98-31931. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations  

  • [Federal Register Volume 63, Number 231 (Wednesday, December 2, 1998)]
    [Notices]
    [Pages 66590-66609]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: 98-31931]
    
    
    -----------------------------------------------------------------------
    
    NUCLEAR REGULATORY COMMISSION
    
    
    Biweekly Notice; Applications and Amendments to Facility 
    Operating Licenses Involving No Significant Hazards Considerations
    
    I. Background
    
        Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
    Commission (the Commission or NRC staff) is publishing this regular 
    biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
    Energy Act of 1954, as amended (the Act), to require the Commission to 
    publish notice of any amendments issued, or proposed to be issued, 
    under a new provision of section 189 of the Act. This provision grants 
    the Commission the authority to issue and make immediately effective 
    any amendment to an operating license upon a determination by the 
    Commission that such amendment involves no significant hazards 
    consideration, notwithstanding the pendency before the Commission of a 
    request for a hearing from any person.
        This biweekly notice includes all notices of amendments issued, or 
    proposed to be issued from November 6, 1998, through November 19, 1998. 
    The last biweekly notice was published on November 18, 1998 (63 FR 
    64106).
    
    Notice of Consideration of Issuance of Amendments to Facility 
    Operating Licenses, Proposed No Significant Hazards Consideration 
    Determination, and Opportunity for a Hearing
    
        The Commission has made a proposed determination that the following 
    amendment requests involve no significant hazards consideration. Under 
    the Commission's regulations in 10 CFR 50.92, this means that operation 
    of the facility in accordance with the proposed amendment would not (1) 
    involve a significant increase in the probability or consequences of an 
    accident previously evaluated; or (2) create the possibility of a new 
    or different kind of accident from any accident previously evaluated; 
    or (3) involve a significant reduction in a margin of safety. The basis 
    for this proposed determination for each amendment request is shown 
    below.
        The Commission is seeking public comments on this proposed 
    determination. Any comments received within 30 days after the date of 
    publication of this notice will be considered in making any final 
    determination.
        Normally, the Commission will not issue the amendment until the 
    expiration of the 30-day notice period.
    
    [[Page 66591]]
    
    However, should circumstances change during the notice period such that 
    failure to act in a timely way would result, for example, in derating 
    or shutdown of the facility, the Commission may issue the license 
    amendment before the expiration of the 30-day notice period, provided 
    that its final determination is that the amendment involves no 
    significant hazards consideration. The final determination will 
    consider all public and State comments received before action is taken. 
    Should the Commission take this action, it will publish in the Federal 
    Register a notice of issuance and provide for opportunity for a hearing 
    after issuance. The Commission expects that the need to take this 
    action will occur very infrequently.
        Written comments may be submitted by mail to the Chief, Rules and 
    Directives Branch, Division of Administration Services, Office of 
    Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
    20555-0001, and should cite the publication date and page number of 
    this Federal Register notice. Written comments may also be delivered to 
    Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
    Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
    written comments received may be examined at the NRC Public Document 
    Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
    filing of requests for a hearing and petitions for leave to intervene 
    is discussed below.
        By January 4, 1999, the licensee may file a request for a hearing 
    with respect to issuance of the amendment to the subject facility 
    operating license and any person whose interest may be affected by this 
    proceeding and who wishes to participate as a party in the proceeding 
    must file a written request for a hearing and a petition for leave to 
    intervene. Requests for a hearing and a petition for leave to intervene 
    shall be filed in accordance with the Commission's ``Rules of Practice 
    for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
    persons should consult a current copy of 10 CFR 2.714 which is 
    available at the Commission's Public Document Room, the Gelman 
    Building, 2120 L Street, NW., Washington, DC and at the local public 
    document room for the particular facility involved. If a request for a 
    hearing or petition for leave to intervene is filed by the above date, 
    the Commission or an Atomic Safety and Licensing Board, designated by 
    the Commission or by the Chairman of the Atomic Safety and Licensing 
    Board Panel, will rule on the request and/or petition; and the 
    Secretary or the designated Atomic Safety and Licensing Board will 
    issue a notice of a hearing or an appropriate order.
        As required by 10 CFR 2.714, a petition for leave to intervene 
    shall set forth with particularity the interest of the petitioner in 
    the proceeding, and how that interest may be affected by the results of 
    the proceeding. The petition should specifically explain the reasons 
    why intervention should be permitted with particular reference to the 
    following factors: (1) the nature of the petitioner's right under the 
    Act to be made a party to the proceeding; (2) the nature and extent of 
    the petitioner's property, financial, or other interest in the 
    proceeding; and (3) the possible effect of any order which may be 
    entered in the proceeding on the petitioner's interest. The petition 
    should also identify the specific aspect(s) of the subject matter of 
    the proceeding as to which petitioner wishes to intervene. Any person 
    who has filed a petition for leave to intervene or who has been 
    admitted as a party may amend the petition without requesting leave of 
    the Board up to 15 days prior to the first prehearing conference 
    scheduled in the proceeding, but such an amended petition must satisfy 
    the specificity requirements described above.
        Not later than 15 days prior to the first prehearing conference 
    scheduled in the proceeding, a petitioner shall file a supplement to 
    the petition to intervene which must include a list of the contentions 
    which are sought to be litigated in the matter. Each contention must 
    consist of a specific statement of the issue of law or fact to be 
    raised or controverted. In addition, the petitioner shall provide a 
    brief explanation of the bases of the contention and a concise 
    statement of the alleged facts or expert opinion which support the 
    contention and on which the petitioner intends to rely in proving the 
    contention at the hearing. The petitioner must also provide references 
    to those specific sources and documents of which the petitioner is 
    aware and on which the petitioner intends to rely to establish those 
    facts or expert opinion. Petitioner must provide sufficient information 
    to show that a genuine dispute exists with the applicant on a material 
    issue of law or fact. Contentions shall be limited to matters within 
    the scope of the amendment under consideration. The contention must be 
    one which, if proven, would entitle the petitioner to relief. A 
    petitioner who fails to file such a supplement which satisfies these 
    requirements with respect to at least one contention will not be 
    permitted to participate as a party.
        Those permitted to intervene become parties to the proceeding, 
    subject to any limitations in the order granting leave to intervene, 
    and have the opportunity to participate fully in the conduct of the 
    hearing, including the opportunity to present evidence and cross-
    examine witnesses.
        If a hearing is requested, the Commission will make a final 
    determination on the issue of no significant hazards consideration. The 
    final determination will serve to decide when the hearing is held.
        If the final determination is that the amendment request involves 
    no significant hazards consideration, the Commission may issue the 
    amendment and make it immediately effective, notwithstanding the 
    request for a hearing. Any hearing held would take place after issuance 
    of the amendment.
        If the final determination is that the amendment request involves a 
    significant hazards consideration, any hearing held would take place 
    before the issuance of any amendment.
        A request for a hearing or a petition for leave to intervene must 
    be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
    Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
    Adjudications Staff, or may be delivered to the Commission's Public 
    Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
    by the above date. A copy of the petition should also be sent to the 
    Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
    Washington, DC 20555-0001, and to the attorney for the licensee.
        Nontimely filings of petitions for leave to intervene, amended 
    petitions, supplemental petitions and/or requests for a hearing will 
    not be entertained absent a determination by the Commission, the 
    presiding officer or the Atomic Safety and Licensing Board that the 
    petition and/or request should be granted based upon a balancing of 
    factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
        For further details with respect to this action, see the 
    application for amendment which is available for public inspection at 
    the Commission's Public Document Room, the Gelman Building, 2120 L 
    Street, NW., Washington, DC, and at the local public document room for 
    the particular facility involved.
    
    Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414, 
    Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
    
        Date of amendment request: November 11, 1998.
    
    [[Page 66592]]
    
        Description of amendment request: The proposed amendments would 
    revise the Technical Specifications (TS) to correct Surveillance 
    Requirements (SRs) 3.6.11.6 and 3.6.11.7 and the associated Bases. 
    These SRs currently are incorrect and do not reflect the Containment 
    Pressure Control System (CPCS) as designed. Therefore, the proposed 
    amendments would only revise the SRs; no change to the CPCS design is 
    involved.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
    First Standard
    
        Implementation of this amendment would not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated. Approval of this amendment will have no 
    significant effect on accident probabilities or consequences.
        The CPCS is not an accident initiating system; therefore, there 
    will be no impact on any accident probabilities by the approval of 
    this amendment. The design of the CPCS is not being modified by this 
    proposed amendment. The amendment merely aligns [TS] surveillance 
    requirements with the existing design and function of the system. 
    Therefore, there will be no impact on any accident consequences.
    
    Second Standard
    
        Implementation of this amendment would not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated. No new accident causal mechanisms are created 
    as a result of NRC approval of this amendment request. No changes 
    are being made to the plant which will introduce any new accident 
    causal mechanisms. This amendment request does not impact any plant 
    systems that are accident initiators, since the CPCS is an accident 
    mitigating system.
    
    Third Standard
    
        Implementation of this amendment would not involve a significant 
    reduction in a margin of safety. Margin of safety is related to the 
    confidence in the ability of the fission product barriers to perform 
    their design functions during and following an accident situation. 
    These barriers include the fuel cladding, the reactor coolant 
    system, and the containment system. The performance of these fission 
    product barriers will not be impacted by implementation of this 
    proposed amendment. The CPCS is already capable of performing as 
    designed. No safety margins will be impacted.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: York County Library, 138 East 
    Black Street, Rock Hill, South Carolina.
        Attorney for licensee: Mr. Paul R. Newton, Legal Department 
    (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
    North Carolina.
        NRC Project Director: Herbert N. Berkow.
    
    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
    Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
    
        Date of amendment request: October 15, 1998.
        Description of amendment request: The proposed amendments would 
    revise the pressure-temperature limits in the Technical Specifications 
    for Units 1, 2, and 3. The proposed amendments would revise the heatup, 
    cooldown, and inservice test limitations for the reactor coolant system 
    of each unit to a maximum of 26 effective full-power years. The 
    proposed amendments would also revise the Technical Specification for 
    low temperature overpressure protection to reflect the revised 
    pressure-temperature limits of the reactor vessels.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        A. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated?
        NO.
        Each accident analysis addressed in the Oconee UFSAR [Updated 
    Final Safety Analysis Report] has been examined with respect to the 
    changes to the Reactor Pressure Vessel (RPV) pressure-temperature 
    limit curves and related Low Temperature Overpressure settings. The 
    probability of any design basis accident (DBA) is not affected by 
    this change, nor are the consequences of a DBA affected by this 
    change. The revised pressure-temperature limits, which were 
    developed based on NRC approved methodology or ASME Code [American 
    Society of Mechanical Engineers Boiler and Pressure Vessel Code] 
    Case N-514 as described in the Technical Justification, are not 
    considered to be an initiator or contributor to any accident 
    analysis addressed in the Oconee UFSAR. The added requirement to 
    deactivate one pressurizer heater bank during low temperature 
    operation does not significantly change the probability or 
    consequence of any accident previously analyzed. No existing 
    Technical Specification requirements are being deleted with this 
    revision.
        B. Create the possibility of a new or different kind of accident 
    from the accident previously evaluated?
        NO.
        This license amendment revises Oconee RPV pressure-temperature 
    limits. The revised pressure-temperature limits were developed based 
    on NRC approved methodology or ASME Code Case N-514 as described in 
    the Technical Justification. Operation of Oconee in accordance with 
    these proposed new Technial Specifications will not create any 
    failure modes not bounded by previously evaluated accidents. 
    Consequently, this change will not create the possibility of a new 
    or different accident from any accident previously evaluated.
        C. Involve a significant reduction in a margin of safety?
        NO.
        This license amendment revises Oconee RPV pressure-temperature 
    limits. The revised pressure-temperature limits were developed based 
    on NRC approved methodology or ASME Code Case N-514 as described in 
    the Technical Justification. The purpose of this license amendment 
    is to assure that sufficient operating margin to safety is 
    maintained in the operation of the Oconee reactor pressure vessels 
    by establishing new, more limiting pressure-temperature limit curves 
    and adding the requirement to deactivate one pressurizer heater 
    bank. No plant safety limits, set points, or design parameters are 
    adversely affected. The fuel, fuel cladding, and Reactor Coolant 
    System are not impacted. Therefore, there will be no significant 
    reduction in any margin of safety.
        Duke [Duke Energy Corporation] has concluded based on this 
    information that there are no significant hazards considerations 
    involved in this amendment request.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Oconee County Library, 501 
    West South Broad Street, Walhalla, South Carolina.
        Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
    1200 17th Street, NW., Washington, DC.
        NRC Project Director: Herbert N. Berkow. Duquesne Light Company, et 
    al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1, 
    Shippingport, Pennsylvania
    
        Date of amendment request: November 11, 1998.
        Description of amendment request: The proposed amendment would 
    modify License Condition 2.C(9) to allow, on a one time only basis, an 
    extension to the steam generator inspection interval of technical 
    specification surveillance 4.4.5.3.b. This
    
    [[Page 66593]]
    
    would allow the steam generator inspection interval to coincide with 
    the 13th refueling outage or the end of 500 effective full power days, 
    whichever occurs sooner.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        1. Does the change involve a significant increase in the 
    probability or consequences of an accident previously evaluated?
        The proposed change is temporary and allows a one time extension 
    of specific surveillance requirements for Cycle 13 to allow 
    surveillance testing to coincide with the 13th (1R13) refueling 
    outage. The proposed surveillance interval extension will not cause 
    a significant reduction in system reliability nor affect the ability 
    of a system to perform its design function. Current monitoring of 
    plant conditions and the surveillance monitoring required during 
    normal plant operation will be performed as usual to assure 
    conformance with technical specification operability requirements.
        The technical specification steam generator tube inspection is 
    intended to prevent the Steam Generator Tube Rupture analyzed in 
    [Updated Final Safety Analysis Report] UFSAR Section 14.2.4 by 
    maintenance of the integrity of the primary to secondary coolant 
    boundary represented by steam generator tubes. The process by which 
    this integrity is maintained is inspection of steam generator tubes 
    at prescribed intervals, and the removal of defective tubes from 
    service. Inspection intervals are based on preventing corrosion 
    growth from exceeding tube structural limits, thereby preventing 
    tube failure. The 1997 steam generator inspection characterized 
    existing steam generator tube degradation, and degraded tubes were 
    removed from service at that time. Degradation growth rates were 
    evaluated for the next operating interval and it was determined that 
    the steam generator tube structural integrity is maintained. 
    Degradation of steam generator tubes was prevented during the 
    extended outage by a carefully controlled, corrosion prevention 
    program.
        The proposed change does not affect the UFSAR and is consistent 
    with changes granted for other plants. The surveillance extension 
    does not involve a change to plant equipment and does not affect the 
    performance of plant equipment used to mitigate an accident. This 
    change, therefore, does not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        2. Does the change create the possibility of a new or different 
    kind of accident from any accident previously evaluated?
        Extending the surveillance interval for the performance of 
    specific inspections will not create the possibility of any new or 
    different kind of accidents. No change is required to any system 
    configurations, plant equipment or analyses.
        Steam generator tube inspections determine tube integrity and 
    provide reasonable assurance that a tube rupture or primary to 
    secondary leak will not occur. Accidents involving steam generator 
    tube rupture are analyzed in UFSAR Section 14.2.4, ``Steam Generator 
    Tube Rupture.'' The only type of accident that can be postulated 
    from extending the steam generator inspection interval would be a 
    tube leak or rupture which are analyzed in the UFSAR. No new failure 
    modes are created by the surveillance extension. Therefore, this 
    change will not create the possibility of a new or different kind of 
    accident from any accident previously evaluated.
        3. Does the change involve a significant reduction in a margin 
    of safety?
        Surveillance interval extensions will not impact any plant 
    safety analyses since the assumptions used will remain unchanged. 
    The safety limits assumed in the accident analyses and the design 
    function of the equipment required to mitigate the consequences of 
    any postulated accidents will not be changed since only the 
    surveillance interval is being extended. Based on engineering 
    judgement, extending the surveillance interval for the performance 
    of these specific inspections does not involve a significant 
    reduction in the margin of safety derived from the required 
    surveillances.
        The margin of safety depends upon maintenance of specific 
    operating parameters within design limits. In the case of steam 
    generators, that margin is maintained through assurance of tube 
    integrity as the primary to secondary boundary. Assurance of tube 
    integrity is provided through periodic in-service inspection of 
    tubes and removal of defective tubes from service. Additional margin 
    is provided through protection from possible consequences of steam 
    generator tube failure by mitigation systems. Radiation monitors 
    provide a detection capability of primary to secondary leakage to 
    enable a prompt response. Maintenance of the steam generator water 
    chemistry in accordance with [Electric Power Research Institute] 
    EPRI guidelines provides additional margin of safety. Therefore, the 
    plant will be maintained within the analyzed limits and the proposed 
    extension will not significantly reduce the margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: B. F. Jones Memorial Library, 
    663 Franklin Avenue, Aliquippa, PA 15001.
        Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
    Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
        NRC Project Director: Robert A. Capra Entergy Operations, Inc., 
    Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, 
    Arkansas
        Date of amendment request: June 30, 1998.
        Description of amendment request: The proposed change modifies the 
    Engineered Safety Features Actuation System (ESFAS) portion of the 
    Arkansas Nuclear One, Unit-2 (ANO-2) Plant Protection System (PPS). 
    This modification is designed to defeat the backup power supply for the 
    auctioneered power sources for channel A and D Reactor Protective 
    System (RPS) and ESFAS bistables, and to provide selective logic for 
    Emergency Feedwater Actuation Signals and Main Steam Isolation Signals. 
    This will ensure that ESFAS will have the redundancy and independence 
    sufficient to assure that (1) no single failure results in loss of the 
    protection function with a channel in indefinite bypass, and (2) 
    removal from service of any component or channel does not result in 
    loss of the required minimum redundancy required by the ANO-2 Technical 
    Specifications (TSs). The proposed modification to the ANO-2 PPS has 
    been determined to involve an Unreviewed Safety Question in accordance 
    with 10 CFR 50.59(a)(2).
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        An evaluation of the proposed change has been performed in 
    accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
    considerations using the standards in 10 CFR 50.92(c). A discussion 
    of these standards as they relate to this amendment request follows:
        Criterion 1--Does Not Involve a Significant Increase in the 
    Probability or Consequences of an Accident Previously Evaluated.
        The ANO-2 Plant Protection System (PPS) includes the electrical 
    and mechanical devices and circuitry (from sensors to actuation 
    device input terminals) involved in generating signals associated 
    with the two protective functions, Engineered Safety Feature 
    Actuation System (ESFAS) and Reactor Protective System (RPS). The 
    RPS is that portion of the PPS which generates signals that actuate 
    a reactor trip. The ESFAS is that portion of the PPS which generates 
    signals that actuate Engineered Safety Features (ESF) to mitigate 
    the consequences of an accident.
        The ANO-2 Safety Analysis Report (SAR) section 15.1.31 ``Loss Of 
    One DC System'' analyzes failure of a DC bus (FODCB) as initiator 
    and its causes. The causes for the FODCB are DC leg to leg fault in 
    the bus or in the power distribution circuit from the battery. Since 
    the proposed change has no impact on the accident initiator, the 
    frequency of occurrence is not changed. In order for the FODCB as a 
    single failure with an accident to de-energize two [Vital Instrument 
    Buses (]VIBs[)], the FODCB would have to occur prior to the safety 
    bus
    
    [[Page 66594]]
    
    energization by offsite bus fast transfer or prior to safety bus 
    energization by the emergency diesel generator (EDG). The potential 
    for de-energization of one pair of VIBs is, therefore, limited to 
    time from initiation of the accident to time for safety bus response 
    to the secondary plant and Reactor Protective System trips.
        The effects of the FODCB are being revised to assume a secondary 
    plant trip that results in de-energization of one power division. 
    The existing analysis conclusions remain unchanged. The accident 
    analysis is being revised to include de-energization of a pair of 
    vital AC instrument channels. De-energization of two vital AC 
    sources has not been previously documented as a design bases event.
        Auctioneered bistable power supplies for Plant Protection System 
    (PPS) channels A and D are being modified to a single power source 
    for each of these two channels. Single channel trips will result for 
    all PPS functions in channels A or D for loss of its single channel 
    bistable power source. The PPS channels B and C auctioneered power 
    supplies remain unchanged to maintain Recirculation Actuation Signal 
    (RAS) response to a FODCB.
        Regarding PPS measurement channels with increasing signal 
    setpoints, de-energization of a single power supply either results 
    in failure of a measurement channel (B or C) to a non-tripped state 
    or in failure of a measurement channel (A or D) to a tripped state. 
    Neither single channel failure scenario impacts accident initiation 
    or mitigation. For PPS measurement channels with decreasing signal 
    setpoints the single channel de-energization events result in 
    failure of a single affected measurement channel to a tripped state. 
    The PPS two out of three logic design with a channel bypassed 
    ensures operability with a single channel failure. Neither condition 
    impacts accident frequency or consequences.
        With the exception of Recirculation Actuation Signal (RAS) and 
    Emergency Feedwater Actuation Signal (EFAS), a FODCB results in an 
    automatic ESFAS initiation for those functions with decreasing 
    signal setpoints. For other ESFAS functions with a decreasing 
    signal, channels A and C or channels B and D fail to the tripped 
    state. For those functions with an increasing signal setpoint 
    (including EFAS), a FODCB results in a single channel failing not 
    tripped, one channel tripping, and two channels remaining 
    functional. System level functions remain operable with either a one 
    out of two logic (no channels bypassed) or a one out of one logic 
    (with a channel bypassed).
        Interposing relay actuation logic has changed from single trip 
    path to selective trip path logic. This change insures emergency 
    feedwater (EFW) discharge valves will receive an automatic open or 
    close demand based on steam generator level and pressure demands.
        Therefore, this change does not involve a significant increase 
    in the probability or consequences of any accident previously 
    evaluated.
        Criterion 2--Does Not Create the Possibility of a New or 
    Different Kind of Accident from any Previously Evaluated.
        In response to de-energization of a pair of Vital Instrument 
    Buses (VIBs), those ESFAS functions with increasing signal 
    setpoints, as a minimum, remain functional with one out of one 
    logic. One channel trips, one channel does not trip, and two 
    channels remain functional. One of the functional channels may be 
    bypassed without impact on operability. The trip response of those 
    ESFAS functions with decreasing signal to trip setpoints remains 
    unchanged.
        EFAS coincidence logic to close the EFW discharge valves 
    requires three out of four channels to be in a non-tripped state. 
    With a FODCB one channel is tripped, one channel is not tripped, and 
    two channels are functional. The close logic becomes two out of two 
    with a FODCB.
        By defeating the auctioneered bistable power sources for PPS 
    channel A and D bistables, PPS measurement channel A or D will fail 
    to its tripped state. This change ensures no more than one channel 
    (B or C) fails to a non-tripped state for the FODCB.
        With selective logic EFAS pump discharge valves will receive 
    control signals to initiate emergency feedwater and to terminate 
    emergency feedwater flow by open and close demands generated 
    independent of the 120 Volt channel pair de-energization.
        The existing ANO-2 Failure Modes and Effects Analysis does not 
    document failure of a pair of vital instrument AC channels. Neither 
    the 120 Volts AC nor the 125 Volt DC system single failure analysis 
    assumes failure of two channels of 120 Volts AC. Even though the 
    failure of either pair of VIBs caused by a FODCB is not a result of 
    the proposed change, the SAR change will address the potential for 
    de-energization of a pair of instrument buses. The ANO-2 SAR will be 
    updated to reflect the documentation and modification of the PPS 
    design to ensure safe plant response.
        Even though the plant response to FODCB is being modified, the 
    proposed ANO-2 PPS design resolution does not create the possibility 
    of a new or different kind of accident from any previously evaluated 
    in the SAR. The PPS will have the redundancy and independence 
    sufficient to assure that (1) no single failure results in loss of 
    the protection function, and (2) removal from service of any 
    component or channel does not result in loss of the required minimum 
    redundancy required by the TS. PPS will also meet the single failure 
    criterion of IEEE 279-1971 to the extent that any single failure 
    within the system does not prevent proper protective action at the 
    system level and no single failure will defeat more than one of the 
    four protective channels associated with any one trip function.
        Criterion 3--Does Not Involve a Significant Reduction in the 
    Margin of Safety.
        Technical Specification Bases 3/4.3.1 & 3/4.3.2 assure 
    sufficient PPS redundancy is maintained to permit a channel to be 
    bypassed. Under the current design, a FODCB will result in reduction 
    of margin by decreasing the number of functional channels to less 
    than two. However, with the proposed modification removal from 
    service of any component or channel for indefinite bypass will not 
    result in loss of the minimum redundancy required by the TS. This 
    activity will restore the margin by ensuring ESFAS required 
    functions remain capable of automatic actuation with a FODCB.
        Therefore, this change does not involve a significant reduction 
    in the margin of safety.
        Based upon the reasoning presented above and the previous 
    discussion of the amendment request, Entergy Operations has 
    determined that even though the proposed PPS design description 
    results in an accident or malfunction of a different type, the 
    requested change does not involve a significant hazards 
    consideration.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Tomlinson Library, Arkansas 
    Tech University, Russellville, AR 72801.
        Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
    Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
        NRC Project Director: John N. Hannon.
    
    Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
    Plant, Unit No. 1, St. Lucie County, Florida
    
        Date of amendment request: October 29, 1998.
        Description of amendment request: The proposed amendment would 
    revise the terminology used in the St. Lucie Plant Technical 
    Specifications (TS) relative to the implementation and automatic 
    removal of certain reactor protection system trip bypasses to ensure 
    that the meaning of explicit terms used in the TS are consistent with 
    the intent of the stated requirements.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        (1) Operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        The proposed amendments are administrative in nature, and do not 
    change the function or the setpoints of the RPS trip bypass 
    features. The revisions simply make corrections to the Notation of 
    TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit 
    terms used in the Notes is consistent with the intent of the stated 
    requirements based on the St. Lucie plant design. The proposed 
    technical specification changes do not involve accident initiators, 
    do not change the configuration or method of operation of any plant 
    equipment that is used to mitigate
    
    [[Page 66595]]
    
    the consequences of an accident, and do not alter any conditions 
    assumed in the plant accident analyses. Therefore, operation of 
    either facility in accordance with its proposed amendment would not 
    involve a significant increase in the probability or consequences of 
    an accident previously evaluated.
        (2) Operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        The proposed amendments are administrative in nature and will 
    not change the physical plant or the modes of plant operation 
    defined in the facility operating licenses. The changes do not 
    involve the addition or modification of equipment nor do they alter 
    the design or operation of plant systems. Therefore, operation of 
    either facility in accordance with its proposed amendment would not 
    create the possibility of a new or different kind of accident from 
    any accident previously evaluated.
        (3) Operation of the facility in accordance with the proposed 
    amendment would not involve a significant reduction in a margin of 
    safety.
        The proposed amendments are administrative in nature and do not 
    change the function or the setpoints of the RPS trip bypass 
    features. The revisions simply make corrections to the Notation of 
    TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit 
    terms used in the Notes is consistent with the intent of the stated 
    requirements based on the St. Lucie plant design. The proposed 
    changes do not alter the basis for any technical specification that 
    is related to the establishment of, or the maintenance of, a nuclear 
    safety margin. Therefore, operation of either facility in accordance 
    with its proposed amendment would not involve a significant 
    reduction in a margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration. This 
    notice is intended to replace an exigent notice of consideration of 
    issuance of amendment for St. Lucie Unit 1, previously published as 
    exigent TS amendments for both St. Lucie Units 1 and 2 in the Federal 
    Register (63 FR 59809). The amendment request for St. Lucie Unit 2 will 
    continue to be considered as an exigent amendment as noticed in the 
    Federal Register (63 FR 59809).
        Local Public Document Room location: Indian River Junior College 
    Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
        Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
    P.O. Box 14000, Juno Beach, Florida 33408-0420.
        NRC Project Director: Frederick J. Hebdon.
    
    GPU Nuclear, Inc, et al., Docket No. 50-219, Oyster Creek Nuclear 
    Generating Station, Ocean County, New Jersey
    
        Date of amendment request: November 10, 1998.
        Description of amendment request: The proposed Technical 
    Specification (TS) change would remove the restriction on the sale or 
    lease of property within the exclusion area and replace the restriction 
    with a requirement to retain complete authority to determine and 
    maintain sufficient control of all activities including the authority 
    to exclude or remove personnel and property within the minimum 
    exclusion distance. A TS Bases page for the proposed change is 
    included. Also included are clarifications and administrative changes 
    which (1) clarify TS definition 1.38 to become ``Site Boundry'' from 
    the current term ``Exclusion Area'' to be consistent with 10 CFR 
    20.1003 definition for Site Boundry and the 10 CFR 100.3 definition of 
    Exclusion Area, (2) convert the one occurrence of the use of TS 
    definition from Exclusion Area to Site Boundry in TS 6.8.4(a)(9), and 
    (3) revise and update the Table of Contents for Section I Definitions.`
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        1. Would operation of the facility in accordance with the 
    proposed change involve a significant increase in the probability or 
    consequences of an accident previously evaluated?
        The proposed change is administrative in nature and does not 
    affect the purpose, function, performance, operability or testing of 
    and does not make any physical or procedural changes to plant 
    systems, structures or components. Also, all existing technical 
    specification limiting conditions for operation and surveillance 
    requirements are retained.
        [Technical Specification Change Request] TSCR 264 does not 
    change the size or location of the exclusion area. Since the 
    exclusion area size and location are not being changed and no 
    physical or procedural changes are being made to the plant, 
    radiological consequences in the exclusion area are not affected by 
    this TSCR.
        This change addresses the existing technical specification 
    restriction on the sale or lease of property within the ``exclusion 
    area'' by ensuring that the licensee will retain at all times the 
    complete authority to determine and maintain sufficient control of 
    all activities through ownership, easement, contract and/or other 
    legal instruments on property within the minimum exclusion distance 
    including the authority to exclude or remove personnel and property 
    within the minimum exclusion distance.
        Therefore, since no physical or procedural changes are being 
    made to existing plant systems, structures or components and since 
    the proposed change requires the licensee to retain complete 
    authority and sufficient control of all activities in the exclusion 
    area, operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        2. Would operation of the facility in accordance with the 
    proposed change create the possibility of a new or different kind of 
    accident from any accident previously evaluated?
        The p[ro]posed change is administrative in nature and does not 
    affect the purpose, function, performance, operability or testing of 
    and does not make any physical or procedural changes to plant 
    systems, structures or components. Also, all existing technical 
    specification limiting conditions for operation and surveillance 
    requirements are retained.
        This change addresses the existing technical specification 
    restriction on the sale or lease of property within the ``exclusion 
    area'' by ensuring that the licensee will retain at all times the 
    complete authority to determine and maintain sufficient control of 
    all activities through ownership, easement, contract and/or other 
    legal instruments on property within the minimum exclusion distance 
    including the authority to exclude or remove personnel and property 
    within the minimum exclusion distance.
        Therefore, since no physical or procedural changes are being 
    made to existing plant systems, structures or components and since 
    the proposed change requires the licensee to retain complete 
    authority and sufficient control of all activities in the exclusion 
    area, operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        3. Would operation of the facility in accordance with the 
    proposed change involve a significant reduction in a margin of 
    safety?
        The p[ro]posed change is administrative in nature and does not 
    affect the purpose, function, performance, operability or testing of 
    and does not make any physical or procedural changes to plant 
    systems, structures or components. Also, all existing technical 
    specification limiting conditions for operation and surveillance 
    requirements are retained.
        This change addresses the existing technical specification 
    restriction on the sale or lease of property within the ``exclusion 
    area'' by ensuring that the licensee will retain at all times the 
    complete authority to determine and maintain sufficient control of 
    all activities through ownership, easement, contract and/or other 
    legal instruments on property within the minimum exclusion distance 
    including the authority to exclude or remove personnel and property 
    within the minimum exclusion distance.
        Therefore, since no physical or procedural changes are being 
    made to existing plant
    
    [[Page 66596]]
    
    systems, structures or components and since the proposed change 
    requires the licensee to retain complete authority and sufficient 
    control of all activities in the exclusion area, operation of the 
    facility in accordance with the proposed amendment will not involve 
    a significant reduction in margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Ocean County Library, 
    Reference Department, 101 Washington Street, Toms River, NJ 08753.
        Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
    Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
        NRC Project Director: Cecil O. Thomas.
    
    Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
    Nuclear Station, Unit 2 (NMP2), Oswego County, New York
    
        Date of amendment request: October 16, 1998.
        Description of amendment request: The proposed amendment would make 
    the following revisions to Technical Specifications (TSs) 3/4.7.1.1: 
    (1) Ensure that four service water (SW) pumps are operating with the 
    divisional cross connect valves open during Operational Condition 1, 2 
    and 3 (current TS requires two SW pumps associated with one loop to be 
    operating); (2) Increase the number of division 1 and 2 heaters 
    required to be operable from 7 per division per intake to 14 per 
    division per intake; (3) The actions necessary for having less than the 
    required equipment is being revised to reflect the new limits for SW 
    equipment; and (4) SW supply header discharge water temperature is 
    being increased from 81 to 82  deg.F. TS 3.7.1.2, Table 3.3.9-1, and 
    Table 4.3.9.1-1 are revised to add ``when handling irradiated fuel in 
    the secondary containment'' to the applicability section. Table 3.3.9-1 
    is being revised to decrease the temperature at which the Intake 
    Deicing Heaters are required to be in service from 39 to 38 degrees F. 
    TS 3.7.1.2 proposed change is to specify that the necessary portions of 
    the SW system needed to support equipment required to be operable shall 
    be operable; the Action Section proposed revision reflects this change. 
    TS 4.7.1.2.1 surveillance requirement proposed change is to increase 
    the flow rate of SW pumps from 6500 GPM to 9000 GPM and to change the 
    SW pumps pressure from 80 psi discharge pressure to 70 psi differential 
    pressure; TS 4.7.1.2.2 is being revised to decrease the intake tunnel 
    water temperature from 39 to 38 degrees F. The surveillance for the 
    Intake Deicing Heaters is being changed to reflect the increase in the 
    number of heaters required. The title of ``Plant Service Water System'' 
    is being changed to ``Service Water System.''
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        The operation of Nine Mile Point Unit 2, in accordance with the 
    proposed amendment, will not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        The SW System is a once-through system which supplies water from 
    Lake Ontario to various essential and non-essential components, as 
    required, during normal plant operation and shutdown conditions. The 
    System is designed with suitable redundancy to provide a reliable 
    source of cooling water for the removal of heat from essential plant 
    components, including the RHR [residual heat removal] heat 
    exchangers, the EDGs [emergency diesel generators], and room coolers 
    for ECCS [emergency core cooling system] equipment, which are 
    required for safe reactor shutdown following a LOCA.
        LCO 3.7.1.1 and LCO 3.7.1.2 each currently requires two 
    independent SW System loops to be operable, with one of the loops in 
    operation. The current LCOs do not provide adequate guidance 
    regarding the minimum number of operating pumps. NMPC [Niagara 
    Mohawk Power Corporation] proposes to revise LCO 3.7.1.1 and its 
    associated Actions and SRs to provide assurance that four SW pumps 
    are operable and are operating within acceptable system parameters, 
    with the divisional cross-connect valves open, during Operational 
    Conditions 1, 2, and 3 to meet the limiting LOCA analysis 
    assumptions.
        TS Section 3/4.7.1 currently specifies a maximum SW supply 
    header discharge water temperature of 81 degrees F and a limiting 
    temperature for Intake Deicing Heater ystem operability (intake 
    water) temperature of 39 degrees F. In addition, TS Table 3.3.9-1, 
    Action 144, requires the Intake Deicing Heater System heaters to be 
    placed in service when the Lake Ontario water temperature reaches 39 
    degrees F. NMPC proposes to revise Action 144 of TS Table 3.3.9-1 
    and TS LCO 3.7.1.1, including its associated Actions and SRs 
    [surveillance requirements], to increase the supply header discharge 
    water temperature to its analytical limit of 82 degrees F and reduce 
    the limiting temperature for the Intake Deicing Heater System Action 
    and operability requirements to 38 degrees F.
        Appropriate changes to LCO 3.7.1.2 and its associated Actions 
    and SRs are also proposed in order to assure consistency with the SW 
    System analyses assumptions during shutdown conditions. The current 
    LCO Actions do not account for the varying flows and heat loads that 
    may be required for various plant shutdown conditions. The revision 
    to the Applicability for LCO 3.7.1.2 and TS Tables 3.3.9-1 and 
    4.3.9.1-1 will assure that the SW System is operable during periods 
    when irradiated fuel is being handled in the secondary containment 
    and essential loads cooled by the SW System are required to be 
    operable (e.g., EDG). A footnote has been added to define 
    Operational Condition * and is consistent with similar footnotes in 
    the TSs. The proposed changes will assure that the necessary ortions 
    of the SW System and the necessary Divisions of the Intake Deicing 
    Heater System heaters are operable that are supporting equipment 
    required to be operable.
        It is further proposed to change the system title identified in 
    the Index and in TS Section 3/4.7.1, including the LCOs and SRs, 
    from ``Plant Service Water System'' to ``Service Water System'' to 
    be consistent with the NMP2 [Nine Mile Unit 2] UFSAR [Updated Final 
    Safety Analysis Report].
        The changes do not involve any physical alteration of the plant, 
    and the SW System will remain capable of providing sufficient 
    cooling flow for the essential cooling loads during plant operation 
    and also during plant shutdown. The changes will have no impact on 
    the design or function of the SW System and its components, thus 
    assuring that the characteristics and functional performance are 
    maintained consistent with the event precursors and the conditions 
    and assumptions of the current design basis accident and transient 
    analyses. The changes to the LCO AOTs [allowed outage times] are 
    either consistent with or are more conservative than the current 
    AOTs. Based on the above, adequate assurance is provided that the 
    probability of event initiation will remain as previously analyzed. 
    Maintaining four pumps operating within acceptable system 
    parameters, with the divisional cross connect valves open, during 
    Operational Conditions 1, 2, and 3 provides assurance that the 
    essential functions supported by the SW System are maintained. 
    Particularly, adequate SW flow assures that the primary and 
    secondary containments can perform their intended functions of 
    limiting the release of radioactive materials to the environment 
    following a LOCA. The small (1 degree F) change in the SW supply 
    header discharge water (UHS) temperature and Intake Deicing Heater 
    System actuation temperature maintain the current design basis for 
    the UHS and SW Systems such that there will be no impact on the LOCA 
    analyses assumptions or conclusions. The proposed changes to the SW 
    System TSs do not adversely affect the capability of plant systems, 
    structures, and components to respond to any accident in Operational 
    Conditions 4, 5, and *. As a result, there will be no degradation of 
    the primary or secondary containment or any other fission product 
    barriers which could increase the radiological consequences of an 
    accident. In addition, other essential accident mitigation equipment 
    supported by the SW System will not be adversely impacted. It is, 
    therefore,
    
    [[Page 66597]]
    
    concluded that operation of NMP2, in accordance with the proposed 
    amendment, will not involve a significant increase in the 
    probability or consequences of an accident previously evaluated. The 
    operation of Nine Mile Point Unit 2, in accordance with the proposed 
    amendment, will not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        The changes do not result in any hardware changes or physical 
    alteration of the plant which could introduce new equipment failure 
    modes, and there will be no impact on the design or function of the 
    SW System or its components. The primary and secondary containment 
    post-LOCA responses remain within previously assessed limits of 
    temperature and pressure. Furthermore, adequate cooling flow is 
    assured during plant operation and also during shutdown conditions 
    such that essential systems and components remain within their 
    applicable design limits. It is, therefore, concluded that no 
    requirements are eliminated or new requirements imposed which could 
    affect equipment or plant operation such that new credible accidents 
    are introduced. Accordingly, operation of NMP2, in accordance with 
    the proposed amendment, will not create the possibility of a new or 
    different kind of accident from any accident previously evaluated.
        The operation of Nine Mile Point Unit 2, in accordance with the 
    proposed amendment, will not involve a significant reduction in a 
    margin of safety.
        The changes provide assurance that the SW System will remain 
    capable of providing sufficient cooling flow for the essential 
    cooling loads during plant operation and also during plant shutdown 
    such that essential systems and components remain within their 
    applicable design limits. The changes will have no impact on the 
    design or function of the SW System and its components, thus 
    assuring that the characteristics and functional performance are 
    maintained consistent with the conditions and assumptions of the 
    current design basis accident and transient analyses. Maintaining 
    four pumps operating within acceptable system parameters, with the 
    divisional cross connect valves open, during Operational Conditions 
    1, 2, and 3 provides assurance that post-LOCA radioactive releases 
    are maintained within 10 CFR 100 limits. The small (1 degree F) 
    change in the SW supply header discharge water (UHS) temperature and 
    the limiting temperature for the Intake Deicing Heater System Action 
    and operability requirements maintains the current design basis for 
    the UHS and SW Systems such that there will be no impact on the LOCA 
    analyses assumptions or conclusions.
        These changes will not result in a reduction in margin to the 
    System analytical limits. Furthermore, maintaining the intake bar 
    surface temperature at least 1 degree F above freezing provides an 
    adequate margin to prevent the adherence of ice, and provides 
    assurance that sufficient flow area is always heated such that the 
    SW System will remain capable of providing adequate cooling flow in 
    the event of a LOCA. Similarly, maintaining the required SW System 
    flow and temperature during Operational Conditions 4, 5, and * will 
    assure that the associated equipment is operable such that 
    radioactive releases are maintained within 10 CFR 100 limits. It is, 
    therefore, concluded that the changes do not eliminate any 
    requirements, impose any new requirements, or alter any physical 
    parameters which significantly reduce the margin to an acceptance 
    limit or adversely affect the margins associated with the fission 
    product barriers as established by the design basis accident and 
    transient analyses. Accordingly, operation of NMP2, in accordance 
    with the proposed amendment, will not involve a significant 
    reduction in a margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
    
        Local Public Document Room location: Reference and Documents 
    Department, Penfield Library, State University of New York, Oswego, New 
    York 13126.
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
    Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
        NRC Project Director: S. Singh Bajwa.
    
    Northeast Nuclear Energy Company (NNECO) et al., Docket No. 50-336, 
    Millstone Nuclear Power Station, Unit No. 2, New London County, 
    Connecticut
    
        Date of amendment request: September 28, 1998.
        Description of amendment request: The proposed amendment would 
    change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered 
    Safety Features Actuation System''; 3.4.6.2, ``Reactor Coolant System--
    Reactor Coolant System Leakage''; 3.4.8, ``Reactor Coolant System--
    Specific Activity''; 3.6.2.1, ``Containment Systems--Depressurization 
    and Cooling Systems Containment Spray and Cooling Systems''; 3.6.5.1, 
    ``Containment Systems--Secondary Containment Enclosure Building 
    Filtration System''; 3.7.6.1, ``Plant Systems--Control Room Emergency 
    Ventilation System''; and 3.9.15, ``Refueling Operations--Storage Pool 
    Area Ventilation System--Fuel Storage.'' Information would also be 
    added to the Bases of the associated Technical Specifications to 
    address the proposed changes.
        The proposed amendment would also revise the Operating License DPR-
    65 by incorporating a change to the Millstone Unit No. 2 Final Safety 
    Analysis Report (FSAR). The change to the FSAR is associated with the 
    revised main steamline break analyses, new determination of the 
    radiological consequences of a main steamline break, and a revised 
    determination of the radiological consequences of the design basis 
    loss-of-coolant accidents (LOCAs).
        The proposed changes to the main steamline break analysis, as 
    described in the FSAR, are based on the revised Siemens Power 
    Corporation steamline break methodology. The report describing the 
    revised methodology was submitted by Siemens Power Corporation to the 
    NRC for approval in a letter dated June 30, 1998. The revised 
    methodology was used to perform the Millstone Unit No. 2 plant-specific 
    analysis for post-scram main steamline break. This plant-specific 
    analysis was submitted by NNECO in a letter dated August 12, 1998, 
    which proposed to change the list of documents in the Technical 
    Specifications that describe the analytical methods used to determine 
    the core operating limits. The proposed changes contained in this 
    letter assume approval of the previously submitted revised Siemens 
    Power Corporation steamline break methodology, and the changes to the 
    list of documents in the Millstone Unit No. 2 Technical Specifications 
    that describe the analytical methods used to determine the core 
    operating limits.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        In accordance with 10CFR50.92, NNECO has reviewed the proposed 
    changes and has concluded that they do not involve a significant 
    hazards consideration (SHC). The basis for this conclusion is that 
    the three criteria of 10CFR50.92(c) are not compromised. The 
    proposed changes do not involve an SHC because the changes would 
    not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
    
    Analyses Changes
    
        The main steam line break analyses and the determinations of the 
    radiological consequences of the main steam line break and loss of 
    coolant accident have been revised. A brief summary of the 
    significant changes to the main steam line break analyses and the 
    radiological consequences of the main steam line break and loss of 
    coolant accident is presented below.
        1. The limited fuel failure following a main steam line break 
    outside containment results in an increase in the calculated 
    radiological consequences both off-site and in the control room. To 
    limit the consequences of a main steam line break outside 
    containment, the
    
    [[Page 66598]]
    
    Technical Specification allowed steam generator tube leakage will be 
    reduced to 0.035 gpm [gallons per minute] per steam generator.
        2. Credit will now be taken for iodine removal from the 
    containment atmosphere by the Containment Spray System (CSS). The 
    use of the CSS for iodine removal has not been previously approved 
    by the NRC.
        3. The proposed increase to the allowable control room in-
    leakage will provide additional operational flexibility to address 
    expected minor system degradation over time. The increase in the 
    allowable control room in-leakage will result in an increase in the 
    calculated dose to the Control Room Operators.
        4. The addition of the dose consequences from containment sump 
    backleakage to the Refueling Water Storage Tank (RWST) has been 
    included in the off-site and control room loss of coolant accident 
    (LOCA) analyses increases the consequences of previously evaluated 
    accidents.
        The containment sump backleakage into the RWST results in sump 
    water entering the RWST when the RWST is at its minimum level. The 
    RWST will become a radioactive source and contribute a shine dose to 
    the surrounding areas. The increase in dose rates onsite will not 
    prevent operators from remaining in the control room or from 
    accessing equipment needed to mitigate the accident.
        All piping and valves associated with RWST backleakage are 
    located in a harsh radiation area. Backflow from the sump might 
    increase dose rates in the area where these components are located. 
    Additional dose contributions, where they occur, do not adversely 
    impact the environmental qualification of the vital equipment 
    located there. All vital equipment would continue to perform its 
    safety function.
        5. Credit will be taken in the main steam line break analyses 
    for the recently installed cavitating venturis in the Auxiliary 
    Feedwater System. However, this will not change the amount of fuel 
    failure. Therefore, credit for this equipment will not impact the 
    radiological consequences of a main steam line break.
        6. Credit will be taken for the Reactor Coolant System (RCS) low 
    flow reactor trip for the pre-scram inside containment main steam 
    line break analysis. This equipment will be qualified for the 
    expected containment environment following a main steam line break 
    inside containment and will be added to the Environmental 
    Qualification Master List.
        7. Millstone Unit No. 1 design basis accidents, loss of coolant 
    and main steam line break, will no longer be evaluated for impact on 
    Millstone Unit No. 2 control room habitability. This credits the 
    decision to decommission Millstone Unit No. 1. [Footnote--B.D. 
    Kenyon letter to the NRC, ``Millstone Nuclear Power Station, Unit 
    No. 1 Certification of Permanent Cessation of Power Operations and 
    that Fuel Has Been Permanently Removed from the Reactor,'' dated 
    July 21, 1998.]
        The revised main steam line break analyses and the revised 
    determinations of the radiological consequences of the main steam 
    line break and design basis LOCA analyses take credit for equipment 
    not previously assumed in the analyses, and for plant or equipment 
    operating restrictions not currently contained in the Technical 
    Specifications. The changes to the analyses will not adversely 
    affect the probability of an accident previously evaluated, but the 
    revised analyses results do indicate that the consequences of an 
    accident previously evaluated will increase. Specifically, the 
    following changes cause an increase in the consequences of an 
    accident previously evaluated.
        1. The increase in allowable control room in-leakage from 100 
    SCFM [standard cubic feet per minute] to 130 SCFM when the Control 
    Room Emergency Ventilation System is operating in the recirculation/
    filtration mode.
        The dose to the Control Room Operators from a Millstone Unit No. 
    2 LOCA increased from 9.25 to 25.8 rem to the thyroid and from 0.205 
    to 2.29 rem to the skin. The dose to the whole body decreased. (Both 
    low wind speed and high wind speed release conditions were analyzed. 
    The low wind speed condition bounds the high wind speed condition.) 
    The dose to the Control Room Operators from a Millstone Unit No. 3 
    LOCA increased from 2.67 to 14 rem to the skin and from 0.209 to 
    1.484 rem to the whole body. The dose to the thyroid decreased. The 
    doses to the Control Room Operators from either a Millstone Unit No. 
    2 or Unit No. 3 LOCA remain below the GDC [General Design Criterion] 
    19 criteria of 30 rem thyroid, 5 rem whole body and 30 rem to the 
    skin.
        The new calculated doses to the Millstone Unit No. 2 Control 
    Room Operators from a main steam line break outside containment are 
    29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. The doses to 
    the Millstone Unit No. 2 Control Room Operators are below the GDC 19 
    criteria of 30 rem thyroid, 5 rem whole body, and 30 rem to the 
    skin. (Note: The dose to the Control Room Operators from a main 
    steam line break was not previously evaluated because fuel failure 
    was not predicted to occur.)
        2. The limited fuel failure that is predicted in the revised 
    main steam line break analyses.
        Previously, the radiological consequences of a main steam line 
    break were not determined and were not presented in the FSAR because 
    fuel failure was not predicted to occur. Because of the predicted 
    limited fuel failure for the main steam line break outside of 
    containment, the radiological consequences were analyzed. The 
    results to the Exclusion Area Boundary (EAB) are 4.8 rem thyroid and 
    0.06 rem whole body. The results to the Low Population Zone (LPZ) 
    are 2.3 rem thyroid and 0.02 rem whole body. To meet the dose 
    acceptance criteria to the Millstone Unit No. 2 Control Room 
    Operators, the maximum allowable Technical Specification primary to 
    secondary leak rate is being reduced to 0.035 gpm per steam 
    generator. The results to the Millstone Unit No. 2 Control Room 
    Operators are 29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. 
    The main steam line break outside containment is the limiting 
    accident for the Millstone Unit No. 2 Control Room Operators. 
    However, the dose consequences of a main steam line break are less 
    than the 10CFR100 limits off-site of 300 rem thyroid and 25 rem 
    whole body, and the doses to the Millstone Unit No. 2 Control Room 
    Operators are below the GDC 19 criteria of 30 rem thyroid, 5 rem 
    whole body, and 30 rem to the skin.
        3. Taking credit for the low RCS flow reactor trip for the pre-
    scram inside containment main steam line break analysis.
        Previous analyses did not credit the low RCS flow reactor trip 
    in a harsh environment. This credits the low flow trip in a manner 
    not previously reviewed by the NRC for Millstone Unit No. 2. Without 
    credit for this reactor trip, the predicted fuel failure for steam 
    line breaks inside containment would be higher.
        4. Taking credit for the removal of radioactive iodine from the 
    containment atmosphere by containment spray.
        Previous analyses did not rely on the spray function to reduce 
    iodine concentration in the post-accident atmosphere inside 
    containment. This adds a mitigation function to the CSS that has not 
    been previously reviewed by the NRC for Millstone Unit No. 2. 
    Without credit for the removal of iodine, the predicted dose 
    consequences following a LOCA would be higher.
        5. The addition of sump backleakage to the RWST during a LOCA.
        The resultant dose contribution to the LPZ from RWST backleakage 
    is 1.487 rem thyroid and 0.11 rem whole body. The total dose to the 
    LPZ from a design basis LOCA is 21.86 rem thyroid and 0.941 rem 
    whole body. The dose is well below the 10CFR100 limits of 300 rem 
    thyroid and 25 rem whole body. The dose to the EAB was not affected 
    because leakage into the RWST does not start until 25.45 hours post-
    LOCA and the EAB is a 2-hour dose.
        The resultant dose contribution to the Millstone Unit No. 2 
    Control Room Operators from RWST backleakage is 3.75 rem thyroid, 
    0.017 rem whole body and 0.296 to the skin. The total dose to the 
    Millstone Unit No. 2 Control Room Operators from the LOCA is 25.8 
    rem thyroid, 0.718 rem whole body and 2.29 rem to the skin. These 
    doses are below the GDC 19 limits of 30 rem thyroid and skin, and 5 
    rem whole body.
        The analyses results meet the guidance contained in SRP 
    [Standard Review Plan] 15.1.5, SRP 15.6.5, and the limits of 
    10CFR100 and GDC 19. Therefore, there will be no significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
    
    Technical Specification Changes
    
        Technical Specification Non-Technical Changes
        The minor editorial and non-technical changes to correct 
    spelling (Technical Specification 3.3.2.1), modify the title of a 
    table column (Technical Specification 3.4.8), clarify the type of 
    measurement performed (Technical Specification 3.4.8), and establish 
    consistent terminology (Technical Specification 3.7.6.1) will not 
    result in any technical changes to the Millstone Unit No. 2 
    Technical Specifications. The proposed changes will have no adverse 
    effect on plant
    
    [[Page 66599]]
    
    operation. Therefore, there will be no significant increase in the 
    probability or consequences of an accident previously evaluated.
    
    Technical Specification 3.4.6.2
    
        The reduction in the maximum allowable value of primary to 
    secondary leakage per steam generator is consistent with the new 
    radiological assessment of the potential control room operator 
    exposure following a main steam line break outside of containment. 
    The wording change to SR [Surveillance Requirement] 4.4.6.2.1 will 
    clarify that the water inventory balance is used to verify 
    compliance with the identified and unidentified leakage limits. 
    Pressure boundary leakage would first show up as unidentified 
    leakage during performance of SR 4.4.6.2.1. Further investigation, 
    (plant walkdown) would be necessary to classify the unidentified 
    leakage as pressure boundary leakage. This is consistent with 
    established plant practices to detect pressure boundary leakage.
        The addition of the new SR 4.4.6.2.2 will address the primary to 
    secondary leakage limit. The new SR will include an exception to 
    Technical Specification 4.0.4 that will allow the determination of 
    primary to secondary leakage to be deferred until after Mode 4 is 
    entered. Even though verification of compliance with the primary to 
    secondary limit will not be done prior to entering Mode 4, the limit 
    is still expected to be met.
        The proposed changes will have no adverse effect on plant 
    operation. Therefore, there will be no significant increase in the 
    probability or consequences of an accident previously evaluated.
    
    Technical Specification 3.4.8
    
        The addition of the words ``of gross specific activity'' to the 
    Limiting Condition for Operation (LCO), Action Statements, and SR 
    will clarify what the E-Bar limit applies to. This is consistent 
    with the Technical Specification Definition (1.20) for E-Bar.
        The addition of a footnote (*) to state the power history 
    requirements for the determination of E-Bar will ensure that the 
    necessary plant conditions are established prior to performing the 
    analysis. This will not affect the E-Bar LCO limit or the 
    requirement to perform the analysis. The proposed change is 
    consistent with NUREG--0212 and NUREG--1432.
        The footnote will also specify that the provisions of 
    Specification 4.0.4 are not applicable. This will allow entry into 
    Mode 1, without determining the value of E-Bar, assuming that the 
    power history requirements will not be met until after Mode 1 is 
    entered. This will normally only apply following an extended 
    shutdown.
        The Isotopic Analysis for Iodine (including I-131, I-133, and I-
    135) sample requirement will be expanded to include the LCO 
    requirement for 100/E-Bar. This is consistent with the requirements 
    of Action Statement d. This change will expand the sampling 
    requirement for iodine. Minor wording changes will also be made to 
    be consistent with the proposed changes to the LCO wording.
        The proposed changes will have no adverse effect on plant 
    operation. Therefore, there will be no significant increase in the 
    probability or consequences of an accident previously evaluated.
    
    Technical Specification 3.6.2.1
    
        The revised radiological assessment calculation for the design 
    basis accident credits iodine removal from the containment 
    atmosphere by the CSS. This will require a reduction in the allowed 
    outage time (AOT) of one containment spray train from seven days to 
    seventy two hours. This AOT is consistent with NUREG-0212 and NUREG-
    1432. This will help ensure that plant equipment assumed in the 
    safety analyses will be available. This is a more restrictive change 
    which will have no adverse effect on plant operation. Therefore, 
    there will be no significant increase in the probability or 
    consequences of an accident previously evaluated.
    
    Technical Specification 3.6.5.1
    
        The value for the pressure drop across the combined HEPA [high-
    efficiency particulate air] filters and charcoal adsorber banks 
    specified in SR 4.6.5.1.d.1 will be changed from a generic value 
    [less than or equal to] 6 inches water gauge) to a plant specific 
    value [less than or equal to] 2.6 inches water gauge). This is a 
    more restrictive change which will have no adverse effect on plant 
    operation. Therefore, there will be no significant increase in the 
    probability or consequences of an accident previously evaluated.
    
    Technical Specification 3.7.6.1
    
        The value for the pressure drop across the combined HEPA filters 
    and charcoal adsorber banks specified in SR 4.7.6.1.e.1 will be 
    changed from a generic value [less than or equal to] 6 inches water 
    gauge) to a plant specific value [less than or equal to] 3.4 inches 
    water gauge). This is a more restrictive change which will have no 
    adverse effect on plant operation.
        SR 4.7.6.1.e.2 will be expanded to clarify that the test of the 
    capability of the Control Room Emergency Ventilation Trains to 
    switch to the recirculation mode is performed with the trains 
    initially operating in the normal mode and the smoke purge mode of 
    operation. This will not affect the requirement that the trains be 
    capable of switching to the recirculation mode.
        The value of allowable control room air in-leakage specified in 
    SR 4.7.6.1.e.3 will be increased from 100 SCFM to 130 SCFM. This is 
    consistent with the recently revised control room radiological 
    analysis for the design basis accidents.
        The proposed increase will provide additional operational 
    flexibility to address expected minor system degradation over time. 
    This increase is supported by the new analysis.
        The proposed changes will have no adverse effect on plant 
    operation. Therefore, there will be no significant increase in the 
    probability or consequences of an accident previously evaluated.
    
    Technical Specification 3.9.15
    
        The value for the pressure drop across the combined HEPA filters 
    and charcoal adsorber banks specified in SR 4.9.15.d.1 will be 
    changed from a generic value [less than or equal to] 6 inches water 
    gauge) to a plant specific value [less than or equal to] 2.6 inches 
    water gauge). This is a more restrictive change which will have no 
    adverse effect on plant operation. Therefore, there will be no 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        The proposed changes have no adverse effect on how any of the 
    associated systems or components function to prevent or mitigate the 
    consequences of design basis accidents. Also, the proposed changes 
    have no adverse effect on any design basis accident previously 
    evaluated since the changes are consistent with the revised 
    analyses, and the appropriate acceptance criteria are met for the 
    revised analyses. Therefore, the license amendment request does not 
    impact the probability of an accident previously evaluated nor does 
    it involve a significant increase in the consequences of an accident 
    previously evaluated.
        2. Create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        The proposed changes will not alter the plant configuration (no 
    new or different type of equipment will be installed) or require any 
    new or unusual operator actions. They do not alter the way any 
    structure, system, or component functions and do not alter the 
    manner in which the plant is operated. The proposed changes do not 
    introduce any new failure modes.
        Also, the response of the plant and the operators following 
    these accidents is unaffected by the change. Therefore, the proposed 
    changes will not create the possibility of a new or different kind 
    of accident from any accident previously evaluated.
        3. Involve a significant reduction in a margin of safety.
    
    Analyses Changes
    
        The acceptance criteria for a main steam line break in the SRP 
    15.1.5 does not exclude the prediction of fuel failure. Instead, the 
    SRP requires that ``Any fuel damage calculated to occur must be of 
    sufficiently limited extent that the core will remain in place and 
    intact with no loss of core cooling.'' The limited fuel failure that 
    is now predicted in the revised main steam line break analyses meets 
    this acceptance criterion. In addition, the RCS low flow reactor 
    trip that is now being credited to function in a harsh environment 
    to limit fuel failure is already required to be operable by 
    Technical Specifications.
        The revised dose consequences for the design basis accidents 
    assumes a control room in-leakage of 130 SCFM. In addition, iodine 
    removal by the CSS, which is already required to be operable by 
    Technical Specifications, is assumed. The acceptance criteria for 
    the dose consequences of the design basis accidents to the EAB, LPZ 
    and the control room personnel is met in the revised analyses. 
    Therefore, the revisions to the dose consequence analyses for the 
    design basis accidents do not involve a significant reduction in the 
    margin of safety.
    
    [[Page 66600]]
    
    Technical Specification Changes
    
        The proposed changes will correct spelling and terminology 
    errors, reduce the maximum allowable primary to secondary leakage, 
    add a new surveillance requirement, modify surveillance requirements 
    for RCS specific activity, reduce the allowed outage time for a 
    containment spray train, reduce the allowed pressure drop across the 
    control room and enclosure building HEPA [high-efficiency 
    particulate air] filters, and increase the control room maximum 
    allowed in-leakage. These changes will have no adverse effect on 
    equipment important to safety. The equipment will continue to 
    function as assumed in the design basis accident analysis. 
    Therefore, there will be no significant reduction of the margin of 
    safety as defined in the Bases for the Technical Specifications 
    affected by these proposed changes.
        The only adverse impact of the proposed changes is that the dose 
    consequences following an accident may increase. However, the 
    revised analyses show that the acceptance criteria for the accident 
    analyses are met. Therefore, based on the responses above, the 
    proposed changes are deemed safe.
        The NRC has provided guidance concerning the application of 
    standards in 10CFR50.92 by providing certain examples (March 6, 
    1986, 51 FR 7751) of amendments that are considered not likely to 
    involve an SHC. The minor editorial and non-technical changes 
    proposed herein to correct reference, spelling, and terminology 
    errors are enveloped by example (i), a purely administrative change 
    to Technical Specifications. The changes proposed herein to add a 
    new surveillance requirement to verify primary to secondary leakage 
    and to reduce the allowable pressure drop across various ventilation 
    filters are enveloped by example (ii), a change that constitutes an 
    additional limitation, restriction, or control not presently 
    included in the Technical Specifications. All of the other changes 
    proposed herein are not enveloped by any specific example.
        As described above, this License Amendment Request does not 
    impact the probability of an accident previously evaluated, does not 
    involve a significant increase in the consequences of an accident 
    previously evaluated, does not create the possibility of a new or 
    different kind of accident from any accident previously evaluated, 
    and does not result in a significant reduction in a margin of 
    safety. Therefore, NNECO has concluded that the proposed changes do 
    not involve an SHC.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Learning Resources Center, 
    Three Rivers Community-Technical College, 574 New London Turnpike, 
    Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
    49 Rope Ferry Road, Waterford, Connecticut.
        Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
    Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
    Connecticut.
        NRC Project Director: William M. Dean.
    
    Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
    Nuclear Power Station, Unit No. 2, New London County, Connecticut
    
        Date of amendment request: October 22, 1998.
        Description of amendment request: The licensee is proposing to 
    change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered 
    Safety Feature Actuation System Instrumentation''; 3.4.9.3, ``Reactor 
    Coolant System [RCS]--Overpressure Protection Systems''; and 3.5.3, 
    ``Emergency Core Cooling Systems--ECCS Subsystems--Tavg < 300="" [degrees]="" f.''="" the="" proposed="" changes="" will="" allow="" millstone="" unit="" no.="" 2="" to="" prevent="" an="" automatic="" start="" of="" any="" high-pressure="" safety="" injection="" (hpsi)="" pump="" when="" the="" shutdown="" cooling="" system="" (sdcs)="" is="" in="" operation="" (mode="" 4="" and="" below).="" an="" inadvertent="" start="" of="" an="" hpsi="" pump="" could="" result="" in="" overpressurization="" of="" the="" sdcs.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" in="" accordance="" with="" 10cfr50.92,="" northeast="" nuclear="" energy="" company="" (nneco)="" has="" reviewed="" the="" proposed="" changes="" and="" has="" concluded="" that="" they="" do="" not="" involve="" a="" significant="" hazards="" consideration="" (shc).="" the="" basis="" for="" this="" conclusion="" is="" that="" the="" three="" criteria="" of="" 10cfr50.92(c)="" are="" not="" compromised.="" the="" proposed="" changes="" do="" not="" involve="" an="" shc="" because="" the="" changes="" would="" not:="" 1.="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" changes="" to="" technical="" specifications="" 3.3.2.1="" and="" 3.5.3="" will="" no="" longer="" require="" the="" hpsi="" pump,="" required="" to="" be="" operable="" in="" mode="" 4,="" to="" start="" automatically="" on="" a="" safety="" injection="" actuation="" signal="" (sias).="" (the="" automatic="" siass="" on="" low="" pressurizer="" pressure="" and="" high="" containment="" pressure="" are="" not="" required="" to="" be="" operable="" in="" mode="" 4.="" however,="" the="" manual="" safety="" injection="" pushbuttons="" are="" required="" in="" mode="" 4).="" this="" will="" allow="" the="" operable="" hpsi="" pump="" control="" switch="" to="" be="" placed="" in="" the="" pull-to-lock="" position="" without="" affecting="" the="" operability="" of="" that="" pump.="" all="" hpsi="" pumps="" will="" be="" prevented="" from="" automatically="" starting="" when="" the="" plant="" is="" in="" mode="" 4,="" and="" the="" shutdown="" cooling="" system="" (sdcs)="" is="" aligned="" to="" the="" rcs="" to="" prevent="" an="" inadvertent="" start="" of="" a[n]="" hpsi="" pump="" which="" could="" overpressurize="" the="" sdcs.="" these="" changes="" will="" not="" reduce="" the="" requirement="" for="" at="" least="" one="" hpsi="" pump="" to="" be="" operable="" in="" mode="" 4.="" the="" changes="" will="" require="" an="" additional="" operator="" action="" to="" remove="" the="" operable="" hpsi="" pump="" breaker="" control="" switch="" from="" the="" pull-to-lock="" position,="" in="" addition="" to="" initiating="" safety="" injection="" by="" use="" of="" the="" manual="" pushbuttons,="" if="" safety="" injection="" system="" actuation="" is="" needed="" in="" mode="" 4.="" the="" requirement="" to="" manually="" initiate="" a[n]="" hpsi="" pump,="" in="" addition="" to="" manually="" initiating="" a[n]="" sias,="" does="" not="" involve="" complicated="" equipment="" manipulations="" nor="" require="" extensive="" time="" for="" performing="" the="" required="" operator="" actions.="" the="" hpsi="" pump="" control="" switches="" are="" located="" in="" the="" control="" room="" on="" the="" same="" panels="" as="" the="" manual="" sias="" pushbuttons.="" the="" additional="" step="" required="" to="" start="" a[n]="" hpsi="" pump="" will="" not="" add="" any="" appreciable="" time="" for="" initiating="" hpsi="" flow="" while="" in="" mode="" 4.="" in="" addition,="" considering="" the="" lower="" probability="" of="" a="" significant="" loss="" of="" coolant="" accident="" in="" mode="" 4,="" and="" the="" slower="" plant="" response="" to="" a="" loss="" of="" coolant="" accident="" in="" mode="" 4,="" the="" time="" required="" for="" the="" additional="" operator="" action="" will="" have="" no="" significant="" effect="" on="" the="" consequences="" of="" the="" accident.="" therefore,="" there="" will="" be="" no="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" change="" to="" technical="" specification="" 3.4.9.3,="" surveillance="" requirement="" (sr)="" 4.4.9.3.3,="" will="" allow="" the="" use="" of="" the="" new="" pull-to-lock="" feature="" of="" the="" hpsi="" pump="" control="" switches="" to="" satisfy="" low="" temperature="" overpressure="" protection="" mass="" input="" requirements.="" this="" will="" not="" affect="" either="" the="" ltop="" [low-temperature="" overpressure="" protection]="" hpsi="" pump="" mass="" input="" restrictions="" or="" the="" level="" of="" control="" to="" ensure="" the="" hpsi="" pumps="" are="" not="" capable="" of="" injecting="" into="" the="" rcs.="" the="" proposed="" changes="" will="" have="" no="" adverse="" effect="" on="" plant="" operation.="" therefore,="" there="" will="" be="" no="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" minor="" editorial="" and="" non-technical="" changes="" to="" add="" amendment="" numbers="" to="" page="" 3/4="" 3-12="" and="" to="" revise="" the="" wording="" of="" srs="" 4.4.9.3.2="" and="" 4.4.9.3.3="" will="" not="" result="" in="" any="" technical="" changes="" to="" the="" millstone="" unit="" no.="" 2="" technical="" specifications.="" the="" proposed="" changes="" will="" have="" no="" adverse="" effect="" on="" plant="" operation.="" therefore,="" there="" will="" be="" no="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" changes="" to="" the="" bases="" reflect="" the="" proposed="" changes="" to="" the="" applicable="" technical="" specifications.="" the="" proposed="" changes="" will="" have="" no="" adverse="" effect="" on="" plant="" operation.="" therefore,="" there="" will="" be="" no="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" 2.="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" the="" proposed="" changes="" will="" allow="" the="" use="" of="" the="" hpsi="" pump="" breaker="" control="" switch="" [[page="" 66601]]="" pull-to-lock="" feature.="" operation="" of="" the="" hpsi="" pump="" in="" mode="" 4="" will="" change="" since="" the="" operator="" will="" have="" to="" start="" the="" hpsi="" pump,="" in="" addition="" to="" manually="" initiating="" safety="" injection.="" however,="" hpsi="" pump="" operation="" is="" not="" an="" accident="" initiator.="" therefore,="" the="" proposed="" changes="" will="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" 3.="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" proposed="" technical="" specification="" changes="" will="" no="" longer="" require="" the="" hpsi="" pump,="" required="" to="" be="" operable="" in="" mode="" 4,="" to="" start="" automatically="" on="" a[n]="" sias,="" will="" allow="" the="" use="" of="" the="" new="" pull-to-="" lock="" feature="" of="" the="" hpsi="" pump="" control="" switches="" to="" satisfy="" low="" temperature="" overpressure="" protection="" mass="" input="" requirements,="" and="" will="" make="" minor="" editorial="" and="" non-technical="" changes.="" these="" changes="" will="" have="" no="" adverse="" effect="" on="" equipment="" important="" to="" safety.="" the="" equipment="" will="" continue="" to="" function="" as="" assumed="" in="" the="" design="" basis="" accident="" analysis.="" therefore,="" there="" will="" be="" no="" significant="" reduction="" in="" the="" margin="" of="" safety="" as="" defined="" in="" the="" bases="" for="" the="" technical="" specifications="" affected="" by="" these="" proposed="" changes.="" the="" only="" adverse="" impact="" of="" the="" proposed="" changes="" is="" that="" an="" additional="" operator="" action="" will="" be="" necessary="" to="" initiate="" hpsi="" flow="" in="" mode="" 4,="" if="" needed.="" however,="" considering="" the="" lower="" probability="" of="" a="" significant="" loss="" of="" coolant="" accident="" in="" mode="" 4,="" and="" the="" slower="" plant="" response="" to="" a="" loss="" of="" coolant="" accident="" in="" mode="" 4,="" the="" time="" required="" for="" the="" additional="" operator="" action="" will="" have="" no="" significant="" effect="" on="" the="" consequences="" of="" the="" accident.="" therefore,="" based="" on="" the="" responses="" above,="" the="" proposed="" changes="" are="" deemed="" safe.="" the="" nrc="" has="" provided="" guidance="" concerning="" the="" application="" of="" standards="" in="" 10cfr50.92="" by="" providing="" certain="" examples="" (march="" 6,="" 1986,="" 51="" fr="" 7751)="" of="" amendments="" that="" are="" considered="" not="" likely="" to="" involve="" an="" shc.="" the="" minor="" editorial="" and="" non-technical="" changes="" proposed="" herein="" to="" add="" page="" amendment="" numbers="" and="" clarify="" wording="" are="" enveloped="" by="" example="" (i),="" a="" purely="" administrative="" change="" to="" technical="" specifications.="" all="" of="" the="" other="" changes="" proposed="" herein="" are="" not="" enveloped="" by="" any="" specific="" example.="" as="" described="" above,="" this="" license="" amendment="" request="" does="" not="" impact="" the="" probability="" of="" an="" accident="" previously="" evaluated,="" does="" not="" involve="" a="" significant="" increase="" in="" the="" consequences="" of="" an="" accident="" previously="" evaluated,="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated,="" and="" does="" not="" result="" in="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" therefore,="" nneco="" has="" concluded="" that="" the="" proposed="" changes="" do="" not="" involve="" an="" shc.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 10="" cfr="" 50.92(c)="" are="" satisfied.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" local="" public="" document="" room="" location:="" learning="" resources="" center,="" three="" rivers="" community-technical="" college,="" 574="" new="" london="" turnpike,="" norwich,="" connecticut,="" and="" the="" waterford="" library,="" attn:="" vince="" juliano,="" 49="" rope="" ferry="" road,="" waterford,="" connecticut.="" attorney="" for="" licensee:="" lillian="" m.="" cuoco,="" esq.,="" senior="" nuclear="" counsel,="" northeast="" utilities="" service="" company,="" p.o.="" box="" 270,="" hartford,="" connecticut.="" nrc="" project="" director:="" william="" m.="" dean.="" peco="" energy="" company,="" docket="" nos.="" 50-352="" and="" 50-353,="" limerick="" generating="" station,="" units="" 1="" and="" 2,="" montgomery="" county,="" pennsylvania="" date="" of="" amendment="" request:="" october="" 30,="" 1998.="" description="" of="" amendment="" request:="" limerick="" generating="" station="" (lgs),="" units="" 1="" and="" 2,="" technical="" specifications="" (ts)="" surveillance="" requirements="" 4.8.4.3.b.1,="" 4.8.4.3.b.2,="" and="" 4.8.4.3.b.3="" list="" the="" overvoltage="" (ov),="" undervoltage="" (uv),="" and="" underfrequency="" (uf)="" values="" for="" the="" protective="" instrumentation="" for="" the="" rps="" electric="" power="" monitoring="" channels.="" the="" proposed="" changes="" correct="" a="" discrepancy="" between="" the="" general="" electric="" nuclear="" engineering="" (gene)="" design="" specification="" for="" power="" supply="" monitoring="" relays="" and="" the="" existing="" ts="" allowable="" values="" (avs).="" the="" changes="" will="" revise="" the="" ov,="" us,="" and="" uf="" values="" from="" 132vac,="" 109vac,="" and="" 57hz="" to="" 127.6vac,="" 110.7vac,="" and="" 57.05hz="" respectively.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" the="" proposed="" technical="" specifications="" (ts)="" changes="" do="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" tech="" spec="" changes="" to="" section="" 4.8.4.3.b="" for="" the="" overvoltage="" (ov),="" undervoltage="" (uv),="" and="" underfrequency="" (uf)="" relays="" are="" more="" conservative="" than="" the="" existing="" ts="" values.="" this="" change="" provides="" more="" protection="" for="" the="" associated="" rps="" components,="" thus="" decreasing="" the="" probability="" of="" a="" failure="" in="" rps.="" the="" associated="" non-="" conformance="" report="" and="" calculation="" provide="" assurance="" that="" the="" ov/uv/="" uf="" settings="" are="" acceptable="" since="" the="" calculated="" values="" assure="" that="" the="" rps="" components="" will="" operate="" within="" their="" ratings.="" there="" are="" no="" physical="" changes="" to="" the="" associated="" protective="" relays="" by="" the="" ts="" change;="" thus,="" original="" design="" basis="" redundancy="" and="" separation="" is="" maintained.="" there="" is="" no="" change="" in="" the="" interface="" of="" the="" rps="" and="" its="" power="" supplies.="" the="" safety="" function="" of="" the="" rps="" is="" to="" initiate="" a="" reactor="" scram="" in="" order="" to="" protect="" the="" primary="" fission="" products="" barrier,="" the="" reactor="" fuel.="" the="" proposed="" ts="" change="" to="" impose="" more="" conservative="" allowable="" values="" for="" the="" ov,="" uv,="" and="" uf="" relays="" will="" provide="" additional="" assurance="" that="" the="" rps="" will="" operate="" within="" equipment="" voltage="" and="" frequency="" ratings,="" and="" will="" not="" be="" damaged="" by="" power="" system="" anomalies.="" this="" change="" will="" not="" affect="" the="" scram="" function="" of="" rps;="" thus,="" the="" consequences="" of="" any="" design="" basis="" events="" will="" not="" be="" affected.="" therefore,="" the="" proposed="" ts="" changes="" do="" not="" involve="" an="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" 2.="" the="" proposed="" ts="" changes="" do="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" the="" proposed="" ts="" allowable="" values="" changes="" will="" not="" result="" in="" any="" physical="" changes="" to="" the="" rps="" electric="" power="" monitoring="" system.="" existing="" setpoints="" will="" not="" be="" changed,="" only="" the="" ts="" allowable="" values="" are="" being="" modified="" to="" be="" more="" conservative.="" the="" system="" redundancy="" and="" independence="" are="" not="" changed,="" and="" no="" new="" failure="" modes="" are="" introduced.="" therefore,="" the="" proposed="" ts="" changes="" do="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" previously="" evaluated.="" 3.="" the="" proposed="" ts="" changes="" do="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" currently,="" there="" are="" no="" ts="" bases="" for="" the="" existing="" rps="" electric="" power="" monitoring="" system="" ov,="" uv,="" and="" uf="" allowable="" values.="" specific="" analytical="" limits="" for="" system="" voltage="" and="" frequency="" are="" not="" defined="" in="" the="" safety="" analysis="" report,="" nor="" discussed="" in="" any="" design="" basis="" allowed="" outage="" time="" or="" accident="" evaluation.="" investigation="" into="" the="" licensing="" basis="" has="" identified="" nominal="" values="" of="" +/-10%="" of="" 120="" vac="" and="" -5%="" of="" 60="" hz="" for="" the="" allowable="" values.="" these="" values="" are="" included="" in="" nureg="" 0123,="" from="" which="" lgs's="" tss="" were="" developed.="" nureg="" 0123="" also="" provides="" no="" bases="" for="" these="" values.="" the="" proposed="" changes="" in="" the="" ts="" allowable="" values="" is="" based="" on="" a="" revision="" to="" the="" calculation="" for="" rps="" breaker="" panel--rps="" ups="" [uninterruptible="" power="" supply]="" system="" bus="" relay="" settings.="" this="" revision="" determines="" the="" new="" allowable="" values="" based="" on="" the="" design="" ratings="" of="" rps="" components,="" and="" factors="" in="" instrument="" inaccuracies="" and="" margin.="" these="" changes="" will="" also="" provide="" bases="" for="" the="" associated="" ts="" section.="" the="" proposed="" changes="" bring="" tss="" into="" agreement="" with="" plant="" design="" specifications.="" therefore,="" the="" proposed="" ts="" changes="" do="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 10="" cfr="" 50.92(c)="" are="" satisfied.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" [[page="" 66602]]="" local="" public="" document="" room="" location:="" pottstown="" public="" library,="" 500="" high="" street,="" pottstown,="" pa="" 19464.="" attorney="" for="" licensee:="" j.w.="" durham,="" sr.,="" esquire,="" sr.="" v.p.="" and="" general="" counsel,="" peco="" energy="" company,="" 2301="" market="" street,="" philadelphia,="" pa="" 19101.="" nrc="" project="" director:="" robert="" a.="" capra.="" public="" service="" electric="" &="" gas="" company,="" docket="" no.="" 50-354,="" hope="" creek="" generating="" station,="" salem="" county,="" new="" jersey="" date="" of="" amendment="" request:="" october="" 22,="" 1998.="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" would="" revise="" technical="" specification="" (ts)="" 4.8.2.1.b.3="" to="" increase="" the="" minimum="" battery="" electrolyte="" temperature="" limit="" from="" 60="" deg.f="" to="" 72="" deg.f.="" this="" change="" resolves="" a="" discrepancy="" in="" the="" electrolyte="" temperature="" assumed="" in="" the="" class="" 1-e="" battery="" sizing="" calculations="" versus="" the="" limit="" specified="" in="" the="" tss.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" (1)="" the="" proposed="" changes="" do="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" ts="" change="" does="" not="" involve="" any="" physical="" changes="" to="" plant="" structures,="" systems="" or="" components="" (ssc).="" the="" class-1e="" batteries="" will="" continue="" to="" function="" as="" designed.="" the="" class-1e="" battery="" system="" is="" designed="" to="" mitigate="" the="" consequences="" of="" an="" accident,="" and="" therefore,="" can="" not="" contribute="" to="" the="" initiation="" of="" any="" accident.="" the="" proposed="" ts="" surveillance="" testing="" and="" monitoring="" requirements="" will="" continue="" to="" ensure="" that="" the="" class-1e="" batteries="" are="" capable="" of="" performing="" their="" required="" safety="" functions.="" in="" addition,="" this="" proposed="" ts="" change="" will="" not="" increase="" the="" probability="" of="" occurrence="" of="" a="" malfunction="" of="" any="" plant="" equipment="" important="" to="" safety,="" since="" the="" manner="" i[n]="" which="" the="" class-1e="" battery="" system="" is="" operated="" is="" not="" affected="" by="" these="" proposed="" changes.="" the="" proposed="" changes="" merely="" establish="" ts="" surveillance="" acceptance="" criteria="" that="" more="" appropriately="" reflect="" the="" actual="" plant="" design.="" therefore,="" the="" proposed="" ts="" changes="" would="" not="" result="" in="" an="" increase="" of="" the="" consequences="" of="" an="" accident="" previously="" evaluated.="" therefore,="" the="" proposed="" ts="" change="" does="" not="" involve="" an="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" (2)="" the="" proposed="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" the="" proposed="" ts="" changes="" do="" not="" involve="" any="" physical="" changes="" to="" the="" design="" of="" plant="" systems,="" structures="" or="" components.="" the="" design="" and="" operation="" of="" the="" class-1e="" battery="" system="" is="" not="" changed="" from="" that="" currently="" described="" in="" the="" [updated="" final="" safety="" analysis="" report]="" ufsar,="" only="" the="" allocation="" of="" battery="" capacity="" design="" margin="" is="" affected="" by="" the="" increased="" ts="" minimum="" battery="" electrolyte="" temperature="" limit.="" the="" class-1e="" battery="" system="" will="" continue="" to="" function="" as="" designed="" to="" mitigate="" the="" consequences="" of="" an="" accident.="" implementing="" new="" ts="" surveillance="" acceptance="" criteria="" that="" more="" appropriately="" reflect="" the="" actual="" plant="" design="" does="" not="" permit="" plant="" operation="" in="" a="" configuration="" that="" would="" create="" a="" different="" type="" of="" malfunction="" to="" the="" class-1e="" batteries="" than="" any="" previously="" evaluated.="" in="" addition,="" the="" proposed="" ts="" changes="" do="" not="" alter="" the="" conclusions="" described="" in="" the="" ufsar="" regarding="" the="" safety="" related="" functions="" of="" the="" class-1e="" batteries="" or="" their="" support="" systems.="" therefore,="" the="" proposed="" ts="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" previously="" evaluated.="" (3)="" the="" proposed="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" proposed="" ts="" change="" involves="" the="" implementation="" of="" new="" ts="" surveillance="" acceptance="" criteria="" that="" more="" appropriately="" reflect="" the="" actual="" plant="" design.="" the="" new="" ts="" minimum="" battery="" electrolyte="" temperature="" limit="" enables="" the="" class-1e="" battery="" capacity="" margin="" to="" be="" allocated="" in="" a="" manner="" which="" conforms="" to="" hope="" creek's="" current="" licensing="" basis.="" the="" ability="" of="" the="" class-1e="" batteries="" to="" independently="" supply="" their="" required="" loads="" for="" four="" hours="" without="" support="" from="" battery="" chargers="" is="" not="" affected="" by="" these="" proposed="" changes.="" the="" safety-related="" class-1e="" support="" systems="" will="" ensure="" that="" the="" proposed="" ts="" minimum="" electrolyte="" temperature="" limit="" is="" met.="" therefore,="" the="" proposed="" ts="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 10="" cfr="" 50.92(c)="" are="" satisfied.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" local="" public="" document="" room="" location:="" pennsville="" public="" library,="" 190="" s.="" broadway,="" pennsville,="" nj="" 08070.="" attorney="" for="" licensee:="" jeffrie="" j.="" keenan,="" esquire,="" nuclear="" business="" unit--n21,="" p.o.="" box="" 236,="" hancocks="" bridge,="" nj="" 08038.="" nrc="" project="" director:="" robert="" a.="" capra.="" southern="" nuclear="" operating="" company,="" inc.,="" et="" al.,="" docket="" nos.="" 50-424="" and="" 50-425,="" vogtle="" electric="" generating="" plant,="" units="" 1="" and="" 2,="" burke="" county,="" georgia="" date="" of="" amendment="" request:="" october="" 15,="" 1998,="" as="" supplemented="" by="" letter="" dated="" november="" 11,="" 1998.="" description="" of="" amendment="" request:="" the="" proposed="" amendments="" would="" change="" the="" vogtle="" electric="" generating="" plant,="" unit="" 1="" and="" unit="" 2="" facility="" operating="" licenses="" to="" delete="" or="" modify="" certain="" license="" conditions,="" which="" have="" become="" obsolete="" or="" inappropriate.="" in="" addition,="" the="" technical="" specifications="" would="" be="" reconstituted="" to="" reflect="" revised="" word="" processing.="" no="" change="" in="" technical="" requirements="" would="" be="" involved;="" however,="" the="" font="" would="" be="" changed="" to="" arial="" 11="" point;="" page="" numbers="" would="" be="" revised="" to="" a="" limiting="" condition="" for="" operation="" specific="" numbering="" scheme;="" and="" intentional="" blank="" pages="" would="" be="" deleted.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" the="" proposed="" changes="" do="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" changes="" either="" remove="" or="" modify="" provisions="" in="" the="" vegp="" [vogtle="" electric="" generating="" plant]="" unit="" 1="" and="" [unit]="" 2="" operating="" licenses="" that="" have="" been="" completed="" or="" are="" otherwise="" obsolete.="" each="" proposed="" change="" is="" summarized="" below:="" certain="" surveillance="" requirements="" (srs)="" that="" were="" either="" added="" or="" modified="" at="" the="" time="" of="" improved="" technical="" specifications="" (its)="" implementation="" were="" listed="" in="" the="" operating="" licenses="" with="" a="" schedule="" for="" performance.="" with="" the="" exception="" of="" unit="" 2="" sr="" 3.8.1.20,="" all="" srs="" are="" deleted="" from="" the="" operating="" licenses,="" because="" they="" have="" since="" been="" performed="" according="" to="" schedule,="" and="" will="" henceforth="" be="" performed="" in="" accordance="" with="" the="" technical="" specifications.="" a="" condition="" concerning="" changes="" to="" the="" unit="" 1="" initial="" test="" program="" is="" deleted="" due="" to="" the="" completion="" of="" the="" program.="" a="" condition="" related="" to="" fema="" [federal="" emergency="" management="" agency]="" procedures="" and="" the="" emergency="" plan="" is="" deleted="" from="" the="" unit="" 1="" license="" due="" to="" the="" obsolescence="" of="" the="" condition.="" conditions="" requiring="" the="" submission="" of="" unit="" 1="" reports="" concerning="" the="" steam="" generator="" tube="" rupture="" analysis,="" the="" reactor="" vessel="" level="" instrumentation="" system,="" the="" safety="" parameter="" display="" system,="" the="" detailed="" control="" room="" design="" review,="" and="" the="" zinc="" coating="" of="" the="" diesel="" fuel="" storage="" tanks="" are="" deleted="" due="" to="" completion="" of="" the="" required="" activities.="" a="" condition="" requiring="" modification="" of="" the="" unit="" 1="" ventilation="" exhaust="" of="" the="" alternate="" radwaste="" facility="" is="" deleted="" due="" to="" completion="" of="" the="" required="" activity.="" an="" exemption="" related="" to="" the="" seismic="" adequacy="" of="" the="" unit="" 1="" spent="" fuel="" racks="" is="" deleted="" because="" the="" required="" actions="" are="" completed="" and="" the="" exemption="" has="" been="" determined="" to="" be="" no="" longer="" in="" effect.="" a="" condition="" in="" both="" the="" unit="" 1="" and="" unit="" 2="" licenses="" containing="" reporting="" requirements="" for="" other="" license="" conditions="" is="" revised="" due="" to="" ambiguities="" between="" the="" requirements="" in="" the="" license="" condition="" and="" those="" published="" in="" nrc="" regulations.="" a="" schedular="" exemption="" for="" the="" unit="" 2="" decommissioning="" funding="" report="" is="" deleted="" [[page="" 66603]]="" because="" the="" report="" was="" submitted="" as="" required="" and="" the="" exemption="" is="" no="" longer="" in="" effect.="" the="" technical="" specifications="" and="" associated="" bases="" have="" been="" converted="" from=""> for DOS version 5.1 to 
    Microsoft Word 97. There were no changes to technical 
    requirements. The only visible changes to the document are as 
    follows: (1) the font was changed to Arial 11 point; [(2)] page 
    numbers were revised to an LCO [limiting condition for operation] 
    specific numbering scheme; and [(3)] intentionally blank pages were 
    deleted.
        The proposed changes discussed above are strictly 
    administrative/editorial and do not affect the operation or function 
    of any plant system, component, or structure. Therefore, the 
    proposed changes do not increase the probability of occurrence or 
    the consequences of a previously evaluated accident.
        2. The proposed changes do not create the possibility of a new 
    and different type of accident from any previously evaluated.
        The proposed administrative/editorial changes do not alter the 
    operation of any plant system or equipment and do not introduce a 
    new mode of operation. Each requirement contained in the license 
    conditions proposed for deletion has either been completed or is 
    obsolete. Since these parts of the license are no longer applicable, 
    deletion of these items does not provide the potential for an 
    accident to be created. The conversion of the Technical 
    Specifications from one word processing format to another did not 
    involve any changes to technical requirements. Thus, the proposed 
    changes cannot create a new accident initiating mechanism, and do 
    not create the possibility of a new and different type of accident 
    from any previously evaluated.
        3. The proposed changes do not involve a significant reduction 
    in the margin of safety.
        The license conditions proposed for deletion are obsolete and 
    each requirement has been completed. The conversion of the Technical 
    Specifications from one word processing format to another did not 
    involve any changes to technical requirements. Since the proposed 
    changes are strictly administrative/editorial and do not involve any 
    physical or procedural changes to the plant, the margin of safety, 
    as defined in the bases for any Technical Specification is not 
    affected by the proposed changes.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Burke County Public Library, 
    412 Fourth Street, Waynesboro, Georgia.
        Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
    NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
    Georgia.
        NRC Project Director: Herbert N. Berkow.
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
    
        Date of application for amendments: November 16, 1996 (TS 98-06).
        Brief description of amendments: The proposed amendments would 
    change the Sequoyah Nuclear Plant Technical Specifications (TSs) by 
    revising the emergency diesel generator (EDG) surveillance requirements 
    (SRs) to add a note that allows the SR to be performed in Modes 1, 2, 3 
    or 4, if the associated components are already out-of-service for 
    testing or maintenance and to remove the SR that verifies certain 
    lockout features prevent EDG starting.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the Tennessee Valley 
    Authority (TVA), the licensee, has provided its analysis of the issue 
    of no significant hazards consideration, which is presented below:
    
        TVA has concluded that operation of SQN Units 1 and 2, in 
    accordance with the proposed change to the TSs, does not involve a 
    significant hazards consideration. TVA's conclusion is based on its 
    evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
    standards set forth in 10 CFR 50.92(c).
        A. The proposed amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The probability of occurrence or the consequences for an 
    accident or malfunction of equipment is not increased by this 
    request. The proposal does not alter the way any structure, system 
    or component functions, does not modify the manner in which the 
    plant is operated, and does not alter equipment out-of-service time. 
    This request does not degrade the ability of the D/G [emergency 
    diesel generator] or equipment downstream of the load sequencers to 
    perform their intended function. Deleting the surveillance of a 
    nonsafety-related equipment protection function from TS likewise 
    does not change the probability or consequences of analyzed accident 
    scenarios. Dose consequences remain unchanged by this request.
        B. The proposed amendment does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        A possibility for an accident or malfunction of a different type 
    than any evaluated previously in SQN's FSAR [Final Safety Analysis 
    Report] is not created; nor is the possibility for an accident or 
    malfunction of a different type. The proposal does not alter the way 
    any structure, system or component functions and does not modify the 
    manner in which the plant is operated.
        C. The proposed amendment does not involve a significant 
    reduction in a margin of safety.
        The margin of safety has not been reduced since the test 
    methodologies are not being changed and LCO [Limiting Condition for 
    Operation] allowed outage times are not being changed. Deleting the 
    surveillance of a nonsafety-related equipment protection function 
    from TS likewise does not reduce the margin of safety. The results 
    of accident analysis remain unchanged by this request.
    
        The NRC has reviewed the licensee's analysis and, based on this 
    review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
        NRC Project Director: Frederick J. Hebdon.
    
    The Cleveland Electric Illuminating Company, Centerior Service Company, 
    Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
    Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
    Plant, Unit 1, Lake County, Ohio
    
        Date of amendment request: October 27, 1998.
        Description of amendment request: The proposed amendment would 
    modify the existing Minimum Critical Power Ratio (MCPR) Safety Limit 
    contained in Technical Specification 2.1.1.2. The change would apply 
    additional conservatism by modifying the MCPR Safety Limit values, as 
    calculated by General Electric, by maintaining the limit of 1.09 for 
    two recirculation loop operation and by increasing the limit from 1.10 
    to 1.11 for single loop operation.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration which is presented below:
    
        1. The proposed change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        There is no change to any plant equipment. Per USAR Section 
    4.2.1, the fuel system design bases are provided in General Electric 
    Standard Application for Reactor Fuel (GESTAR II). The Minimum 
    Critical Power Ratio (MCPR) Safety Limit protects the fuel in 
    accordance with the design basis. The MCPR Safety Limit calculations 
    limit the bundle power to ensure the critical power ratio remains 
    unchanged. Therefore, there is not an increase in the probability of 
    transition boiling. The basis of the MCPR Safety Limit calculation 
    remains the same,
    
    [[Page 66604]]
    
    ensuring that greater than 99.9% of all fuel rods in the core avoid 
    transition boiling if the limit is not violated. Therefore, there is 
    no increase in the probability of the occurrence of a previously 
    analyzed accident.
        The fundamental sequences of accidents and transients have not 
    been altered. The MCPR Operating Limits are selected such that 
    potentially limiting plant transients and accidents prevent the MCPR 
    from decreasing below the MCPR Safety Limit anytime during the 
    transient. Therefore, there is no impact on any of the limiting USAR 
    Appendix 15B transients. The radiological consequences are the same 
    as previously stated in the USAR, and as approved in the NRC Safety 
    Evaluation for GESTAR II. Therefore, the consequences of an accident 
    do not increase over previous evaluations in the USAR.
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The MCPR Safety Limit values are designed to ensure that fuel 
    damage from transition boiling does not occur in at least 99.9% of 
    the fuel rods in the core as a result of the limiting postulated 
    accident. The values are calculated in accordance with GESTAR II and 
    the fuel vendor's interim implementing procedures, which incorporate 
    cycle-specific parameters.
        The GESTAR II analysis has been accepted by the NRC as 
    comprehensive for ensuring that fuel designs will perform within 
    acceptable bounds. The MCPR Safety Limit ensures that the fuel is 
    protected in accordance with the design basis. The function, 
    location, operation, and handling of the fuel remain unchanged. In 
    addition, the initiating sequence of events has not changed. 
    Therefore, no new or different kind of accident is created.
        3. The proposed change does not involve a significant reduction 
    in a margin of safety.
        The MCPR Safety Limit values do not alter the design or function 
    of any plant system, including the fuel. The new MCPR Safety Limit 
    values were calculated using NRC-approved methods described in 
    GESTAR II and the fuel vendor's interim implementing procedures, 
    which incorporate cycle-specific parameters. The MCPR Safety Limit 
    values are consistent with GESTAR II, the NRC Safety Evaluation of 
    GESTAR II, the NRC Safety Evaluation Report for the Perry Nuclear 
    Power Plant and its Supplements for USAR Sections 4.4.1 and 
    15.0.3.3.1, and the Technical Specification Bases (Section 2.1.1.2) 
    for the MCPR Safety Limit. This change incorporates a cycle-specific 
    MCPR Safety Limit, as opposed to relying on the generic limit. 
    Therefore, the implementation of the proposed change to the MCPR 
    Safety Limit does not involve a reduction in the margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Perry Public Library, 3753 
    Main Street, Perry, OH 44081.
        Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
    Trowbridge, 2300 N Street, NW., Washington, DC 20037.
        NRC Project Director: Stuart A. Richards.
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of application request: October 27, 1998 (supersedes the April 
    12, 1996, amendment request). This notice supersedes the staff's 
    proposed no significant hazards consideration determination evaluation 
    for the requested changes that was published on May 8, 1996 (61 FR 
    20858).
        Description of amendment request: The proposed amendment 
    application would change the technical specifications (TS) for the 
    reactor coolant system and associated Bases to allow the installation 
    of electrosleeves in the Callaway steam generators for two fuel cycles.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
    
        1. The proposed change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        The electrosleeve configuration has been designed and analyzed 
    in accordance with the requirements of the ASME [American Society of 
    Mechanical Engineers] Code. The applied stresses and fatigue usage 
    for the sleeve are bounded by the limits established in the ASME 
    Code. ASME Code minimum material property values are used for the 
    structural and plugging limit analysis. Mechanical testing has shown 
    that the structural strength of nickel electrosleeves under normal, 
    upset and faulted conditions provides margin to the acceptance 
    limits. These acceptance limits bound the most limiting (3 times 
    normal operating pressure differential) burst margin recommended by 
    RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\'', 
    \11/16\'' and \3/4\'' tube sleeves has demonstrated that no 
    unacceptable levels of primary to secondary leakage are expected 
    during any plant condition.
        The sleeve nominal wall thickness (used for developing the 
    depth-based plugging limit for the sleeve) is determined using the 
    guidance of Regulatory Guide 1.121 and the pressure stress equation 
    of Section III of the ASME Code. The limiting requirement of 
    Regulatory Guide 1.121, which applies to part throughwall 
    degradation, is that the minimum acceptable wall must maintain a 
    factor of safety of three against tube failure under normal 
    operating (design) conditions. A bounding set of design and 
    transient loading input conditions was used for the minimum wall 
    thickness evaluation in the generic evaluation. Evaluation of the 
    minimum acceptable wall thickness for normal, upset and postulated 
    accident condition loading per the ASME Code indicates these 
    conditions are bounded by the design condition requirement minimum 
    wall thickness.
        A bounding tube wall degradation growth rate per cycle and a NDE 
    [Non-Destructive Examination] uncertainty has been assumed for 
    determining the sleeve TS plugging limit. The sleeve wall 
    degradation extent is determined by NDE. The degradation which would 
    require plugging sleeved tubes is developed using the guidance of RG 
    1.121 and is defined in BAW-10219P, to be 20% throughwall for any 
    service induced degradation.
        The consequences of failure of the sleeve are bounded by the 
    current steam generator tube rupture analysis included in the 
    Callaway FSAR [Final Safety Analysis Report]. Due to the slight 
    reduction in diameter caused by the sleeve wall thickness, primary 
    coolant release rates would be slightly less than assumed for the 
    steam generator tube rupture analysis (depending on the break 
    location), and therefore, would result in lower total primary fluid 
    mass release to the secondary system.
        A risk assessment for installation of Electrosleeves at Callaway 
    Plant was performed for a two-cycle operating period. The results of 
    this evaluation determined that sufficient margins against 
    postulated tube rupture during bounding accident conditions exist 
    for all types of degradation of the Electrosleeve material. The 
    calculated probability of burst for a hypothetical population of 
    10,000 axial flaws, 100% throughwall of the parent tube and 0.40'' 
    long, is 4.4 x 10-11 at the end of the second operating cycle. The 
    probability of burst for postulated circumferential flaws and pits 
    is determined to be essentially zero.
        The proposed change does not adversely impact any other 
    previously evaluated design basis accident or the results of LOCA 
    [Loss of Coolant Accident] and non-LOCA accident analyses for the 
    current technical specification minimum reactor coolant system flow 
    rate. The results of the analyses and testing demonstrate that the 
    electrosleeve is an acceptable means of maintaining tube integrity. 
    Furthermore, per Regulatory Guide 1.83 recommendations, the sleeved 
    tube can be monitored through periodic inspections with present NDE 
    techniques. These measures demonstrate that installation of sleeves 
    spanning degraded areas of the tube will restore the tube to a 
    condition consistent with its original design basis.
        Conformance of the electrosleeve design with the applicable 
    sections of the ASME Code and results of the leakage and mechanical 
    tests, support the conclusion that installation of electrosleeves 
    will not involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
    
    [[Page 66605]]
    
        Electrosleeving does not represent a potential to adversely 
    affect any plant component. Stress and fatigue analysis of the 
    repair has shown that the ASME Code and Regulatory Guide 1.121 
    criteria are not exceeded. Implementation of electrosleeving 
    maintains overall tube bundle structural and leakage integrity at a 
    level consistent to that of the originally supplied tubing during 
    all plant conditions. Leak and mechanical testing of electrosleeves 
    support the conclusions of the calculations that each sleeve retains 
    both structural and leakage integrity during all conditions. 
    Sleeving of tubes does not provide a mechanism resulting in an 
    accident outside of the area affected by the sleeves. Any accident 
    as a result of potential tube or sleeve degradation in the repaired 
    portion of the tube is bounded by the existing tube rupture accident 
    analysis.
        Implementation of sleeving will reduce the potential for primary 
    to secondary leakage during a postulated steam line break while not 
    significantly impacting available primary coolant flow area in the 
    event of a LOCA. By effectively isolating degraded areas of the tube 
    through repair, the potential for steam line break leakage is 
    reduced. These degraded intersections now are returned to a 
    condition consistent with the Design Basis. While the installation 
    of a sleeve reduces primary coolant flow, the reduction is far below 
    that caused by plugging. Therefore, far greater primary coolant flow 
    area is maintained through sleeving versus plugging.
        3. The proposed change does not involve a significant reduction 
    in a margin of safety.
        The electrosleeve repair of degraded steam generator tubes has 
    been shown by analysis to restore the integrity of the tube bundle 
    consistent with its original design basis condition, i.e., tube/
    sleeve operational and faulted condition stresses are bounded by the 
    ASME Code requirements and the repaired tubes are leaktight. The 
    safety factors used in the design of sleeves for the repair of 
    degraded tubes are consistent with the safety factors in the ASME 
    Code used in steam generator design. The portions of the installed 
    sleeve assembly which represent the reactor coolant pressure 
    boundary can be monitored for the initiation and progression of 
    sleeve/tube wall degradation, thus satisfying the requirements of 
    Regulatory Guide 1.83. The portion of the tube bridged by the sleeve 
    is effectively removed from the pressure boundary, and the sleeve 
    then forms the new pressure boundary. The areas of the sleeved tube 
    assembly which require inspection are defined in BAW-10219P.
        In addition, since the installed sleeve represents a portion of 
    the pressure boundary, a baseline inspection of these areas is 
    required prior to operation with sleeves installed. The effect of 
    sleeving on the design transients and accident analyses has been 
    reviewed based on the installation of sleeves up to the level of 
    steam generator tube plugging coincident with the minimum reactor 
    flow rate and the Callaway Safety Analysis.
        Provisional requirements cited in other NRC Safety Evaluation 
    Reports addressing the implementation of sleeving have required the 
    reduction of the individual steam generator normal operation primary 
    to secondary leakage limit from 500 to 150 gpd [gallons per day]. 
    Consistent with these evaluations, Union Electric will reduce the 
    per steam generator leak rate of 500 gpd in TS 3.4.6.2.c to 150 gpd. 
    The establishment of this leakage limit at 150 gpd provides 
    additional safety margin. [The staff notes that this leakage limit 
    has been incorporated into the Callaway Technical Specifications via 
    license amendment #119 dated October 1, 1996.]
        Finally, Union Electric will reduce the tube plugging limit from 
    48% through wall to 40% through wall to be consistent with NUREG-
    1431. The establishment of the plugging limit at 40% through wall 
    provides additional safety margin. [The staff notes that this 
    plugging limit has been incorporated into the Callaway Technical 
    Specifications via license amendment #119 dated October 1, 1996.]
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: University of Missouri-
    Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    & Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
        NRC Project Director: William H. Bateman.
    
    Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
    Yankee Nuclear Power Station, Vernon, Vermont
    
        Date of amendment request: November 3, 1998.
        Description of amendment request: The licensee proposes to make 
    administrative changes to the Technical Specifications to correct 
    errors, add consistency within the Technical Specifications, and make 
    nomenclature changes to support and enhance usability of the Technical 
    Specifications.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration which is presented below:
    
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated, because:
        The proposed changes are purely administrative in nature and 
    have no effect on plant hardware, plant design, safety limit 
    setting, or plant system operation and therefore do not modify or 
    add any initiating parameters that would significantly increase the 
    probability or consequences of an accident previously evaluated.
        No new modes of operation are introduced by the proposed changes 
    such that adverse consequences would result. Accordingly, the 
    consequences of previously analyzed accidents are not affected by 
    this proposed license amendment.
        2. Create the possibility of a new or different kind of accident 
    from any accident previously evaluated, because:
        These changes do not affect the operation of any systems or 
    components, nor do they involve any potential initiating events that 
    would create any new or different kind of accident. Therefore, the 
    proposed changes do not create the possibility of a new or different 
    kind of accident from any accident previously evaluated for the 
    Vermont Yankee Nuclear Power Station.
        3. Involve a significant reduction in a margin of safety, 
    because:
        These proposed changes do not affect any equipment involved in 
    potential initiating events or safety limits. Therefore, it is 
    concluded that the proposed changes do not involve a significant 
    reduction in a margin of safety.
        Administrative changes, as such, do not constitute any 
    significant hazards considerations.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Brooks Memorial Library, 224 
    Main Street, Brattleboro, VT 05301.
        Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
    Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
        NRC Project Director: Cecil O. Thomas.
    
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
    North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
    Virginia
    
        Date of amendment request: November 10, 1998.
        Description of amendment request: The proposed changes to North 
    Anna Power Station (NAPS), Units 1 and 2, Technical Specification (TS) 
    3.4.4 will clarify the operability requirements for the pressurizer 
    heaters and eliminate a potential verbatim compliance issue associated 
    with the pressurizer heaters and emergency power supply. The verbatim 
    compliance issue was created when the Emergency Diesel Generator 
    allowed outage time was changed from 72 hours to 14 days.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the
    
    [[Page 66606]]
    
    licensee has provided its analysis of the issue of no significant 
    hazards consideration, which is presented below:
    
        Virginia Electric and Power Company has reviewed the 
    requirements of 10 CFR 50.92 as they relate to the proposed changes 
    for the North Anna Units 1 and 2 and determined that a significant 
    hazards consideration is not involved. The proposed changes will 
    revise the LCO [limiting condition for operation] 3.4.4 to require 
    that the pressurizer have two groups of pressurizer heaters operable 
    with a capacity of greater than or equal to 125 kW and capable of 
    being powered from its associated emergency bus. The Action 
    Statement will also be revised to focus on heater operability. The 
    following is provided to support this conclusion.
        (a) Does the change involve a significant increase in the 
    probability or consequences of an accident previously evaluated?
        The pressurizer heaters are not an initiator of any accident 
    previously evaluated. As a result, the probability of any accident 
    previously evaluated is not increased. The pressurizer heaters 
    remain operable as assumed in the accident analysis to mitigate the 
    consequences of any accident. Therefore, the proposed changes to 
    clarify the operability requirements do not significantly increase 
    the probability of occurrence or the consequences of any previously 
    analyzed accident.
        (b) Does the change create the possibility of a new or different 
    kind of accident from any accident previously evaluated?
        The proposed Technical Specifications changes do not involve any 
    physical alteration of the plant or changes in methods governing 
    normal plant operation. Operation of and the design of the 
    pressurizer heaters and the associated power supplies are not 
    changed by the proposed changes. The proposed changes do not impose 
    any new or eliminate any existing requirements. Therefore, it is 
    concluded that no new or different kind of accident or malfunction 
    from any previously evaluated has been created.
        (c) Does the change involve a significant reduction in a margin 
    of safety?
        The proposed Technical Specifications changes will not reduce 
    the margin of safety since the change has no effect on any safety 
    analyses assumptions. The pressurizer heaters remain operable as 
    assumed in the safety analysis to mitigate the consequences of any 
    accident previously analyzed. The proposed changes only clarify the 
    operability requirements for the pressurizer heaters and associated 
    emergency power supplies. Therefore, the proposed changes do not 
    result in a significant reduction in a margin of safety.
    
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: The Alderman Library, Special 
    Collections Department, University of Virginia, Charlottesville, 
    Virginia 22903-2498.
        Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
    Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
    23219.
        NRC Project Director: Herbert N. Berkow.
    
    Previously Published Notices of Consideration of Issuance of 
    Amendments to Facility Operating Licenses, Proposed No Significant 
    Hazards Consideration Determination, and Opportunity for a Hearing
    
        The following notices were previously published as separate 
    individual notices. The notice content was the same as above. They were 
    published as individual notices either because time did not allow the 
    Commission to wait for this biweekly notice or because the action 
    involved exigent circumstances. They are repeated here because the 
    biweekly notice lists all amendments issued or proposed to be issued 
    involving no significant hazards consideration.
        For details, see the individual notice in the Federal Register on 
    the day and page cited. This notice does not extend the notice period 
    of the original notice.
    
    Florida Power and Light Company, et al., Docket Nos. 50-335, and 50-
    389, St. Lucie Plant, Unit Nos. 1, and 2, St. Lucie County, Florida
    
        Date of amendment request: October 29, 1998.
        Description of amendment request: Technical Specification changes 
    (TS) relating to the implementation and automatic removal of certain 
    reactor protection system trip bypasses to ensure that the meaning of 
    explicit terms used in the TSs are consistent with the intent of the 
    stated requirements.
        Date of publication of individual notice in the Federal Register: 
    November 5, 1998 (63 FR 59809).
        Expiration date of individual notice: November 19, 1998.
        Local Public Document Room location: Indian River Junior College 
    Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    
    Notice of Issuance of Amendments to Facility Operating Licenses
    
        During the period since publication of the last biweekly notice, 
    the Commission has issued the following amendments. The Commission has 
    determined for each of these amendments that the application complies 
    with the standards and requirements of the Atomic Energy Act of 1954, 
    as amended (the Act), and the Commission's rules and regulations. The 
    Commission has made appropriate findings as required by the Act and the 
    Commission's rules and regulations in 10 CFR Chapter I, which are set 
    forth in the license amendment.
        Notice of Consideration of Issuance of Amendment to Facility 
    Operating License, Proposed No Significant Hazards Consideration 
    Determination, and Opportunity for A Hearing in connection with these 
    actions was published in the Federal Register as indicated.
        Unless otherwise indicated, the Commission has determined that 
    these amendments satisfy the criteria for categorical exclusion in 
    accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
    no environmental impact statement or environmental assessment need be 
    prepared for these amendments. If the Commission has prepared an 
    environmental assessment under the special circumstances provision in 
    10 CFR 51.12(b) and has made a determination based on that assessment, 
    it is so indicated.
        For further details with respect to the action see: (1) The 
    applications for amendment, (2) the amendment, and (3) the Commission's 
    related letter, Safety Evaluation and/or Environmental Assessment as 
    indicated. All of these items are available for public inspection at 
    the Commission's Public Document Room, the Gelman Building, 2120 L 
    Street, NW., Washington, DC, and at the local public document rooms for 
    the particular facilities involved.
    
    Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs 
    Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
    
        Date of application for amendment: July 20, 1998.
        Brief description of amendment: The amendment implements a 
    modification that constitutes an unreviewed safety question as 
    described in 10 CFR 50.59. The modification involves replacing the 
    service water heat exchangers with new plate and frame heat exchangers 
    having an increased thermal performance capability. The planned 
    modification is similar to the one completed on Unit 1. In addition, by 
    a separate letter dated July 20, 1998, the licensee submitted a request 
    to obtain approval for a temporary one time cooling lineup needed to 
    support emergency diesel generator operability for the installation of 
    the Unit 2 service water heat exchanger replacement, which is currently 
    being reviewed by the NRC
    
    [[Page 66607]]
    
    staff. Therefore, since the implementation of the proposed service 
    water heat exchanger modification is dependent on the staff's issuance 
    of the one time Technical Specification (TS) change regarding 
    installation of the modification, this modification should not be 
    implemented prior to the issuance of the one-time TS change for 
    installing the modification.
        Date of issuance: November 5, 1998.
        Effective date: This license amendment is effective as of the date 
    of its issuance to be implemented after the staff's issuance of the 
    one-time TS change regarding the installation of the service water heat 
    exchanger modification.
        Amendment No.: 203.
        Facility Operating License No. DPR-69: Amendment revised the 
    Updated Final Safety Analysis Report.
        Date of initial notice in Federal Register: August 12, 1998 (63 FR 
    43201).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 5, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Calvert County Library, Prince 
    Frederick, Maryland 20678.
    
    Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
    Station, Plymouth County, Massachusetts
    
        Date of application for amendment: June 26, 1998.
        Brief description of amendment: The amendment modifies various 
    Technical Specification pages to correct typographical errors, remove 
    inadvertent replication of information, and updates various Bases 
    sections.
        Date of issuance: November 10, 1998.
        Effective date: November 10, 1998.
        Amendment No: 178.
        Facility Operating License No. DPR-35: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: September 23, 1998 (63 
    FR 50933).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 10, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Plymouth Public Library, 11 
    North Street, Plymouth, Massachusetts 02360.
    
    Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
    Electric Plant, Unit No. 2, Darlington County, South Carolina
    
        Date of application for amendment: March 6, 1998, as supplemented 
    September 11, 1998. The September 11, 1998, supplemental letter 
    contained clarifying information only, and did not change the no 
    significant hazards consideration determination.
        Brief description of amendment: The amendment revises Technical 
    Specification 3.9.2 relating to the use of Post-Accident Monitoring 
    Source Range neutron flux detectors as a compensatory measure in the 
    event that one of the two required BF3 neutron flux detectors becomes 
    inoperable during Mode 6 operations (refueling).
        Date of issuance: November 12, 1998.
        Effective date: November 12, 1998.
        Amendment No: 180.
        Facility Operating License No. DPR-23: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 3, 1998 (63 FR 
    30262).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 12, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Hartsville Memorial Library, 
    147 West College Avenue, Hartsville, South Carolina 29550.
    
    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
    Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
    
        Date of application of amendments: September 17, 1998, as 
    supplemented October 15, 1998.
        Brief description of amendments: The amendments revised the Updated 
    Final Safety Analysis Report to perform a Keowee Emergency Power 
    Engineered Safeguards Functional Test during the 1998 Unit 3 refueling 
    outage at Oconee.
        Date of Issuance: November 12, 1998.
        Effective date: As of the date of issuance to be implemented during 
    the 1998 Unit 3 refueling outage.
        Amendment Nos.: Unit 1--233; Unit 2--233; Unit 3--232.
        Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
    Amendments revised the Updated Final Safety Analysis Report.
        Date of initial notice in Federal Register: September 30, 1998 (63 
    FR 52304).
        The October 15, 1998, letter provided clarifying information that 
    did not change the scope of the September 17, 1998, application and the 
    initial proposed no significant hazards consideration determination.
        The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated November 12, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Oconee County Library, 501 
    West South Broad Street, Walhalla, South Carolina.
    
    GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
    Station, Dauphin County, Pennsylvania
    
        Date of application for amendment: December 2, 1996.
        Brief description of amendment: This amendment would revise audit 
    frequency requirements and relocate them from the Technical 
    Specifications to the Quality Assurance Plan.
        Date of issuance: November 12, 1998.
        Effective date: This amendment is effective immediately to be 
    implemented written 60 days.
        Amendment No.: 52.
        Facility Operating License No. DPR-73: The amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 30, 1997 (62 FR 
    40850).
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Government Publications 
    Section, State Library of Pennsylvania Walnut Street and Commonwealth 
    Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    
    Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
    362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
    Diego County, California
    
        Date of application for amendments: May 11, 1998, as supplemented 
    by letter dated October 9, 1998.
        Brief description of amendments: The amendments modify the 
    technical specifications (TS) for San Onofre Nuclear Generating Station 
    Unit Nos. 2 and 3 to implement 10 CFR Part 50 Appendix J, Option B for 
    performance-based reactor containment leakage testing.
        Date of issuance: November 6, 1998.
        Effective date: November 6, 1998, to be implemented within 30 days 
    from the date of issuance.
        Amendment Nos.: Unit 2 -144; Unit 3 -135.
        Facility Operating License No. NPF-10 and NPF-15: The amendments 
    revised the Technical Specifications.
    
    [[Page 66608]]
    
        Date of initial notice in Federal Register: September 9, 1998 (63 
    FR 48265).
        The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated November 6, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Main Library, University of 
    California, P. O. Box 19557, Irvine, California 92713.
    
    STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
    Texas Project, Units 1 and 2, Matagorda County, Texas
    
        Date of amendment request: July 6, 1998.
        Brief description of amendments: Relocates the description of the 
    reactor coolant system design features in Technical Specification 5.4 
    to the Updated Final Safety Analysis Report, which already contains the 
    information.
        Date of issuance: November 18, 1998.
        Effective date: November 18, 1998, to be implemented within 30 
    days.
        Amendment Nos.: Unit 1--Amendment No. 98; Unit 2--Amendment No. 85.
        Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: September 9, 1998 (63 
    FR 48266).
        The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated November 18, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Wharton County Junior College, 
    J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    
    STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
    Texas Project, Units 1 and 2, Matagorda County, Texas.
    
        Date of amendment request: July 6, 1998 , as supplemented on 
    October 28, 1998.
        Brief description of amendments: Relocate the Technical 
    Specification 3/4.3.3.3 requirements for Seismic Instrumentation to the 
    Technical Requirements Manual.
        Date of issuance: November 18, 1998.
        Effective date: November 18, 1998, to be implemented within 30 
    days.
        Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 86.
        Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: September 9, 1998 (63 
    FR 48267).
        The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated November 18, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Wharton County Junior College, 
    J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    
    STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
    Texas Project, Units 1 and 2, Matagorda County, Texas.
    
        Date of amendment request: July 6, 1998, as supplemented on October 
    28, 1998.
        Brief description of amendments: Relocates the Technical 
    Specification 3/4.7.13 requirements for the Area Temperature Monitoring 
    System to the Technical Requirements Manual.
        Date of issuance: November 18, 1998.
        Effective date: November 18, 1998, to be implemented within 30 
    days.
        Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No. 
    87.
        Facility Operating License Nos. NPF-76 and NPF-80: The amendment 
    revises the Technical Specifications.
        Date of initial notice in Federal Register: September 9, 1998 (63 
    FR 48267). The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated November 18, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Wharton County Junior College, 
    J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.
    
        Date of application for amendments: February 13, 1998 (TS 97-07).
        Brief description of amendments: The amendments incorporate new 
    main steam isolation valve (MSIV) requirements that are consistent with 
    NUREG-1431, the Westinghouse Standard Technical Specifications (TS), 
    including testing requirements for the MSIVs that ensure the valves 
    close on an automatic actuation signal.
        Date of issuance: November 17, 1998.
        Effective date: As of the date of issuance to be implemented no 
    later than 45 days after issuance.
        Amendment Nos.: 236 and 226.
        Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
    revise the technical specifications.
        Date of initial notice in Federal Register: April 22, 1998 (63 FR 
    19980).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 17, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of application for amendments: June 26, 1998 (TS 98-02).
        Brief description of amendments: The amendments change the 
    Technical Specifications and their Bases to lower the specific activity 
    limit for the primary coolant system from 1.0 microcurie/gram dose 
    equivalent iodine-131 to 0.35 microcurie/gram, as provided for in NRC 
    Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse 
    Steam Generator Tubes Affected by Outside Diameter Stress Corrosion 
    Cracking.'' This change allows a proportional increase in main steam 
    line break induced primary-to-secondary leakage when implementing the 
    alternate steam generator tube repair criteria, which the NRC has 
    already approved for Sequoyah Nuclear Plant, Units 1 and 2.
        Date of issuance: November 17, 1998.
        Effective date: As of the date of issuance to be implemented no 
    later than 45 days after issuance.
        Amendment Nos.: 237 and 227.
        Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
    revise the technical specifications.
        Date of initial notice in Federal Register: July 15, 1998 (63 FR 
    38205).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 17, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    
    [[Page 66609]]
    
    Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
    Unit 1, (WBN) Rhea County, Tennessee
    
        Date of application for amendment: August 5, 1998 (TS 98-008).
        Brief description of amendment: This amendment is in response to 
    your application dated August 5, 1998. The amendment revises the WBN 
    Technical Specifications (TS) and associated TS Bases to allow up to 4 
    hours to make the residual heat removal suction relief valve available 
    as a cold overpressure mitigation system relief path.
        Date of issuance: November 10, 1998.
        Effective date: November 10, 1998.
        Amendment No.: 14.
        Facility Operating License No. NPF-90: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: September 23, 1998 (63 
    FR 50940).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 10, 1998.
        No significant hazards consideration comments received: None.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1001 Broad Street, Chattanooga, TN 37402.
    
    TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
    Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
    
        Date of amendment request: July 10, 1996 (TXX-96405), as 
    supplemented by letters dated October 1, 1996 (TXX-96475), and July 1, 
    1998 (TXX-98159).
        Brief description of amendments: The amendment would take credit 
    for the addition of train oriented Fan Coil Units for each UPS and 
    Distribution Room and would provide redundancy to the existing Air 
    Conditioning (A/C) Units (TS 3/4.7.11 and its associated bases).
        Date of Issuance: Date of issuance: November 18, 1998.
        Effective date: November 18, 1998, to be implemented within 30 
    days.
        Amendment Nos.: Unit 1--Amendment No. 61; Unit 2--Amendment No. 47.
        Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: February 12, 1997 (62 
    FR 6579).
        The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated November 18, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: University of Texas at 
    Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
    19497, Arlington, TX 76019.
    
    Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
    Nuclear Power Plant, Kewaunee County, Wisconsin
    
        Date of application for amendment: May 7, 1998.
        Brief description of amendment: This amendment revises Technical 
    Specification 5.4, ``Fuel Storage,'' to increase the allowable mass of 
    uranium-235 (U235) per axial centimeter for fuel storage. 
    The requested change will allow the use of new Siemens Power 
    Corporation heavy fuel assembly designs.
        Date of Issuance: November 12, 1998.
        Effective date: November 12, 1998.
        Amendment No.: 141.
        Facility Operating License No. DPR-43: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 17, 1998 (63 FR 
    33111).
        The Commission's related evaluation of the amendment is contained 
    in a Safety Evaluation dated November 12, 1998.
        No significant hazards consideration comments received: No.
        Local Public Document Room location: University of Wisconsin, 
    Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    
        Dated at Rockville, Maryland, this 24th day of November 1998.
    
        For the Nuclear Regulatory Commission.
    Elinor G. Adensam,
    Acting Director Division of Reactor Projects--III/IV Office of Nuclear 
    Reactor Regulation.
    [FR Doc. 98-31931 Filed 12-1-98; 8:45 am]
    BILLING CODE 7590-01-P
    
    
    

Document Information

Published:
12/02/1998
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
98-31931
Dates:
This license amendment is effective as of the date of its issuance to be implemented after the staff's issuance of the one-time TS change regarding the installation of the service water heat exchanger modification.
Pages:
66590-66609 (20 pages)
PDF File:
98-31931.pdf