    Appendix G. To prevent these valves from lifting as a result of normal 
    operating pressure surges (e.g., reactor coolant pump (RCP) starting 
    and shifting operating charging pumps) with the reactor coolant system 
    in a solid water condition, the operating pressure must be maintained 
    below the PORV setpoint.
        In addition, to prevent damage to RCP seals, the operator must 
    maintain a minimum differential pressure across the RCP seals. Hence, 
    the licensee must operate the plant in a pressure window that is 
    defined as the difference between the minimum required pressure to 
    start a RCP and the operating margin to prevent lifting of the PORVs 
    due to normal operating pressure surges. 10 CFR part 50, Appendix G, 
    safety margin adds instrument uncertainty in the LTOP setpoint. The 
    licensee's current LTOP analysis indicates that using this 10 CFR part 
    50, Appendix G, safety margin to determine the PORV setpoint would 
    result in an operating window between the LTOP setpoint and the minimum 
    pressure required for RCP seals which is significantly restricted when 
    physical conditions such as PORV overshoot, RCP Ps, and static 
    head corrections are taken into account in setpoint determination. 
    Operating with these limits could result in the lifting of the PORVs or 
    damage to the RCP seals during normal operation. Using Code Case N-514 
    would allow the licensee to
    
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    recapture most of the operating margin that is lost by factoring in the 
    instrument uncertainties in the determination of the LTOP setpoint. The 
    net effect of using Code Case N-514 is that the setpoint will not 
    change significantly with the next setpoint analysis. Therefore, the 
    licensee proposed that in determining the setpoint for LTOP events for 
    Ginna, the allowable pressure be determined using the safety margins 
    developed in an alternate methodology in lieu of the safety margins 
    required by 10 CFR part 50, Appendix G. The alternate methodology is 
    consistent with the ASME Code Case N-514. The content of this Code Case 
    had been incorporated into Appendix G of Section XI of the ASME Code 
    and published in the 1993 Addenda to Section XI.
        An exemption from 10 CFR 50.60 is required to use the alternate 
    methodology for calculating the maximum allowable pressure for LTOP 
    considerations. By application dated June 12, 1997, the licensee 
    requested an exemption from 10 CFR 50.60 to allow it to utilize the 
    alternate methodology of Code Case N-514 to compute its LTOP setpoints.
    
    Environmental Impacts of the Proposed Action
    
        Appendix G of the ASME Code requires that the P/T limits be 
    calculated: (a) using a safety factor of two on the principal membrane 
    (pressure) stresses, (b) assuming a flaw at the surface with a depth of 
    one quarter (\1/4\) of the vessel wall thickness and a length of six 
    (6) times its depth, and (c) using a conservative fracture toughness 
    curve that is based on the lower bound of static, dynamic, and crack 
    arrest fracture toughness tests on material similar to the Ginna 
    reactor vessel material.
        In determining the PORV setpoint for LTOP events, the licensee 
    proposed the use of safety margins based on an alternate methodology 
    consistent with the proposed ASME Code Case N-514 guidelines. ASME Code 
    Case N-514 allows determination of the setpoint for LTOP events such 
    that the maximum pressure in the vessel will not exceed 110% of the P/T 
    limits of the existing ASME Appendix G. This results in a safety factor 
    of 1.8 on the principal membrane stresses. All other factors, including 
    assumed flaw size and fracture toughness, remain the same. Although 
    this methodology would reduce the safety factor on the principal 
    membrane stresses, use of the proposed criteria will provide adequate 
    margins of safety to the reactor vessel during LTOP transients.
        The change will not increase the probability or consequences of 
    accidents, no changes are being made in the types of any effluents that 
    may be released offsite, and there is no significant increase in the 
    allowable individual or cumulative occupational radiation exposure. 
    Accordingly, the Commission concludes that there are no significant 
    radiological environmental impacts associated with the proposed action.
        With regard to potential nonradiological impacts, the proposed 
    action does involve features located entirely within the restricted 
    area as defined in 10 CFR Part 20. It does not affect nonradiological 
    plant effluents and has no other environmental impact. Accordingly, the 
    Commission concludes that there are no significant nonradiological 
    environmental impacts associated with the proposed action.
    
    Alternatives to the Proposed Action
    
        Since the Commission has concluded there is no measurable 
    environmental impact associated with the proposed action, any 
    alternatives with equal or greater environmental impact need not be 
    evaluated. As an alternative to the proposed action, the staff 
    considered denial of the proposed action. Denial of the application 
    would result in no change in current environmental impacts. The 
    environmental impacts of the proposed action and the alternative action 
    are similar.
    
    Alternative Use of Resources
    
        This action does not involve the use of any resources not 
    previously considered in the ``Final Environmental Statement For the R. 
    E. Ginna Nuclear Power Plant'' dated December 1975.
    
    Agencies and Persons Consulted
    
        In accordance with its stated policy, on July 2, 1997, the staff 
    consulted with the Mr. Jack Spath of the New York State Energy Research 
    and Development Authority, regarding the environmental impact of the 
    proposed action. The State official had no comments.
    
    Finding of No Significant Impact
    
        Based upon the environmental assessment, the Commission concludes 
    that the proposed action will not have a significant effect on the 
    quality of the human environment. Accordingly, the Commission has 
    determined not to prepare an environmental impact statement for the 
    proposed action.
        For further details with respect to the proposed action, see the 
    licensee's letter dated June 12, 1997, which is available for public 
    inspection at the Commission's Public Document Room, which is located 
    at The Gelman Building, 2120 L Street, NW., Washington, DC, and at the 
    local public document room located at the Rochester Public Library, 115 
    South Avenue, Rochester, New York.
    
        Dated at Rockville, Maryland, this 23rd day of July 1997.
    
        For the Nuclear Regulatory Commission.
    Guy S. Vissing,
    Senior Project Manager, Project Directorate I-1, Division of Reactor 
    Projects I/II, Office of Nuclear Reactor Regulation.
    [FR Doc. 97-19931 Filed 7-28-97; 8:45 am]
    BILLING CODE 7590-01-P
    
    
    

Document Information

Published:
07/29/1997
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
97-19931
Pages:
40554-40555 (2 pages)
Docket Numbers:
Docket No. 50-244
PDF File:
97-19931.pdf