96-27025. Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations  

  • [Federal Register Volume 61, Number 206 (Wednesday, October 23, 1996)]
    [Notices]
    [Pages 55028-55049]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: 96-27025]
    
    
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    UNITED STATES 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 September 30, 1996, through October 10, 
    1996. The last biweekly notice was published on October 9, 1996 (61 FR 
    52962).
    
    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. 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 
    Review and Directives Branch, Division of Freedom of Information and 
    Publications 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 November 22, 1996, 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
    
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    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: Docketing and 
    Services Branch, or may be delivered to the Commission's Public 
    Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
    by the above date. Where petitions are filed during the last 10 days of 
    the notice period, it is requested that the petitioner promptly so 
    inform the Commission by a toll-free telephone call to Western Union at 
    1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
    operator should be given Datagram Identification Number N1023 and the 
    following message addressed to (Project Director): petitioner's name 
    and telephone number, date petition was mailed, plant name, and 
    publication date and page number of this Federal Register notice. 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.
    
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
    Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
    North Carolina
    
        Date of amendment request: September 18, 1996
        Description of amendment request: Revise Technical Specification 
    (TS) 4.8.1.1.2 by removing TS 4.8.1.1.2.h.2 pressure testing 
    requirement since adequate testing will be completed in accordance with 
    American Society of Mechanical Engineers (ASME) Boiler and Pressure 
    Vessel Code, Section XI.
        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:
        This change does not involve a significant hazards consideration 
    for the following reasons:
        1. The proposed amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        Applying ASME Code, Section XI alternative examination/testing 
    will not affect any initiators of any previously evaluated accidents 
    or change the manner in which the emergency diesel generators or any 
    other systems operate. The diesel fuel oil system supports the 
    emergency diesel generators which serve an accident mitigating 
    function. Where portions of piping are non-isolable or where 
    atmospheric tanks are involved, the Section XI ASME alternatives to 
    110% pressure testing continue to ensure the integrity of the fuel 
    oil system without any impact on analyzed accident scenarios or 
    their consequences. Therefore, the proposed amendment does not 
    result in an increase in the probability or consequences of an 
    accident previously evaluated.
        2. The proposed amendment does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        The proposed alternative testing and surveillance will not 
    involve any physical alterations or additions to plant equipment or 
    alter the manner in which any safety-related system performs it 
    function. Using ASME Section XI, or NRC-approved ASME Code cases, as 
    guidance for pressure testing continues to provide assurance that 
    the fuel oil supply system will perform its intended function. 
    Therefore, the proposed changes do not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        3. The proposed amendment does not involve a significant 
    reduction in the margin of safety.
        There are no changes being made to the safety limits or safety 
    settings that would adversely impact plant safety. Further, there is 
    no impact on the margin of safety as defined in the Technical 
    Specifications. Utilizing ASME Section XI as guidance for 
    determining those sections of piping that should be pressure-tested 
    or tested at atmospheric pressure will ensure proper operation of 
    the diesel generator fuel oil supply system. Therefore, the proposed 
    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.
        Local Public Document Room location: Cameron Village Regional 
    Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
        Attorney for licensee: William D. Johnson, Vice President and 
    Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
    Raleigh, North Carolina 27602
        NRC Project Director: F. Mark Reinhart, Acting
    
    Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
    Plant, Unit 1, Monroe County, Michigan
    
        Date of amendment request: August 29, 1996 (Reference NRC-96-0111)
        Description of amendment request: The proposed amendment will: (1) 
    allow certain equipment and instruments to be removed from service for 
    short periods of time to allow for
    
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    maintenance, testing, inspection, modifications, and account for 
    equipment failures; (2) reduce the frequency of environmental liquid 
    effluent monitoring and eliminate one raw water sampling location; (3) 
    eliminate the requirement for moisture intrusion monitoring for the 
    reactor building lower level; and (4) correction of a typographical 
    error.
        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 using the standards in 10 CFR 50.92(c). The licensee's 
    analysis is presented below:
        (1) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in 
    accordance with the proposed amendment will not involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident. Provisions for 
    removing the primary cover gas supply from service for short periods 
    of time will not significantly increase the probability of an 
    accident occurring as long as the probability of a significant water 
    reaction with residual sodium is not significantly increased. This 
    is ensured by prescribing limits on the time that carbon dioxide 
    pressure can be low. The consequences of an accident would not be 
    affected by provisions for removing the primary cover gas supply 
    from service as this equipment does not mitigate accidents or affect 
    the accident sequences. Similarly, the provisions for removing the 
    moisture intrusion and cover gas pressure alarms from service for 
    short period of time will not significantly increase the probability 
    of an accident. The alarms provide a monitoring function to detect 
    degradation in the performance of the cover gas supply and sump 
    systems. Absence of these alarm functions for short periods of time 
    does not increase the probability of such degradation and it does 
    not significantly impact the ability for timely detection of such 
    degradation. The consequences of an accident would not be affected 
    by provisions for removing the moisture intrusion and cover gas 
    pressure alarms from service as this equipment does not mitigate 
    accidents or affect the accident sequences. Elimination of the 
    moisture intrusion alarm for the reactor building lower level does 
    not significantly increase the probability of an accident because 
    the probability that water could accumulate in this area is 
    essentially unchanged. Design features of the foundation, 
    containment structure, and annulus drains are intended to prevent 
    entry of water into the reactor building. These features have 
    prevented any water intrusion into this area. The consequences of an 
    accident would not be affected by elimination of the moisture 
    intrusion alarm for the reactor building lower level because this 
    equipment does not mitigate accidents or affect the accident 
    sequences. The Safety Evaluation Supporting Amendment 9 to the 
    referenced license did not rely on moisture intrusion monitoring and 
    alarm features for any safety function or accident prevention or 
    mitigation function. Environmental monitoring surveillance are 
    unrelated to postulated accident sequences and cannot affect the 
    probability or consequences of an accident. The correction of the 
    typographical error is unrelated to accident initiation and 
    sequences and cannot affect the probability or consequences of any 
    accident.
        (2) The operation of Enrico Fermi Atomic Power Plant, Unit 1, 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 proposed changes do not create the possibility of a new or 
    different accident from any previously evaluated. With the exception 
    of the allowance for composite environmental samples, which are 
    unrelated to any potential accident sequence, these changes propose 
    no new activities or new methods for performing existing activities. 
    Previous evaluations have considered the release of all of the 
    radioactivity in the residual sodium due to postulated fire or other 
    catastrophe and release of radioactive water stored in the liquid 
    waste tanks which bound the only possible radiological accidents at 
    Fermi 1. For these reasons, no new or different type of accident is 
    created by these changes.
        (3) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in 
    accordance with the proposed amendment will not involve a 
    significant reduction in a margin of safety.
        The proposed changes do not involve a significant reduction in a 
    margin of safety. The changes to the primary system cover gas system 
    technical specifications still ensure that any residual sodium is 
    passivated by carbon dioxide. Changes to the alarms affect only 
    monitoring functions and therefore do not cause a change to any 
    parameter that could affect the margin of safety. Similarly, the 
    environmental surveillances are unrelated to margin of safety. The 
    correction of the typographical error is unrelated to margin of 
    safety. For these reasons, the proposed changes do not involve a 
    significant 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 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: Monroe County Library System, 
    3700 South Custer Road, Monroe, Michigan 48161
        Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
    2000 Second Avenue, Detroit, Michigan 48226NRC Branch Chief: Michael F. 
    Weber
    
    Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
    Michigan
    
        Date of amendment request: September 25, 1996 (NRC-96-0085)
        Description of amendment request: The proposed amendment would 
    revise Technical Specification (TS) Surveillance Requirement 4.8.4.3 to 
    remove the requirement to periodically test the thermal overload (TOL) 
    devices for safety-related motor-operated valves (MOVs). The 
    surveillance requirement would continue to require testing of a TOL 
    device following any maintenance activity that could affect the 
    performance of the device. The surveillance requirement would also be 
    clarified by indicating that testing of TOL devices is required upon 
    initial installation. The associated portion of the TS Bases would also 
    be revised to reflect this change.
        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. The deletion of 
    the requirement for testing of the TOL protective devices lessens 
    degradation to the components which can improve MOV reliability. 
    Based on historical data through the years of testing, there is no 
    significant drifting of the trip setpoints of the TOL protective 
    devices. The probability of an accident would not increase since 
    terminating the periodic testing or clarifying the situational 
    testing requirements cannot cause equipment to operate inadvertently 
    and so cannot cause an accident. The periodic testing of the TOL 
    protective devices can temporarily render MOVs inoperable due to the 
    removal of the components from service and can cause safety systems/
    divisions to become unavailable. The deletion of the periodic 
    testing requirement would increase the availability of safety 
    systems insuring that they would be able to respond to accident 
    conditions. The consequences of an accident will not increase since 
    eliminating the periodic testing and clarifying the situational 
    testing requirements will improve reliability of safety-related MOVs 
    to respond to an accident and will not increase the failure rate of 
    equipment. The clarification of the situational testing ensures that 
    the test will be conducted after any maintenance that could affect 
    the performance of the TOL protective devices. Thus, the proposed 
    change increases reliability of the MOVs and increases plant safety. 
    Therefore this change will not result in a significant increase in 
    the probability or consequences of an accident.
        2. The proposed change does not create the possibility of a new 
    or different accident from any previously evaluated. The TOL
    
    [[Page 55031]]
    
    protective devices are not an accident initiator, they only protect 
    equipment provided to mitigate the consequences of an accident. For 
    this reason, no new or different type of accident is created by this 
    change.
        3. The proposed change does not involve a significant reduction 
    in a margin of safety. The trip setpoints of the TOL protective 
    devices depend upon both the current and the length of time the 
    current is applied. The trip setpoints for TOL protective devices 
    are much higher than conditions normally experienced during an MOV 
    stroke and are meant to protect the motor from stall and overload 
    conditions. The difference between the current of the trip setpoints 
    and the normal conditions is great enough that a premature trip of 
    the TOL protective device is highly unlikely, even at degraded 
    voltages. The TOL protective device protects the motor from the 
    stall conditions. Not conducting the periodic testing of the TOL 
    protective devices would not cause the MOVs to fail, nor would the 
    performance of the MOVs be adversely affected. Throughout the life 
    of the plant, there has never been an instance of a safety related 
    MOV failure due to degradation or failure of TOL protective devices. 
    Further, based on maintenance history, the elimination of the 
    periodic testing would eliminate any significant potential 
    degradation of the TOL protective devices, thereby increasing their 
    reliability. Finally, with the removal of the periodic testing of 
    the TOL protective devices, fewer MOVs would have to be removed from 
    service for testing. Since necessary components would no longer be 
    inoperable due to the periodic testing, there would be an increase 
    of availability time of safety systems/divisions. Deletion of the 
    periodic testing could reduce the durations of online system 
    outages. Clarifying the situational testing requirements would 
    better define when the testing of the TOL protective devices is 
    necessary which would ensure operability. The testing would be based 
    on installation or any maintenance that could affect the TOL 
    protective device. For these reasons, the proposed change does not 
    involve a significant 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: Monroe County Library System, 
    3700 South Custer Road, Monroe, Michigan 48161
        Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
    2000 Second Avenue, Detroit, Michigan 48226
        NRC Project Director: John N. Hannon
    
    Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear 
    Station, Units 1 and 2, York County, South Carolina
    
        Date of amendment request: June 21, 1996
        Description of amendment request: The proposed amendments would 
    administratively correct the term ``lifting load'' in Technical 
    Specification 3.9.6b.2 to ``lifting force.'' This correction would 
    clarify that the static loads associated with the lifting tool, drive 
    rod and control rod weights are not included in the lifting force 
    limit. The amendments would also more accurately define auxiliary hoist 
    minimum capacities and give a more expansive description of the 
    activities for which protective measures and surveillance testing are 
    used.
        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:
        Question: Will the change involve a significant increase in the 
    probability or consequences of an accident previously evaluated?
        No. The proposed change[s] [are] administrative in nature, and 
    do[] not represent any changes to the refueling process in the 
    field. It more accurately describes the components for which the 
    LCO's [limiting conditions of operation] protection is intended as 
    well as giving a more accurate description of the auxiliary hoist's 
    minimum capacity. [They] also broaden[] the domain of activities for 
    which protective measures are taken, by including drag load testing 
    into monitored activities. At both MNS [McGuire Nuclear Station] and 
    CNS [Catawba Nuclear Station], the auxiliary hoists and the 
    manipulator cranes are rated at [greater than or equal to] 3000 
    pounds and are surveillance tested to greater than 1000 pounds. This 
    brackets the limit force lifting value change from 600 to 1000 
    pounds in the amendment proposal.
        Question: Will the change create the possibility of a new or 
    different kind of accident from any accident previously evaluated?
        No. Th[ese] proposed administrative change[s] reflect[] no 
    changes in the refueling processes, or any systems, structures or 
    components connected with the refueling process.
        Question: Will the change involve a significant reduction in a 
    margin of safety?
        No. The proposed administrative change[s] [have] no impact on 
    refueling processes, systems, structures or components, and do[] not 
    result in any significant reduction in a margin of safety. The 
    subject change[s] only clarif[y] the original intent of the 
    specification and more accurately describe[] the involved 
    components, component capacities and the domain of activities for 
    which measures are taken to protect the reactor internals.
        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 
    proposed amendments involve no significant hazards consideration.
        Local Public Document Room location: York County Library, 138 East 
    Black Street, Rock Hill, South Carolina 29730
        Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
    South Church Street, Charlotte, North Carolina 28242
        NRC Project Director: Herbert N. Berkow
    
    Duke Power Company, 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: September 17, 1996 (TSC 96-01)
        Description of amendment request: The proposed changes would reduce 
    the Reactor Building pressure setpoint for actuation of the Reactor 
    Building Spray System in Technical Specification (TS) 3.5.3 from a 
    maximum of 30 pounds per square inch gauge (psig) to 15 psig, reduce 
    the maximum allowable Reactor Building internal pressure specified in 
    TS 3.6.4 from 1.5 psig to 1.2 psig when the reactor is critical, revise 
    the corresponding Bases of TS 3.3 to indicate that the Reactor Building 
    sprays and coolers are designed to mitigate the containment temperature 
    response rather than containment pressure response to a loss-of-coolant 
    accident, and make other administrative changes. In addition, the lower 
    Reactor Building pressure limit (a vacuum of 5 inches of mercury (Hg)) 
    in Specification 3.6.4 would be changed to the corresponding value in 
    terms of psig to reflect the units displayed on the control room 
    instrumentation.
        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:
        No. The analysis of the post-LOCA [loss-of-coolant accident] 
    Reactor Building response to high-energy line breaks, using the new 
    methodology, uses assumptions different from the requirements 
    currently delineated in Technical Specifications. The new 
    assumptions used for initial Reactor Building pressure and Reactor 
    Building Spray system
    
    [[Page 55032]]
    
    actuation are 1.2 psig and 20 psig respectively. These values are 
    lower, and hence more conservative, than the values currently 
    specified in Technical Specifications.
        Since the new values for Reactor Building pressure and Reactor 
    Building Spray actuation are more conservative and the analysis 
    methodology has received approval from the NRC via [an] SER, this 
    change does not involve a significant increase in the probability or 
    consequences of an accident previously identified.
        (2) Create the possibility of a new or different kind of 
    accident from any kind of accident previously evaluated:
        No. The methodology for Reactor Building high energy line break 
    analysis is being revised. The revision of the method of analysis 
    does not alter the manner by which plant systems and components 
    function for accident mitigation.
        (3) Involve a significant reduction in a margin of safety.
        No. By letter dated March 15, 1995, the NRC stated that the new 
    analyses described in the topical report, DPC-NE-3003-P, expand the 
    scope of analyzed piping failures in containment for the Oconee 
    facilities. The NRC further stated that this new analysis method has 
    been used to reanalyze existing licensing basis pipe failure events 
    in containment, and to examine the potential effects of previously 
    unanalyzed assumptions and initial conditions which the NRC staff 
    finds to be consistent with current NRC staff acceptance criteria or 
    produce equally conservative results. In conclusion, the NRC 
    confirmed that this methodology, with appropriate adjustments to 
    reflect potential plant modifications, may be used by Duke Power to 
    perform future analyses in support of licensing applications related 
    to containment accident response. This proposed change to Technical 
    Specifications reflects the use of this new methodology. Based on 
    this new methodology, changes have been made to setpoint assumptions 
    for initial Reactor Building pressure and Reactor Building Spray 
    actuation. This proposed Technical Specification change reflects 
    those assumption changes. This methodology has been accepted by the 
    NRC. This proposed change to Technical Specifications does not 
    involve a significant reduction in the margin of safety.
        The NRC has reviewed the licensee's analysis and, based on 
    thisreview, 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 29691
        Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
    1200 17th Street, NW., Washington, DC 20036
        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: September 9, 1996
        Description of amendment request: The proposed amendment would 
    revise the Minimum Channels Operable requirement of Item 4.c (Steam 
    Line Isolation, Containment Pressure Intermediate -- High-High) of 
    Technical Specification (TS) Table 3.3-3 from 3 to 2. This proposed 
    change would make this Unit 1 TS consistent with the comparable Unit 2 
    TS.
        The proposed amendment would also revise the minimum charging pump 
    discharge pressure in TS 3.5.5 from 2311 psig to 2397 psig. This change 
    is required to ensure that safety analysis assumptions for safety 
    injection flow are met. Conforming changes would also be made to the 
    Bases for TS 3/4.5.5 to reflect the proposed changes to TS 3.5.5.
        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 amendment does not add or modify any existing plant 
    equipment. Since normal charging pump discharge pressure is greater 
    than or equal to approximately 2440 psig, no additional plant 
    configuration changes or modifications will be required to comply 
    with this revised charging pump discharge pressure value. The 
    proposed amendment does not change the design or function of the 
    containment pressure intermediate-high-high channels.
        The consequences of an accident previously evaluated are not 
    significantly increased. The ability of the containment pressure 
    intermediate-high-high function to initiate steam line isolation 
    will not be affected. Since steam line isolation will continue to 
    occur at the same required trip setpoint, the amount of mass and 
    energy released to containment along with the ability to maintain at 
    least one unfaulted steam generator (SG) as a heat sink for the 
    reactor remains unchanged. The amount of seal injection flow will 
    continue to be adequately limited to ensure sufficient flow to the 
    reactor core during accident conditions. The Bases changes are 
    editorial in nature and do not involve a change to probability or 
    consequences of an accident previously evaluated.
        Based on the above discussion, it is concluded that this change 
    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?
        The proposed amendment does not change the plant configuration 
    in a way which introduces a new potential hazard to the plant. Since 
    design requirements continue to be met and the integrity of the 
    reactor coolant system pressure boundary is not challenged, no new 
    failure mode has been created. As a result, an accident which is 
    different than already evaluated in the Updated Final Safety 
    Analysis Report will not be created due to this change.
        Therefore, the proposed change does 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?
        The margin of safety is not significantly reduced by this 
    proposed change. The trip setpoint for the containment pressure 
    intermediate-high-high function remains unchanged. With one channel 
    inoperable, the remaining two channels will continue to initiate the 
    protective function on a two-out-of-two logic. The action statement 
    limits this condition to 6 hours after which time the inoperable 
    channel must be placed in the trip condition. This action restores 
    the function to be able to meet single failure criteria on a one-
    out-of-two logic basis.
        The proposed revision to the charging pump discharge pressure 
    will not change the flow limit on seal injection. The specification 
    will continue to ensure that seal injection flow is limited. This 
    will ensure that sufficient flow to the reactor core is provided 
    during accident conditions.
        The proposed changes to the Bases for seal injection flow are 
    editorial in nature and do not affect the margin of safety.
        Therefore, this proposed 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: 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: John F. Stolz
    
    Entergy Gulf States Inc., Cajun Electric Power Cooperative, and 
    Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
    Unit 1, West Feliciana Parish, Louisiana
    
        Date of amendment request: August 29, 1996
        Description of amendment request: The proposed amendment would 
    revise the technical specifications (TSs) to reflect the elimination of 
    T-factor adjustments in the Average Power
    
    [[Page 55033]]
    
    Range Monitors (APRM) setpoints, a decrease in the calibration 
    frequency of the Local Power Range Monitors (LPMR), and an improvement 
    in the calculation of Reactivity Anomaly.
        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 request does not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        This change replaces the APRM setpoints T-factor limit with 
    power and flow-dependent minimum critical power ratio (MCPR) and 
    linear heat generation rate (LHGR) limits. These new power and flow-
    dependent thermal limits eliminate the need for manual setpoint 
    adjustment resulting from power peaking conditions. The new power 
    and flow-dependent thermal limits are automatically applied by 
    computer software during the calculation of the core thermal limits 
    and, therefore, do not require manual setpoint adjustments based on 
    the power peaking conditions in the reactor. Extensive transient 
    analyses at a variety of power and flow conditions have been 
    performed and were utilized to study the trend of transient severity 
    without the setpoints T-factor limit. A large data base was 
    established by analyzing limiting transients over a range of power 
    and flow conditions. The data base included evaluations 
    representative of a variety of plant configurations and parameters 
    such that the conclusions drawn from the studies would be applicable 
    to the broad range of boiling water reactors (BWRs). This data base 
    was utilized to develop plant specific operating limits (MCPR and 
    LHGR), which assures that margins to fuel safety limits are equal to 
    or larger than those currently in existence with the APRM setpoints 
    T-factor limit applied. Therefore, this change does not involve an 
    increase in the probability of any event previously evaluated.
        The consequences of an accident previously evaluated have not 
    been increased because, in all cases, the new power and flow-
    dependent thermal limits (MCPR and LHGR) assure that margins to fuel 
    safety limits are equal to or larger than those currently in 
    existence with the APRM setpoints T-factor limit applied. Protection 
    of other thermal limits for all previously analyzed events is 
    accomplished by specific limits that are independent of the APRM 
    setpoints T-factor. These are the power and flow-dependent MCPR 
    Operating Limits which provide protection from fuel dryout and the 
    rated maximum average planner linear heat generation rate (MAPLHGR) 
    limit which provides protection of the peak clad temperature for the 
    design basis accident-loss of coolant accident (DBA LOCA). 
    Therefore, the proposed change does not involve a significant 
    increase in the consequences of any event previously evaluated.
        No new equipment is introduced by the change in the local power 
    range monitor (LPRM) calibration frequency and, therefore, the 
    probability for an accident previously evaluated is unchanged. The 
    consequences of an accident can be affected by the thermal limits 
    prior to the accident but LPRM chamber and cycle exposure have no 
    significant effect on the calculated thermal limits. The thermal 
    limit calculation is not significantly effected because the LPRM 
    sensitivity versus exposure function is well defined. This allows 
    accurate LPRM end-of-life calculations so that detectors can be 
    replaced before their behavior significantly deteriorates. In the 
    event deterioration is noted late in the cycle for a few chambers, 
    they can be bypassed with no significant effect on uncertainties. 
    Also, the total nodal power uncertainty remains less than the 
    uncertainty assumed in the General Electric BWR Thermal Analysis 
    Basis (GETAB) safety limit. Therefore, the thermal limit calculation 
    is not affected by the LPRM calibration frequency and the 
    consequences of an accident previously evaluated are not changed.
        The change in the parameters used to measure reactivity for 
    calculation of the reactivity anomaly has no affect on either the 
    consequences or the probability of an accident previously evaluated 
    because the allowed reactivity anomaly criteria is unchanged. The 
    only change is the parameters used to measure reactivity.
        Therefore, the proposed elimination of the APRM setpoints T-
    factor maintains adequate off-rated MCPR and LHGR margin for all 
    operating conditions. Also, the change in the LPRM calibration 
    frequency continues to maintain the accuracy of the thermal limit 
    calculation. Therefore, the consequences of an accident previously 
    evaluated are not affected by this change. Finally, the change in 
    the parameters used to measure reactivity for calculation of the 
    reactivity anomaly has no affect on either the consequences nor the 
    probability of an accident previously evaluated. Since no new plant 
    equipment is introduced by any of the proposed changes, the 
    probability of accidents previously evaluated are not changed. 
    Therefore, none of the proposed changes involve an increase in the 
    probability or consequences of any event previously evaluated.
        2. The request does not create the possibility of occurrence of 
    a new or different kind of accident from any accident previously 
    evaluated.
        This change only replaces the APRM setpoints T-factor limit with 
    power and flow-dependent MCPR and LHGR limits, changes the LPRM 
    calibration frequency, and a change to the parameter(s) used to 
    measure reactivity. None of the proposed changes involve any new 
    modes of operation or any plant modifications. Therefore, the 
    proposed changes do not create the possibility of a new or different 
    type of accident from any accident previously analyzed.
        3. The request does not involve a significant reduction in a 
    margin of safety.
        The replacement of the APRM setpoints T-factor limit with power 
    and flow-dependent thermal limits has been confirmed to provide 
    adequate MCPR and LHGR protection at all reactor operation 
    conditions. Operation with higher peaking without APRM gains or flow 
    bias trip setpoints adjustment does not involve a reduction in a 
    margin of safety because the higher power peaking resulting from 
    elimination of the APRM setpoints T-factor has been analyzed to 
    assure that the margins to fuel safety limits are equal to or larger 
    than those currently in existence with the APRM setpoints T-factor 
    limit applied. Therefore, the replacement of the APRM setpoint T-
    factor with power and flow-dependent thermal limits does not involve 
    a reduction in the margin of safety.
        Protection of other thermal limits for all previously analyzed 
    events is accomplished by specific limits that are independent of 
    the APRM setpoint T-factor limit. These are the power and flow-
    dependent
        MCPR Operating Limits which provide protection from fuel dryout 
    and the rated MAPLHGR limit which provides protection of the peak 
    clad temperature for the DBA LOCA.
        The margin of safety can be affected by the thermal limits prior 
    to an accident but LPRM chamber exposure and cycle exposure have no 
    significant effect on the calculated thermal limits. The thermal 
    limit calculation is not significantly affected because the LPRM 
    sensitivity versus exposure function is well defined. This allows 
    accurate LPRM end of life calculations so that detectors can be 
    replaced before their behavior significantly deteriorates. In the 
    event deterioration is noted late in the cycle for a few chambers, 
    they can be bypassed with no significant effect on uncertainties. 
    Also, the total nodal power uncertainty remains less than the 
    uncertainty assumed in the GETAB safety limit. Therefore neither the 
    thermal limit calculation nor the margin of safety are affected by 
    the LPRM calibration.
        The change in the parameters used to measure reactivity for 
    calculation of the reactivity anomaly has no affect on the margin of 
    safety because the allowed reactivity anomaly criteria is unchanged. 
    The only change is the parameters used to measure reactivity.
        Neither the change to APRM setpoints T-factor nor the change to 
    the LPRM calibration frequency significantly effects the thermal 
    limits calculation, and, therefore, do not result in an increase in 
    core damage frequency. The change in the parameters used to measure 
    reactivity for calculation of the reactivity anomaly has no affect 
    on the core damage frequency because the allowable reactivity 
    anomaly criteria remains unchanged. Therefore, the proposed changes 
    do 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 Documenmt Room location: Government Documents 
    Department, Louisiana State University, Baton Rouge, LA 70803
        Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
    
    [[Page 55034]]
    
    1400 L Street, N.W., Washington, D.C. 20005
        NRC Project Director: William D. Beckner
    
    Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
    Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
    Unit 1, West Feliciana Parish, Louisiana
    
        Date of amendment request: August 29, 1996
        Description of amendment request: The proposed amendment would 
    provide a revision to the reactor pressure vessel (RPV) surveillance 
    capsule withdrawal schedule for the River Bend Station. The first 
    surveillance capsule would be withdrawn at 10.4 effective full power 
    years (EFPY) rather than at 6EFPY.
        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.
        Pressure-temperature (P-T) limits (RBS Technical Specifications 
    Figure 3.4.11-1) are imposed on the reactor coolant system to ensure 
    that adequate safety margins against nonductile or rapidly 
    propagating failure exist during normal operation, anticipated 
    operational occurrences, and system hydrostatic tests. The P-T 
    limits are related to the nil-ductility reference temperature, 
    RTNDT, as described in ASME Section III, Appendix G. Changes in 
    the fracture toughness properties of RPV beltline materials, 
    resulting from the neutron irradiation and the thermal environment, 
    are monitored by a surveillance program in compliance with the 
    requirements of 10CFR50, Appendix H. The effect of neutron fluence 
    on the shift in the nil-ductility reference temperature of pressure 
    vessel steel is predicted by methods give in Regulatory Guide 1.99, 
    Rev. 2.
        River Bend's current P-T limits were established based on 
    adjusted reference temperatures developed in accordance with the 
    procedures prescribed in Reg. Guide 1.99, Rev. 2, Regulatory 
    Position 1. Calculation of adjusted reference temperature by these 
    procedures includes a margin term to ensure conservative, upper-
    bound values are used for the calculation of the P-T limits. 
    Revision of the first capsule withdrawal schedule will not affect 
    the P-T limits because they will continue to be established in 
    accordance with Regulatory Position 1 (or other NRC-approved) 
    procedures. When permitted (two or more credible surveillance data 
    sets available), Regulatory Position 2 (or other NRC-approved) 
    methods for determining adjusted reference temperature will be 
    followed.
        This change is not related to any accidents previously 
    evaluated. The proposed change is a revision of the Withdrawal Time 
    for the first surveillance capsule as given in Technical 
    Requirements (TR) Table 3.4.11-1 from 6 EFPY to 10.4 EFPY. This 
    change will not affect P-T limits as given in RBS Technical 
    Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b. 
    This change will not affect any plant safety limits or limiting 
    conditions of operation. The proposed change will not affect reactor 
    pressure vessel performance as no physical changes are involved and 
    RBS vessel P-T limits will remain conservative in accordance with 
    Reg. Guide 1.99, Rev. 2 requirements. The proposed change will not 
    cause the reactor pressure vessel or interfacing systems to be 
    operated outside of their design or testing limits. Also, the 
    proposed change will not alter any assumptions previously made in 
    evaluating the radiological consequences of accidents. Therefore, 
    the probability or consequences of accidents previously evaluated 
    will not be increased by the proposed change.
        2. The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The proposed change is a revision of the Withdrawal Time in TR 
    Table 3.4.11 for the first RPV material surveillance capsule from 6 
    EFPY to 10.4 EFPY. This proposed change does not involve a 
    modification of the design of plant structures, systems, or 
    components. The proposed change will not impact the manner in which 
    the plant is operated as plant operating and testing procedures will 
    not be affected by the change. The proposed change will not degrade 
    the reliability of structures, systems or components important to 
    safety (ITS) as equipment protection features will not be deleted or 
    modified, equipment redundancy or independence will not be reduced, 
    supporting system performance will not be downgraded, the frequency 
    of operation of ITS equipment will not be increased, and increased 
    or more severe testing of ITS equipment will not be imposed. No new 
    accident types or failure modes will be introduced as a result of 
    the proposed change. Therefore, the proposed change does not create 
    the possibility of a new or different kind of accident from that 
    previously evaluated.
        3. The proposed changes do not involve a significant reduction 
    in a margin of safety.
        As stated in the River Bend SER, ``Appendices G and H of 10CFR50 
    describe the conditions that require pressure-temperature limits and 
    provide the general bases for these limits. These appendices 
    specifically require that pressure-temperature limits must provide 
    safety margins at least as great as those recommended in the ASME 
    Code, Section III, Appendix G. .... Until the results from the 
    reactor vessel surveillance program become available, the staff will 
    use RG 1.99, Revision 1 [now Revision 2] to predict the amount of 
    neutron irradiation damage. ... The use of operating limits based on 
    these criteria--as defined by applicable regulations, codes, and 
    standards--will provide reasonable assurance that nonductile or 
    rapidly propagating failure will not occur, and will constitute an 
    acceptable basis for satisfying the applicable requirements of GDC 
    31.''
        Bases for RBS Technical Specification 3/4/11 states: ``The P/T 
    limits are not derived from Design Basis Accident (DBA) analyses. 
    They are prescribed during normal operation to avoid encountering 
    pressure, temperature, and temperature rate of change conditions 
    that might cause undetected flaws to propagate and cause nonductile 
    failure of the RCPB [Reactor Coolant Pressure Boundary], a condition 
    that is unanalyzed. ... Since the P/T limits are not derived from 
    any DBA, there are no acceptance limits related to the P/T limits. 
    Rather, the P/T limits are acceptance limits themselves since they 
    preclude operation in an unanalyzed condition.''
        The proposed change will not affect any safety limits, limiting 
    safety system settings, or limiting conditions of operation. The 
    proposed change does not represent a change in initial conditions, 
    or in a system response time, or in any other parameter affecting 
    the course of an accident analysis supporting the Bases of any 
    Technical Specification. The proposed change does not involve 
    revision of the P-T limits but rather a revision of the Withdrawal 
    Time for the first surveillance capsule. The current P-T limits were 
    established based on adjusted reference temperatures for vessel 
    beltline materials calculated in accordance with Regulatory Position 
    1 of Reg. Guide 1.99, Rev. 2. P-T limits will continue to be revised 
    as necessary for changes in adjusted reference temperature due to 
    changes in fluence according to Regulatory Position 1 until two or 
    more credible surveillance data sets become available. When two or 
    more credible surveillance data sets become available, P-T limits 
    will be revised as prescribed by Regulatory Position 2 of Reg. Guide 
    1.99, Rev. 2 or other NRC-approved guidance. Therefore, the proposed 
    changes do not involve a significant reduction in any margins 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: Government Documents 
    Department, Louisiana State University, Baton Rouge, LA 70803
        Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
    1400 L Street, N.W., Washington, D.C. 20005
        NRC Project Director: William D. Beckner
    
    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
    Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
    
        Date of amendment request: September 23, 1996
        Description of amendment request: The proposed amendment would 
    revise the Crystal River Unit 3 (CR 3) technical specifications (TS) to 
    delete a note
    
    [[Page 55035]]
    
    associated with Surveillance Requirement (SR) 3.3.7.1 for the 
    Engineered Safeguard Actuation System (ESAS) Automatic Actuation Logic. 
    Applicable TS Bases will also be revised to reflect the proposed TS 
    change.
        SR 3.3.7.1 requires periodic testing of the ESAS automatic 
    actuation logic matrix to demonstrate that the required logic 
    combinations are operable. When the ESAS automatic actuation logic is 
    placed in an inoperable status solely for performing of this 
    surveillance, the note associated with the SR 3.3.7.1 provides relief 
    in that it allows not entering into applicable Conditions and Required 
    Actions for up to 8 hours, provided the associated engineering 
    safeguards (ES) function is maintained. The licensee has determined 
    that because of the CR 3 design of the ESAS System and the way the test 
    is performed, maintenance of the ``associated ES function'' is not 
    possible. Thus, the note does not provide the relief intended and 
    therefore, the licensee proposes to delete the note. During the 
    performance of the ESAS test and bypassing the associated ES function, 
    the licensee proposes to enter into applicable TS Conditions.
        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 will not significantly increase the 
    probability or consequences of an accident previously evaluated 
    because unavailability of equipment is recognized in the design of 
    the plant and in the Technical Specifications. The probability and 
    consequences of accidents previously evaluated are bounded by the 
    evaluations done for the allowed outage time of the associated 
    functions.
        2. The proposed change will not create the possibility of a new 
    or different kind of accident from any accident previously evaluated 
    because the bypassing of ES functions for testing purposes does not 
    place the plant in a configuration which would allow the possibility 
    of a new or different kind or accident to be created.
        3. The proposed change will not involve a significant reduction 
    to the margin of safety because deleting the NOTE does not effect 
    the way the test is performed. The test is required by the Technical 
    Specifications and will still be performed in the same manner. Thus, 
    there is no change in the unavailability of the system as a result 
    of this change and the margin of safety is not reduced.
        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: Coastal Region Library, 8619 
    W. Crystal Street, Crystal River, Florida 32629
        Attorney for licensee: A. H. Stephens, General Counsel, Florida 
    Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida 
    33733
        NRC Project Director: Frederick J. Hebdon
    
    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
    Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
    
        Date of amendment request: September 27, 1996
        Description of amendment request: The proposed amendment would 
    revise the Crystal River 3 (CR3) post-accident monitoring (PAM) 
    instrumentation technical specification (TS). Specifically, the 
    following TS changes are proposed:
        A. Table 3.3.17-1, Function 8: The descriptor is changed from 
    ``Containment Pressure (Narrow Range)'' to ``Containment Pressure 
    (Expected Post-Accident Range).''
        B. Table 3.3.17-1, Function 18: The required channels for Core Exit 
    Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core 
    quadrant.''
        C. Table 3.3.17-1: A new Function 20 is added and designated as 
    ``Low Pressure Injection Flow.''
        D. Table 3.3.17-1: A new Function 21 is added and designated as 
    ``Degrees of Subcooling.''
        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 (the letters A, B, C and D correspond to the proposed TS 
    changes), which is presented below:
        1. The proposed changes will not significantly increase the 
    probability or consequences of an accident previously evaluated 
    because:
        A/B. The changes in containment pressure and core exit 
    thermocouple nomenclature do not reflect any physical changes to the 
    facility.
        C/D.The addition of low pressure injection flow and degrees of 
    subcooling to the Post-Accident Monitoring Instrumentation LCO is 
    being done to comply with a commitment made during the technical 
    specification improvement program to include in the technical 
    specifications, that instrumentation which monitors variables 
    classified as Type A in accordance with Regulatory Guide 1.97. These 
    two variables have recently been re-classified as Type A. The 
    associated instruments are used after an accident occurs to prompt 
    the operators to take certain mitigative actions. Therefore, the 
    probability of an accident occurring is unaffected. As part of the 
    re-classification of these variables to Type A, the associated 
    monitoring instrumentation will be under more strict surveillance 
    and control, which provides additional assurance that the prescribed 
    manual operator actions will be implemented when necessary. This, in 
    turn, assures the previously evaluated accident consequences remain 
    valid.
        2. The proposed changes will not create the possibility of a new 
    or different kind of accident from any accident previously evaluated 
    because:
        A/B. The changes in containment pressure and core exit 
    thermocouple nomenclature do not reflect any physical changes to the 
    facility. The changes provide clarification for the instruments 
    which are required to comply with the LCO.
        C/D. The addition of low pressure injection flow and degrees of 
    subcooling to the Post-Accident Monitoring instrumentation LCO is 
    being done to comply with a commitment made during the technical 
    specification improvement program to include in the technical 
    specifications, that instrumentation which monitors variables 
    classified as Type A in accordance with Regulatory Guide 1.97. These 
    two variables have been re-classified as Type A. The associated 
    instruments are used after an accident occurs to prompt the 
    operators to take certain mitigative actions. Since the 
    instrumentation is used only post-accident, these changes do not 
    create the possibility of a new or different kind of accident.
        3. The proposed change will not involve a significant reduction 
    to the margin of safety because:
        A/B. The changes in containment pressure and core exit 
    thermocouple nomenclature have no affect on the margin of safety. 
    The changes provide clarification of the technical specifications. 
    This reduces the potential for confusion regarding this 
    instrumentation.
        C/D. The addition of low pressure injection flow and degrees of 
    subcooling to the post-accident monitoring instrumentation table 
    adds controls on the OPERABILITY of post-accident monitoring 
    instrumentation providing greater assurance it will be available 
    should an accident occur.
        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: Coastal Region Library, 8619 
    W. Crystal Street, Crystal River, Florida 32629
        Attorney for licensee: A. H. Stephens, General Counsel, Florida 
    Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida 
    33733
        NRC Project Director: Frederick J. Hebdon
    
    [[Page 55036]]
    
    Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
    Millstone Nuclear Power Station, Unit 1, New London County, 
    Connecticut
    
        Date of amendment request: September 5, 1996
        Description of amendment request: The proposed change deletes 
    License Condition 2.C.5, Integrated Implementation Schedule.
        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 attached 
    proposed change and has concluded that it does not involve a 
    significant hazards consideration (SHC). The basis for this is that 
    the three criteria of 10CFR50.92(c) are not compromised. The 
    proposed change does not involve an SHC because the change would 
    not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        Operation of the facility in accordance with the proposed change 
    would result in a change in an administrative process for 
    prioritizing and scheduling projects and engineering evaluations. 
    With the limited number of NRC required projects remaining to be 
    implemented, the IIS [Integrated Implementation Schedule] is no 
    longer required to schedule resources for the remaining topics. 
    Since this license condition only involves an administrative 
    process, it does not directly affect the design or operation of the 
    plant. Therefore, no accident analyses are affected by the change, 
    and the change does not increase the probability or consequences of 
    any previously evaluated accident.
        2. Create the possibility of a new or different kind of accident 
    from any previously evaluated.
        The proposed license modification removes a requirement relating 
    to the scheduling of modifications and engineering evaluations. 
    Because the license condition addresses only an administrative 
    scheduling mechanism, it does not affect directly the design or 
    operation of the plant. Therefore, the proposed change does not 
    create a different kind of accident from those previously analyzed.
        3. Involve a significant reduction in a margin of safety.
        The proposed license modification removes a requirement relating 
    to the scheduling of modifications and engineering evaluations. The 
    original purpose of the IIS and the ISAP [Integrated Safety 
    Assessment Program] was to prioritize and schedule modifications and 
    engineering evaluations in a manner that was agreed upon by both 
    NNECO and the NRC. These programs were especially important to 
    Millstone Unit No. 1 for priorization of topics associated with the 
    SEP [Systematic Evaluation Program] and the TMI [Three Mile Island] 
    Action Plan. This program is considered to be no longer necessary. 
    Modifications and engineering evaluations will be scheduled and 
    prioritized using other methodologies. Since this change involves an 
    administrative process only, there is no direct impact on the design 
    or operation of the plant, and therefore, no significant 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: Learning Resources Center, 
    Three Rivers Community-Technical College, 574 New London Turnpike, 
    Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
    Rope Ferry Road, Waterford, CT 06385
        Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
    Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
    CT 06141-0270
        NRC Project Director: Phillip F. McKee
    
    PECO Energy Company, Public Service Electric and Gas Company, 
    Delmarva Power and Light Company, and Atlantic City Electric 
    Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
    Station, Units Nos. 2 and 3, York County, Pennsylvania
    
        Date of application for amendments: August 27, 1996
        Description of amendment request: The proposed amendment revises 
    the required value of control rod drive (CRD) system pressure in 
    technical specification (TS) 3.10.8, ``Shutdown Margin (SDM) Test-
    Refueling.''
        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 any accident previously 
    evaluated.
        The proposed changes are purely administrative and do not 
    involve any physical changes to plant SSC [systems, structures and 
    components]. The change in the minimum CRD charging water header 
    pressure from 955 psig to 940 psig was previously approved in TS 
    Amendments Nos. 211 and 216 for PBAPS [Peach Bottom Atomic Power 
    Station], Units 2 and 3. TS Change Request 95-12 was incomplete by 
    inadvertently failing to identify the need to change requirement (f) 
    of LCO [Limiting Condition for Operation] 3.10.8. Therefore, the 
    proposed changes will not increase the probability of occurrence or 
    the consequences of an accident previously evaluated in the SAR 
    [safety analysis report].
        2) The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The proposed changes are purely administrative and do not 
    involve any physical changes to plant SSC. The proposed changes do 
    not allow plant operation in any mode that is not already evaluated 
    in the SAR. Therefore, the possibility of a different type of 
    accident than previously evaluated in the SAR is not created.
        3) The proposed changes do not result in a significant reduction 
    in the margin of safety.
        The proposed changes are purely administrative and have no 
    impact on any safety analysis assumptions or margins of safety. A 
    change to SR 3.10.8.6 was approved by the NRC by TS Amendment Nos. 
    211 and 216. LCO 3.10.8 requirement (f) should have been changed at 
    the same time to reflect a minimum CRD charging water pressure of 
    940 psig. Changing LCO 3.10.8 requirement (f) to reflect TS 
    Amendment Nos. 211 and 216 is purely administrative, and therefore, 
    does not involve a 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: Government Publications 
    Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
    Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
    PA 17105
        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: John F. Stolz
    
    Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
    Limerick Generating Station, Units 1 and 2, Montgomery County, 
    Pennsylvania
    
        Date of amendment request: May 20, 1996
        Description of amendment request: The proposed Technical 
    Specifications (TS) changes would revise TS Sections 3/4.4.9.2, 3/
    4.9.11.1, 3/4.9.11.2, and the associated TS Bases 3/4.4.9 and 3/4.9.11, 
    to more clearly describe that the Residual Heat Removal (RHR) system 
    Shutdown Cooling mode of operation consists of four (4) ``subsystems.'' 
    These TS sections pertain to plant operations during Operational 
    Conditions (OPCONs) 4, ``Cold Shutdown'' and 5, ``Refueling.'' In 
    addition, the proposed TS change would make administrative changes to 
    TS Section 3/4.4.9.1 to
    
    [[Page 55037]]
    
    ensure consistency in terminology regarding the description of Shutdown 
    Cooling ``subsystems.'' The proposed TS changes are consistent with the 
    guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1, 
    ``Standard Technical Specifications General Electric Plants, BWR/4,'' 
    dated April 1995) which indicates that the RHR Shutdown Cooling mode of 
    operation is comprised of two (2) loops and four (4) subsystems (i.e., 
    two (2) subsystems per loop).
        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 TS changes do not involve any physical changes to 
    plant structures systems, or components. The RHR [Residual Heat 
    Removal] Shutdown Cooling mode of operation is manually controlled 
    and is not required for accident mitigation. The RHR system will 
    continue to function as designed in all modes of operation. The 
    consequences of equipment malfunction are not changed from those in 
    existing analyses, with no increase in onsite or offsite 
    radiological effects. The RHR system will continue to function as 
    designed to mitigate the consequences of an accident and resultant 
    onsite and offsite radiological effects remain as previously 
    evaluated. The proposed TS changes will revise the TS to more 
    clearly describe the RHR system configuration in OPCONs 4 and 5. The 
    proposed changes are consistent with the guidance stipulated in 
    NUREG-1433, Revision 1.
        The four (4) ``subsystem'' Shutdown Cooling designation permits 
    operability of only one (1) RHR heat exchanger for Shutdown Cooling 
    service in Operational Conditions (OPCONs) 4 and 5, as long as both 
    associated RHR pumps are operable and alignable for Shutdown 
    Cooling. TS requirements for RHR Shutdown Cooling operation in Hot 
    Shutdown, Suppression Pool Spray, and Suppression Pool Cooling 
    continue to require two (2) independent loops to be operable in 
    OPCONs 1, 2, and 3*, meaning both RHR heat exchangers will still be 
    required to be operable throughout OPCON 3.
        The four (4) ``subsystem'' Shutdown Cooling designation has no 
    effect on the required operability of the Residual Heat Removal 
    Service Water (RHRSW) system. As required by TS Section 3.7.1.1, the 
    RHRSW subsystem(s) associated with the required operable RHR heat 
    exchanger(s) will continue to remain operable. Each operable RHRSW 
    subsystem consists of two (2) operable pumps and the required 
    operable flowpath to provide decay heat removal via the associated 
    RHR heat exchanger.
        The RHRSW system piping is designed, fabricated, inspected, and 
    tested in accordance with the requirements of ASME [American Society 
    of Mechanical Engineers], Section III Class 3, and each RHRSW 
    subsystem is single active failure proof in that the failure of a 
    motor-operated valve, diesel generator, or pump does not prevent the 
    system from performing its safety function.
        The required availability of four (4) loops of the Low Pressure 
    Coolant Injection (LPCI) mode of RHR during OPCONs 1, 2, and 3 as 
    required by TS Section 3.5.1 is not impacted by the four (4) 
    ``subsystem'' Shutdown Cooling designation. No change to any RHR 
    system instrumentation logic, required Emergency Core Cooling System 
    (ECCS) availability, or method of operation is involved.
        NUREG-1433, Revision 1, also re-affirms that each Shutdown 
    Cooling ``subsystem'' is considered operable if it can be manually 
    aligned, remotely or locally, in the shutdown cooling mode for 
    removal of decay heat. Thus, a LPCI-dedicated pump can be aligned 
    for LPCI automatic initiation, yet still be considered part of an 
    operable shutdown cooling subsystem as long as it can be re-aligned 
    for Shutdown Cooling.
        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 changes do not involve any physical changes to 
    plant structures, systems, or components. The RHR system will 
    continue to function as designed in all modes of operation. No new 
    accident type is created as a result of the proposed changes. No new 
    failure mode for any equipment is created. The changes are 
    consistent with the guidance provided in NUREG-1433, Revision 1, 
    pertaining to RHR Shutdown Cooling operation in OPCONs 4 and 5.
        The four (4) ``subsystem'' Shutdown Cooling designation has no 
    effect on the required operability of the RHRSW system. The RHRSW 
    subsystem(s) associated with the required operable RHR heat 
    exchanger(s) will continue to remain operable as required by TS 
    Section 3.7.1.1. Each operable RHRSW subsystem consists of two (2) 
    operable pumps and the required operable flowpath to provide decay 
    heat removal via the associated RHR heat exchanger.
        The RHRSW system piping is designed, fabricated, inspected, and 
    tested in accordance with the requirements of ASME, Section III, 
    Class 3, and each RHRSW subsystem is single active failure proof in 
    that the failure of a motor-operated valve, diesel generator, or 
    pump does not prevent the system from performing its safety 
    function.
        The required availability of four (4) loops of the LPCI mode of 
    RHR during OPCONs 1, 2, and 3 as required by TS Section 3.5.1 and 
    3.5.2 is not impacted by the four (4) ``subsystem'' Shutdown Cooling 
    designation. No change to any RHR system instrumentation logic, 
    required ECCS availability, or method of operation is involved.
        NUREG-1433, Revision 1, also re-affirms that each Shutdown 
    Cooling ``subsystem'' is considered operable if it can be manually 
    aligned, remotely or locally, in the Shutdown Cooling mode for 
    removal of decay heat. Thus, a LPCI-dedicated pump can aligned be 
    [sic] [be aligned] for automatic LPCI initiation, yet still be 
    considered part of an operable shutdown cooling subsystem as long as 
    it can be re-aligned for Shutdown Cooling.
        Therefore, the proposed TS changes do not create the possibility 
    of a new or different kind of accident from any accident previously 
    evaluated.
        3. The proposed TS changes do not involve a significant 
    reduction in a margin of safety.
        Although the Bases for TS Sections 3/4.4.9.2, 3/4.9.11.1, and 3/
    4.9.11.2 are being revised in support of this proposed TS change, 
    the changes only involve providing clarification regarding the 
    designation of the RHR Shutdown Cooling operation configuration in 
    OPCONs 4 and 5. The proposed TS changes do not involve any physical 
    changes to plant structures, systems, or components. The RHR system 
    will continue to function as designed in all modes of operation. The 
    consequences of equipment malfunction are not changed from those in 
    existing analyses, with no increase in onsite or offsite 
    radiological effects. The RHR system will continue to function as 
    designed to mitigate the consequences of an accident and resultant 
    onsite and offsite radiological effects remain as previously 
    evaluated. The proposed changes are consistent with the guidance 
    stipulated in NUREG-1433, Revision 1.
        The four (4) ``subsystem'' Shutdown Cooling designation has no 
    effect on the required operability of the RHRSW system. As required 
    by TS 3.7.1.1, the RHRSW subsystem(s) associated with the required 
    operable RHR heat exchanger(s) will continue to remain operable. 
    Each operable RHRSW subsystem consists of two (2) operable pumps and 
    the required operable flowpath to provide decay heat removal via the 
    associated RHR heat exchanger.
        The RHRSW system piping is designed, fabricated, inspected, and 
    tested in accordance with the requirements of ASME, Section III, 
    Class 3, and each RHRSW subsystem is single active failure proof in 
    that the failure of a motor-operated valve, diesel generator, or 
    pump does not prevent the system from performing its safety 
    function. (In the same manner that manual action may be required for 
    RHR system alignment in OPCONs 4 and 5 with one (1) RHR heat 
    exchanger operable, a failure of the motor-operated RHRSW inlet or 
    outlet heat exchanger isolation valves may require manual 
    positioning for the required alignment.)
        The required availability of four (4) loops of the LPCI mode of 
    RHR during OPCONs 1, 2, and 3* as required by TS Section 3.5.1 is 
    not affected by the four (4) ``subsystem'' Shutdown Cooling 
    configuration. No change to any RHR system instrumentation logic, 
    required ECCS availability, or method of operation is involved.
        NUREG-1433, Revision 1, also re-affirms that each Shutdown 
    Cooling ``subsystem'' is
    
    [[Page 55038]]
    
    considered operable if it can be manually aligned, remotely or 
    locally, in the Shutdown Cooling mode for removal of decay heat. 
    Thus, a LPCI-dedicated pump can be aligned for LPCI automatic 
    initiation, yet still be considered part of an operable Shutdown 
    Cooling ``subsystem'' as long as it can be re-aligned for Shutdown 
    Cooling.
        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.
        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, Philadelphia Electric Company, 2301 Market Street, 
    Philadelphia, PA 19101
        NRC Project Director: John F. Stolz
    
    Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
    Limerick Generating Station, Units 1 and 2, Montgomery County, 
    Pennsylvania
    
        Date of amendment request: June 28, 1996
        Description of amendment request: The proposed Technical 
    Specifications (TS) changes would incorporate performance-based 
    testing, in accordance with 10 CFR Part 50, Appendix J, ``Primary 
    Reactor Containment Leakage Testing For Water-Cooled Power Reactors,'' 
    Option B. This option allows utilities to extend the frequencies of the 
    Type A Containment (ILRT) Leak Rate Test and Type B and C Local Leak 
    Rate Tests (LLRTs) based on the performance and design of the 
    containment and components.
        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 TS changes do not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        Incorporation of the new 10 CFR 50, Appendix J, Option B at LGS, 
    Units 1 and 2 does not increase the probability of occurrence of an 
    accident previously evaluated. The containment structure including 
    its isolation capability is not an accident initiator.
        These changes do not involve any changes to the containment 
    structure, system or components which could increase the probability 
    of occurrence of an accident previously evaluated or act as a new 
    accident initiator. Implementation of the proposed changes will 
    affect the manner in which these structures, systems, or components 
    (SSCs) are tested; however, the new testing schedule is not an 
    initiator of any analyzed event. No equipment changes are involved 
    with adoption of Option B; therefore, performance-based test 
    intervals for Type A, B, and C tests do not increase the probability 
    of occurrence of a malfunction of equipment important to safety 
    previously evaluated. No physical changes are being made to the 
    plant, nor are there any changes being made in the operation of the 
    plant as the result of increasing the test intervals. Additionally, 
    the proposed TS changes will not alter the operation of equipment 
    available for the mitigation of accidents or transients, therefore, 
    this change will not result in any significant increase to onsite or 
    offsite dose previously evaluated. The potential for time-based and 
    activity-based failure mechanisms which could lead to excessive 
    containment leakage has been determined to be minimal. Performance-
    based test intervals for Type A, B, and C tests will not alter any 
    safety limits which ensure the integrity of fuel barriers, and will 
    not increase the primary containment leakage limits.
        Performance-based test intervals for Type A, B, and C leak tests 
    do not increase the consequences of an accident previously 
    evaluated. NUREG-1493 concluded that reducing the frequency of Type 
    A tests from the current three per ten years to one per ten years 
    was found to lead to an imperceptible increase in risk. NUREG-1493 
    includes the results of a sensitivity study performed to explore the 
    risk impact of several alternative leak rate test schedules. The 
    estimated increase in population exposure risk ranged from 0.02% to 
    0.14%. The risk impact was determined to be very small since Type B 
    and C testing (local leak rate tests) detect a very large percentage 
    of overall containment leakages. The percentage of leakages detected 
    by Type A tests is very small. Past test results experienced at 
    Limerick Units 1 and 2 concur with these determinations. NUREG-1493 
    also concluded that the overall unit risk is not very sensitive to 
    changes in containment leakage rates. Given the insensitivity of 
    risk to containment leak rates and the small fraction of leak paths 
    detected solely by the Type A tests, increasing the interval between 
    Type A tests is possible with minimal impact on public risk.
        NUREG-1493 also concluded that, based on a model of component 
    failure with time, the performance-based alternatives to current, 
    local-leakage testing requirements are feasible without significant 
    risk impact. The LGS design and past performance is bounded by the 
    NUREG study. The NUREG model indicated that the number of components 
    tested could be reduced by about 60% with less than a three-fold 
    increase in the incremental risk due to containment leakage. Since 
    under existing requirements, leakage contributes less than 0.1 
    percent of overall accident risk, the overall impact is very small.
        Therefore, the proposed TS changes will not involve a 
    significant 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.
        Performance-based test intervals for Type A, B, and C leak tests 
    do not introduce a new or different type of accident or create the 
    possibility of a different type of malfunction of equipment 
    important to safety than previously evaluated. No physical changes 
    are being made to the plant, nor are there any changes being made in 
    the operation of the plant as the result of increasing the test 
    intervals. No new failure modes of plant equipment previously 
    evaluated will be introduced. Additionally, the TS changes will not 
    alter the operation of equipment available for the mitigation of 
    accidents or transients. The safety function of the primary 
    containment will be retained since the containment will continue to 
    provide an essentially leak tight barrier against the uncontrolled 
    release of radioactivity to the environment for postulated accidents 
    previously evaluated.
        Therefore, the proposed TS changes will not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        3. The proposed TS changes do not involve a significant 
    reduction in a margin of safety.
        The margin of safety is not reduced as a result of adopting 10 
    CFR 50, Appendix J, Option B. The effect of increasing containment 
    leakage rate testing intervals was evaluated in NUREG-1493 using 
    historical industry leakage rate testing results. Performance 
    history at LGS is consistent with the conclusions reached in NUREG-
    1493 and NEI 94-01. The results of the NUREG evaluation conclude 
    that the increased safety risk corresponding to the extended test 
    intervals is small (less than 0.1% of total risk). The revised TS 
    will continue to maintain the allowable leakage rate for the Type A 
    tests. In addition, the requirement to perform a periodic general 
    visual inspection of the primary containment has been maintained at 
    the original interval of three times in 10 years as part of the 
    performance-based leakage rate testing program.
        The risk of a non-detectable increase of primary containment 
    leakage is considered to be negligible due to the conclusion that 10 
    CFR 50, Appendix J, Type B and C testing program will continue to be 
    conducted between Type A tests. A review of previous LGS Type A test 
    results has concluded that the only failure mechanisms are activity-
    based. There is no indication of time-based failures that would not 
    be identified during the performance of Type B and C tests. 
    Therefore, we have concluded that the proposed adoption of the 
    Option B intervals would not result in a non-detectable primary 
    containment leakage rate in excess of the allowable value (i.e., 
    0.5% wt/day) established by the LGS TS.
        The proposed TS will continue to maintain the allowable leakage 
    rate for the combined Type B and C tests. As supported by the 
    findings of NUREG-1493, the percentage of leakages detected by Type 
    A tests is small (as
    
    [[Page 55039]]
    
    stated above) and Type B and C leakage tests are capable of 
    detecting more than 97% of containment leakages and virtually all 
    such leakages are identified by local leak rate tests of containment 
    isolation valves. The Type B and C test intervals will be 
    established through the PCLRTP for each component based on design 
    and previous LGS test performance history.
        Therefore, the proposed TS changes do not involve a 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: Pottstown Public Library, 500 
    High Street, Pottstown, PA 19464
        Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
    General Counsel, Philadelphia Electric Company, 2301 Market Street, 
    Philadelphia, PA 19101
        NRC Project Director: John F. Stolz
    
    Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
    311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
    County, New Jersey
    
        Date of amendment request: September 25, 1996
        Description of amendment request: The amendments would relocate to 
    the Salem Updated Final Safety Analysis Report the list of containment 
    isolation valves that are currently located in Table 3.6-1 of Technical 
    Specification 3.6.3. In addition, references to the table in 
    specifications 1.7, 3.6.1, and 3.6.3 are being updated.
        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 consequence of an accident previously 
    evaluated.
        The proposed changes simplify the TS, meet the regulatory 
    requirements for control of containment isolation, and are 
    consistent with the guidance provided in Generic Letter (GL) 91-08, 
    ``Removal of Component Lists from Technical Specifications.'' The 
    procedural details of TS Table 3.6-1 have not been changed, only 
    relocated to a different controlling document, the Salem Update 
    [sic] [Updated] Final Safety Analysis Report (UFSAR). The proposed 
    changes are administrative in nature, should result in improved 
    administrative practices, and do not affect plant operations.
        The probability of occurrence of a previously evaluated accident 
    is not increased because this change does not introduce any new 
    potential accident initiating conditions. The consequences of an 
    accident previously evaluated is not increased because the ability 
    of containment to restrict the release of any fission product 
    radioactivity to the environment will not be degraded by this 
    change.
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The proposed changes are administrative in nature, do not result 
    in a physical alterations or changes to the operation of the plant, 
    and cause no change in the method by which any safety-related system 
    performs its functions. Therefore, this proposed change will 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 administrative change to relocate TS Table 3.6-1 to the 
    UFSAR does not alter the basic regulatory requirements for 
    containment isolation and will not adversely affect the containment 
    isolation capability for credible accident scenarios. Adequate 
    control of the content of the relocated table is assured by the 
    10CFR50.59 review process.
        The proposed relocation of TS Table 3.6-1 does not alter the 
    requirements for CIV operability currently in the TS. the Limiting 
    Condition for Operation and the Surveillance Requirements would be 
    retained in the revised TS. Therefore, the proposed changes will not 
    affect the meaning, application, and function of the current TS 
    requirements for the CIVs.
        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: Salem Free Public library, 112 
    West Broadway, Salem, NJ 08079
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
    Strawn, 1400 L Street, NW, Washington, DC 20005-3502
        NRC Project Director: John F. Stolz
    
    Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
    311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
    County, New Jersey
    
        Date of amendment request: September 25, 1996
        Description of amendment request: The amendments would change 
    Technical Specification 3/4.8.1, ``Electrical Power Systems,'' to 
    revise the Emergency Diesel Generator (EDG) voltage and frequency 
    limits as a result of updated EDG load calculations and to eliminate 
    ambiguity in the testing methodology for EDG start timing.
        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.
        Since no change is being made to the offsite power supplies, or 
    to any system or component that interfaces with the offsite power 
    supplies, there is no change in the probability of a Loss of Offsite 
    Power Accident.
        The proposed changes provide the necessary conservatism for 
    voltage and frequency to ensure the EDGs are not run in an 
    overloaded condition and that driven equipment is not damaged during 
    steady state operation following a Loss of Offsite Power coincident 
    with a Loss of Coolant Accident. Since the narrower band of voltage 
    and frequency for the isochronous mode continues to ensure proper 
    steady state operation of the EDG and associated driven equipment, 
    there is no change in the consequences of an accident previously 
    evaluated.
        Based on the above, the proposed amendment does 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.
        The proposed amendment does not result in any design or physical 
    configuration changes to the EDGs. Proposed changes made to the 
    testing parameters and testing methodology will not cause a new or 
    different accident since the EDGs are used for accident mitigation 
    and no new failure modes are being introduced. Therefore, the 
    proposed amendment will 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 amendment provides further conservatism to the 
    voltage and frequency band currently specified in the TSs. The 
    proposed voltage and frequency changes ensure the EDG will not be 
    overloaded from an over-frequency condition and driven equipment 
    will not be damaged from an over-voltage condition.
        The control system is set to control the EDG voltage within the 
    bands specified in the requested changes. The changes are consistent 
    with current calculations and within the capability of the controls. 
    Since the narrower band of voltage and frequency for the isochronous 
    mode is bounded by the existing TS, there is no change in the margin 
    of safety. The increased band for droop mode will ensure the EDG is 
    capable of operating in accordance with normal offsite power 
    parameters and does not 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
    
    [[Page 55040]]
    
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Salem Free Public library, 112 
    West Broadway, Salem, NJ 08079
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
    Strawn, 1400 L Street, NW, Washington, DC 20005-3502
        NRC Project Director: John F. Stolz
    
    Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
    311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
    County, New Jersey
    
        Date of amendment request: October 1, 1996
        Description of amendment request: The proposed amendments would 
    change Technical Specifications (TSs) 3/4.7.1.5, ``Main Steam Line 
    Isolation Valves (MSIVs),'' and 3/4.3.2, ``Engineered Safety Feature 
    Actuation System Instrumentation.'' These changes are needed to 
    accommodate entry into Modes 3 and 2 prior to performing MSIV closure 
    time testing in Mode 2. The proposed amendments would also allow for 
    the repair and testing of inoperable MSIVs in certain operating Modes, 
    and would change the low steam line pressure trip setpoint value for 
    safety injection to make it consistent with the previously approved 
    value for steam line isolation.
        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 isolation capability of the MSIVs and the protective 
    functions of the low steam line pressure channels are necessary for 
    accident mitigation and do not impact the probability of an 
    accident. MSIV testing in the higher modes is necessary to obtain 
    conditions which enable testing of the MSIVs. These conditions are 
    consistent with the current accident analyses for main steam line 
    breaks and secondary system depressurization. Failure of a MSIV, 
    which could be encountered during testing, is accounted for in the 
    accident analyses.
        Provisions for entering Mode 2 within six hours with an 
    inoperable MSIV allows operators to remove the plant from power 
    generation in a more controlled manner without challenging plant 
    safety systems and is consistent with other plant shutdown TS (i.e., 
    TS 3.0.3). The additional six hours to Hot Shutdown, should MSIV 
    closure be infeasible, does not result in a significant increase in 
    the probability or consequence of an accident since this is a very 
    small incremental time addition. The values for the low steam line 
    pressure safety injection are higher and are bounded by the present 
    accident analysis. The elimination of the obsolete stroke time of 
    eight seconds is editorial in nature. As a result, the changes 
    proposed do not involve a significant increase in the probability or 
    consequence 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 changes do not involve any modifications to 
    existing plant equipment, do not alter the function of any plant 
    systems, do not introduce any new operating configurations or new 
    modes of plant operation, nor change the safety analyses. The 
    proposed changes will, therefore, not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        3. The proposed change does not involve a significant reduction 
    in a margin of safety.
        MSIV testing in Mode 2 is within the currently analyzed plant 
    operation as discussed in the Updated Final Safety Analysis Report 
    (UFSAR) Sections 10.3 and 15.4. These UFSAR sections address 
    performance of the TS surveillance test at or near 1000 psig Steam 
    Generator pressure to assure main steam isolation occurs within the 
    accident conditions, where Steam Generator pressure may be lower 
    during Mode 1 operation. The test methodology demonstrating MSIV 
    operability is consistent with the accident analysis.
        Operation in Modes 2 and 3 with one or more isolation valve 
    inoperable and in the closed position does not impact the margin of 
    safety since the valves are already performing the safety function.
        The protective functions that occur as a result of the low steam 
    line pressure initiating signal remain bounded by the values assumed 
    in the safety analyses. That is, the protective functions that occur 
    as a result of this initiating signal already assume a setpoint that 
    is conservative for the revised value. The change to the setpoint 
    eliminates conflicting information in the TS.
        Therefore, the proposed changes 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: Salem Free Public library, 112 
    West Broadway, Salem, NJ 08079
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
    Strawn, 1400 L Street, NW, Washington, DC 20005-3502
        NRC Project Director: John F. Stolz
    
    Public Service Electric & Gas Company, Docket No. 50-311, Salem 
    Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
    
        Date of amendment request: September 20, 1996, as supplemented 
    September 30, 1996
        Description of amendment request: The proposed amendment would 
    change Technical Specification 4.7.7.b.4 to indicate that the specified 
    flowrate for the Auxiliary Building Exhaust Air Filtration System 
    applies only to system testing.
        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 accident considered in this proposed change is the Loss of 
    Coolant Accident (LOCA) as described in Section 15.4 of the UFSAR 
    [Updated Final Safety Analysis Report]. The assumption is that: 
    ``The Auxiliary Building Ventilation System will discharge the vapor 
    (from recirculation liquid leakage) to the atmosphere through 
    charcoal filters which have an efficiency of 90 percent.'' As such 
    the system acts to limit the total offsite and control room 
    radiation doses following a LOCA.
        The Auxiliary Building Ventilation System [ABVS] is designed to 
    maintain the Auxiliary Building at a negative pressure with respect 
    to the atmosphere during normal and emergency operation. Filtration 
    of radio-iodines is accomplished by administratively aligning the 
    ECCS [emergency core cooling system] equipment areas exhaust flows 
    to the standby charcoal adsorber bed if required. The ABVS has no 
    direct impact on reactor operation or on any system connected to the 
    Reactor Coolant Pressure Boundary.
        The emergency operation of the Auxiliary Building Ventilation 
    System is not affected by the proposed changes. The acceptance 
    criteria for system performance are not modified by the requested 
    change. The change clarifies the intent of SR [surveillance 
    requirement] 4.7.7.b.4 and the basis for the flowrates used for 
    system acceptance testing. It has been determined that operation of 
    the system at lower flow rates than those specified for surveillance 
    testing is conservative with respect to the radio-iodine removal 
    efficiency assumed for the charcoal adsorber. A higher removal 
    efficiency results in lower total exposures at the site boundary and 
    within the control room. Additionally, the system is capable of 
    maintaining the required negative pressure at the reduced flowrate.
        Given the above, it is concluded that the proposed change does 
    not result in an increase in the probability or consequences 
    associated with previously analyzed accidents.
    
    [[Page 55041]]
    
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The proposed amendment does not result in any design or 
    operational change to the ABVS, to the Nuclear Steam Supply System, 
    to the ECCS System, to the Containment Building, to the fuel or to 
    the electrical power supplies. Therefore, the proposed amendment 
    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.
        Specification 3/4.7.7 and the associated bases were reviewed to 
    determine if the proposed changes result in a reduction in the 
    margin of safety. The change to SR 4.7.7.b.4 continues to assure 
    that the system is operated consistent with the assumptions of the 
    accident analysis. The proposed changes to Bases 3/4.7.7 clarify the 
    basis for flowrates associated with ABVS surveillance test 
    requirements. All changes result in ABVS operation that is just as 
    conservative as that assumed in existing analyses.
        The proposed changes do not involve the addition or modification 
    of plant equipment, are consistent with the design basis of the ABVS 
    as described in the UFSAR, and appropriately limit operation to be 
    consistent with the assumptions of the accident analysis. As such 
    there is no 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: Salem Free Public library, 112 
    West Broadway, Salem, NJ 08079
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
    Strawn, 1400 L Street, NW, Washington, DC 20005-3502
        NRC Project Director: John F. Stolz
    
    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.
    
    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
    50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
    Units Nos. 1, 2, and 3, Maricopa County, Arizona
    
        Date of application for amendments: June 17, 1996
        Brief description of amendments request: The proposed amendments 
    would modify the technical specifications to change (1) the reference 
    method for calculating dose conversion factors (DCFs) to be used in 
    dose calculations, and (2) the upper and lower limits for operating 
    pressurizer pressure to account for new instrument uncertainties and to 
    reduce the allowed operating band.
        Date of individual notice in Federal Register: September 11, 1996 
    (61 FR 47963)
        Expiration date of individual notice: October 11, 1996
        Local Public Document Room location: Phoenix Public Library, 1221 
    N. Central Avenue, Phoenix, Arizona 85004
    
    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
    50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
    Units Nos. 1, 2, and 3, Maricopa County, Arizona
    
        Date of application for amendments: June 28, 1996
        Brief description of amendments request: The proposed amendments 
    would modify the technical specifications to increase the minimum 
    required amount of anhydrous trisodium phosphate (TSP) in the 
    containment baskets.
        Date of individual notice in Federal Register: September 11, 1996 
    (61 FR 47962), as corrected September 26, 1996 (61 FR 50535).
        Expiration date of individual notice: October 11, 1996
        Local Public Document Room location: Phoenix Public Library, 1221 
    N. Central Avenue, Phoenix, Arizona 85004
    
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
    Station, Unit No. 1, Washington County, Nebraska
    
        Date of application for amendment: August 23, 1996
        Brief description of amendment request: The proposed amendment 
    would revise Paragraph 2.B(2) of
        Facility Operating License No. DPR-40 to allow source materials in 
    the form of depleted or natural uranium as reactor fuel and to revise 
    Technical Specification 4.3.2 to include depleted uranium in describing 
    the reactor core.
        Date of individual notice in Federal Register: August 30, 1996 (61 
    FR 45995)
        Expiration date of individual notice: September 30, 1996
        Local Public Document Room location: W. Dale Clark Library, 215 
    South 15th Street, Omaha, Nebraska 68102
    
    Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
    Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
    Creeks, Manitowoc County, Wisconsin
    
        Date of application for amendment: September 19, 1996
        Brief description of amendment request: The proposed amendments 
    would change Technical Specification requirements related to the low 
    temperature overpressure protection (LTOP) system. Specifically, the 
    reactor coolant system (RCS) temperature below which LTOP is required 
    to be enabled and one high pressure safety injection pump is required 
    to be rendered inoperable would be changed from 275  deg.F to 355 
    deg.F. Also, a specification would be added stating that only one 
    reactor coolant pump shall be operated when the RCS temperature is less 
    than or equal to 125  deg.F. Finally, editorial changes would be made 
    to rename the ``Overpressure Mitigating System'' as the ``Low 
    Temperature Overpressure Protection System.'' Date of individual notice 
    in Federal Register: October 1, 1996 (61 FR 51308) Expiration date of 
    individual notice: October 31, 1996
        Local Public Document Room location: Joseph P. Mann Library, 1516 
    Sixteenth, Two Rivers, Wisconsin 54241
    
    Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
    Nuclear Power Plant, Kewaunee County, Wisconsin
    
        Date of application for amendment: September 27, 1996
        Brief description of amendment request: The proposed amendment 
    would change Technical Specification (TS) requirements related to the 
    low temperature overpressure protection (LTOP) system. Specifically, 
    the LTOP curve would be modified to define 10 CFR Part 50, Appendix G 
    pressure temperature limitations for LTOP evaluation through the end of 
    operating cycle (EOC) 33. In addition, the LTOP enabling temperature 
    and the temperature required for starting a reactor coolant pump would 
    be changed consistent with the design basis for the LTOP system. 
    Finally, the TS bases would be changed consistent with he changes 
    described above.
        Date of individual notice in Federal Register: October 7, 1996 (61 
    FR 52472)
    
    [[Page 55042]]
    
        Expiration date of individual notice: November 6, 1996
        Local Public Document Room location: University of Wisconsin, 
    Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    
    Notice Of Issuance Of Amendments ToFacility 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.
    
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
    Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
    North Carolina
    
        Date of application for amendment: July 19, 1996
        Brief description of amendment: The amendment revises the 
    containment spray nozzle surveillance interval in TS 3/4.6.2 from 5 to 
    10 years.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment No.: 67
        Facility Operating License No. NPF-63. Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44354) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 3, 1996.No significant 
    hazards consideration comments received: No
        Local Public Document Room location:  Cameron Village Regional 
    Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    
    Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
    Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas
    
        Date of amendment request: April 11, 1996, as supplemented August 
    23, 1996
        Brief description of amendments: The amendments revised the 
    Technical Specifications to permit implementation of 10 CFR Part 50, 
    Appendix J, Option B.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment Nos.: 185 and 176
        Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: May 8, 1996 (61 FR 
    20846) The additional information contained in the supplemental letter 
    dated August 23, 1996, was clarifying in nature and thus, within the 
    scope of the initial notice and did not affect the staff's proposed no 
    significant hazards consideration determination.The Commission's 
    related evaluation of the amendments is contained in a Safety 
    Evaluation dated October 3, 1996.No significant hazards consideration 
    comments received: No.
        Public Document Room location: Tomlinson Library, Arkansas Tech 
    University, Russellville, AR 72801
    
    Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
    Unit No. 1, Pope County, Arkansas
    
        Date of amendment request: April 29, 1996
        Brief description of amendment: The amendment relocated cycle 
    specific operating parameters from the Technical Specifications to the 
    Core Operating Limits Report per Generic Letter 88-16. The parameters 
    being relocated by this amendment include the variable low reactor 
    coolant system pressure trip and the variable low reactor coolant 
    system pressure-temperature protective limits.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment No.: 186
        Facility Operating License No. DPR-51: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 5, 1996 (61 FR 
    28613) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 3, 1996.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Tomlinson Library, Arkansas 
    Tech University, Russellville, AR 72801
    
    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
    Electric Station, Unit 3, St. Charles Parish, Louisiana
    
        Date of amendment request: November 7, 1995, as supplemented by 
    letter dated April 11, 1996.
        Brief description of amendment: The amendment modifies the Appendix 
    A Technical Specifications related to Safety Injection Tank level and 
    pressure setpoints.
        Date of issuance: September 27, 1996
        Effective date: September 27, 1996
        Amendment No.: 121
        Facility Operating License No. NPF-38: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: November 27, 1995 (60 
    FR 58401) The additional information contained in the supplemental 
    letter dated April 11, 1996, was clarifying in nature and thus, within 
    the scope of the initial notice and did not affect the staff's proposed 
    no significant hazards consideration determination. The Commission's 
    related evaluation of the amendment is contained in a Safety Evaluation 
    dated September 27, 1996.No significant hazards consideration comments 
    received: No.
        Local Public Document Room location:  University of New Orleans 
    Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    
    Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
    Turkey Point Plant Units 3 and 4, Dade County, Florida
    
        Date of application for amendments: July 17, 1996
        Brief description of amendments: The amendments consist of changes 
    to the Technical Specifications regarding containment leakage tests.
        Date of issuance: October 4, 1996
        Effective date: October 4, 1996
        Amendment Nos.: 192 and 186Facility Operating Licenses Nos. DPR-31 
    and DPR-41: Amendments revised the Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44357)
    
    [[Page 55043]]
    
    The Commission's related evaluation of the amendments is contained in a 
    Safety Evaluation dated October 4, 1996.No significant hazards 
    consideration comments received: No
        Local Public Document Room location: Florida International 
    University, University Park, Miami, Florida 33199
    
    Georgia Power Company, Oglethorpe Power Corporation, Municipal 
    Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
    50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, 
    Appling County, Georgia
    
        Date of application for amendments: May 21, 1996
        Brief description of amendments: The amendments revise the 
    condensate storage tank level indication to ensure that the water level 
    is sufficient to provide 50,000 gallons of water for core spray makeup 
    to the reactor pressure vessel. On September 24, 1996, based on a 
    teleconference between the licensee and the NRC project manager, it was 
    mutually agreed to change the requested implementation schedule from 90 
    days to 30 days.
        Date of issuance: October 2, 1996
        Effective date: As of the date of issuance to be implemented within 
    30 days
        Amendment Nos.: 202 and 143
        Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44358) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated October 2, 1996. No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Appling County Public Library, 
    301 City Hall Drive, Baxley, Georgia 31513
    
    GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
    Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania
    
        Date of application for amendment: January 16, 1995
        Brief description of amendment: This amendment revised the 
    Technical Specification to incorporate an improvement from 
    administrative controls section of the revised standard TS for B&W 
    plants.
        Date of issuance: October 8, 1996
        Effective date: October 8, 1996
        Amendment No.: 50Possession-Only License No. DPR-73: The amendment 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: December 20, 1995 (60 
    FR 65679). The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 8, 1996No 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
    
    IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center, 
    Linn County, Iowa
    
        Date of application for amendment: July 5, 1996
        Brief description of amendment: The amendment will support the 
    implementation of noble metal chemical addition at the Duane Arnold 
    Energy Center as a method to enhance the effectiveness of hydrogen 
    water chemistry in mitigating intergranular stress corrosion cracking 
    in reactor vessel internal components. Specifically, the amendment will 
    permit an increase of the reactor water conductivity limit in Technical 
    Specification (TS) Table 3.6.B.2-1 and several other changes in TS 
    sections 4.6.B.2.c, 4.6.B.2.d, and the associated Bases.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment No.: 218
        Facility Operating License No. DPR-49: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40020) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 3, 1996. No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Cedar Rapids Public Library, 
    500 First Street, S. E., Cedar Rapids, Iowa 52401
    
    IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center, 
    Linn County, Iowa
    
        Date of application for amendment: December 22, 1995, as 
    supplemented September 20, 1996
        Brief description of amendment: The amendment revises the Duane 
    Arnold Energy Center (DAEC) Technical Specifications (TS) Sections 
    3.7.A and 4.7.A, ``Primary Containment,'' by deleting information also 
    contained in 10 CFR Part 50, Appendix J, Option A and incorporating 
    references to the Primary Containment Leakage Rate Testing Program. 
    These changes allow the use of the performance based option of 
    containment leak testing. The amendment also adds Operability and 
    Surveillance Requirements (SRs) for the drywell air lock. Minor 
    administrative changes were also made. These changes are consistent 
    with comparable specifications in the Improved Standard Technical 
    Specifications (ITS), NUREG-1433. In addition, the staff executed 
    administrative changes and corrections to the TS Bases, as submitted in 
    two letters dated February 13, 1995. Sections changed or corrected are 
    Section 1.2, Bases; Section 2.2, Bases Reactor Coolant System 
    Integrity; Section 3.7.H/4.7.H, Bases Containment Atmosphere Dilution; 
    and Section 3.7.I/4.7.I, Bases Oxygen Concentration.
        Date of issuance: October 4, 1996
        Effective date: October 4, 1996
        Amendment No.: 219
        Facility Operating License No. DPR-49: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: January 31, 1996 (61 FR 
    3499) The September 20, 1996, submittal was clarifying in nature and 
    did not affect the no significant hazards determination. The 
    Commission's related evaluation of the amendment is contained in a 
    Safety Evaluation dated October 4, 1996.No significant hazards 
    consideration comments received: No.
        Local Public Document Room location: Cedar Rapids Public Library, 
    500 First Street, S. E., Cedar Rapids, Iowa 52401
    
    Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
    No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
    Illinois
    
        Date of application for amendment: June 28, 1996 and as 
    supplemented on September 17, 1996
        Brief description of amendment: The amendment will allow removal of 
    the Inclined Fuel Transfer System (IFTS) primary containment blind 
    flange while primary containment is required to be operable. This will 
    provide flexibility to operate the IFTS for the purpose of testing and 
    exercising the system during such conditions. Primary containment 
    integrity will be provided by an alternate means while the blind flange 
    is removed. The change will be incorporated via a provisional note into 
    Technical Specification (TS) Surveillance Requirement 3.6.1.3.3, 
    associated with TS 3.6.1.3, ``Primary Containment Isolation Valves 
    (PCIVs).''
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment No.: 107
        Facility Operating License No. NPF-62: The amendment revised the 
    Technical Specifications.
    
    [[Page 55044]]
    
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40021) The information provided in the licensee's letter of September 
    17, 1996 provided clarifying information and did not involve 
    significant changes to the original Federal Register notice.The 
    Commission's related evaluation of the amendment is contained in a 
    Safety Evaluation dated October 3, 1996.No significant hazards 
    consideration comments received: No
        Local Public Document Room location: The Vespasian Warner Public 
    Library, 120 West Johnson Street, Clinton, Illinois 61727
    
    Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
    No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
    Illinois
    
        Date of application for amendment: June 21, 1996, and as 
    supplemented by letter dated August 15, 1996
        Brief description of amendment: The amendment modifies Section 5.7, 
    ``High Radiation Areas,'' of the ``Administrative Controls'' section of 
    the Clinton Power Station technical specifications (TS). The changes 
    include: (1) allowing utilization of a Radiation Work Permit (RWP) ``or 
    equivalent'' to control entry into a high radiation area; (2) 
    clarifying the example given in the TS of individuals who are qualified 
    in radiation protection procedures; (3) clarifying the requirements for 
    when specified access controls and barriers for high radiation areas 
    within large areas like the containment may be established; (4) 
    clarifying that it is acceptable for an RWP to specify a maximum dose, 
    i.e., a specified setpoint on an alarming dosimeter in lieu of a stay 
    time for entry into a high radiation area (where an individual could 
    receive a deep dose equivalent of 3000 mrem in one hour); (5) 
    eliminating the upper dose limit for specifying the applicability of 
    the requirements of Specification 5.7.1; (6) providing additional 
    flexibility regarding the control of keys to locked doors for 
    preventing unauthorized entry into high radiation areas; (7) providing 
    alternate means of informing individuals of dose rates in immediate 
    work areas; (8) reorganizing TS Sections 5.7.1, 5.7.2, and 5.7.3 into 
    four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4); and (9) making minor 
    edits to enhance readability.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996
        Amendment No.: 108
        Facility Operating License No. NPF-62: The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40021) The August 21, 1996, submittal consisted of supporting technical 
    information which did not change the staff's initial proposed no 
    significant hazards consideration determination or expand the scope of 
    the original notice. The Commission's related evaluation of the 
    amendment is contained in a Safety Evaluation dated October 3, 1996.No 
    significant hazards consideration comments received: No.
        Local Public Document Room location: The Vespasian Warner Public 
    Library, 120 West Johnson Street, Clinton, Illinois 61727
    
    Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
    Nuclear Power Station, Unit 1, New London County, Connecticut
    
        Date of application for amendment: May 2, 1996, as supplemented by 
    letter dated August 30, 1996
        Brief description of amendment: The amendment removes Technical 
    Specification Figure 5.1, which was used in maintaining Keff 
    values, and substitutes in its place a defined requirement for maximum 
    Kinfinity for any fuel placed in the Millstone Unit 1 spent fuel 
    pool.
        Date of Issuance: October 4, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 60 days.
        Amendment No.: 97
        Facility Operating License No. DPR-21: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 17, 1996 (61 FR 
    37301) The August 30, 1996, letter provided additional, clarifying 
    information that did not change the scope of the May 2, 1996, 
    application and the initial proposed no significant hazards 
    consideration determination.The Commission's related evaluation of this 
    amendment is contained in a Safety Evaluation dated October 4, 1996.No 
    significant hazards consideration comments received: No.
        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
    
    Northern States Power Company, Docket No. 50-282, Prairie Island 
    Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota
    
        Date of application for amendment: July 15, 1996, and supplemented 
    August 22, 1996
        Brief description of amendment: The amendment allows the use of the 
    moveable in-core detector system for measurement of the core peaking 
    factors with less than 75 percent and greater than or equal to 50 
    percent of the detector thimbles available. The amendment is a one-time 
    only change for Prairie Island, Unit 1, to reduce the number of 
    required in-core detectors necessary for continued operation for the 
    remainder of Operating Cycle 18.
        Date of issuance: October 10, 1996
        Effective date: October 10, 1996, and shall remain effective for 
    the remainder of Cycle 18 only
        Amendment No.: 124
        Facility Operating License No. DPR-42. Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40024) By letter dated August 22, 1996, NSP forwarded a copy of the 
    results of its most recent low power physics tests to the NRC for use 
    as a reference and provided additional clarifying information. This 
    information was within the scope of the original application and did 
    not change the staff's initial proposed no significant hazards 
    considerations determination. Therefore, renoticing was not 
    warranted.The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated October 10, 1996. No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Minneapolis Public Library, 
    Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
    Minnesota 55401
    
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
    Station, Unit No. 1, Washington County, Nebraska
    
        Date of amendment request: May 17, 1996
        Brief description of amendment: The amendment revises Technical 
    Specifications (TS) 2.18, 3.14, 3.3, and 5.10 to relocate the 
    operability requirements for shock suppressors (snubbers) from the TS 
    to the Updated Safety Analysis Report (USAR) and incorporate snubber 
    examination and testing requirements in TS 3.3.
        Date of issuance: September 27, 1996
        Effective date: September 27, 1996
        Amendment No.: 176
    
    [[Page 55045]]
    
        Facility Operating License No. DPR-40: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44360) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 27, 1996.No 
    significant hazards consideration comments received: No.
        Local Public Document Room location: W. Dale Clark Library, 215 
    South 15th Street, Omaha, Nebraska 68102
    
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
    Station,Unit No. 1, Washington County, Nebraska
    
        Date of amendment request: August 23, 1996
        Brief description of amendment: The amendment modifies paragraph 
    2.B.(2) of
        Facility Operating License No. DPR-40 allowing the use of source 
    material, in the form of depleted or natural uranium, as reactor fuel.
        Date of issuance: October 2, 1996
        Effective date: October 2, 1996
        Amendment No.: 177
        Facility Operating License No. DPR-40: Amendment revised the 
    Operating License.
        Date of initial notice in Federal Register: August 30, 1996 (61 FR 
    45995) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 2, 1996.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: W. Dale Clark Library, 215 
    South 15th Street, Omaha, Nebraska 68102
    
    Power Authority of the State of New York, Docket No. 50-333, James 
    A. FitzPatrick Nuclear Power Plant, Oswego County, New York
    
        Date of application for amendment: January 25, 1996
        Brief description of amendment: The amendment would extend the 
    instrumentation surveillance test intervals to support 24-month 
    operating cycles. These proposed changes would eliminate the mid-cycle 
    outages to perform the Technical Specification surveillance 
    requirements.
        Date of issuance: October 2, 1996
        Effective date: As of the date of issuance to be implemented within 
    30 days.
        Amendment No.: 233
        Facility Operating License No. DPR-59: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 22, 1996 (61 FR 
    25709) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 2, 1996.No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Reference and Documents 
    Department, Penfield Library, State University of New York, Oswego, New 
    York 13126
    
    Power Authority of the State of New York, Docket No. 50-333, James 
    A. FitzPatrick Nuclear Power Plant, Oswego County, New York
    
        Date of application for amendment: March 27, 1996, as supplemented 
    April 24, 1996, and August 15, 1996
        Brief description of amendment: The proposed amendment changes 
    would permit implementation of 10 CFR Part 50, Appendix J, Option B 
    with an exception to the guidelines of Regulatory Guide 1.163 for Type 
    C testing of primary containment isolation valves in the reverse (non-
    accident) direction.
        Date of issuance: October 4, 1996
        Effective date: As of the date of issuance to be implemented within 
    30 days.
        Amendment No.: 234
        Facility Operating License No. DPR-59: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 8, 1996 (61 FR 
    20855) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 4, 1996.No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Reference and Documents 
    Department, Penfield Library, State University of New York, Oswego, New 
    York 13126
    
    Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
    Vermont Yankee Nuclear Power Station, Vernon, Vermont
    
        Date of application for amendment: August 9, 1996, as supplemented 
    September 17, 1996
        Brief description of amendment: The amendment revises the Technical 
    Specifications to revise the safety limit minimum critical power ratio 
    for cycle 19 operation from its current value of 1.07 (for the fuel 
    currently in the reactor for cycle 18) for two recirculation loop 
    operation to 1.10, and from 1.08 to 1.12 for single recirculation loop 
    operation.
        Date of issuance: October 4, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days.
        Amendment No.: 150
        Facility Operating License No. DPR-28: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44364) The September 17, 1996, letter provided clarifying information 
    that did not change the scope of the August 9, 1996, application and 
    initial proposed no significant hazards consideration determination.The 
    Commission's related evaluation of the amendment is contained in a 
    Safety Evaluation dated October 4, 1996.No significant hazards 
    consideration comments received: No
        Local Public Document Room location: Brooks Memorial Library, 224 
    Main Street, Brattleboro, VT 05301
    
    South Carolina Electric & Gas Company, South Carolina Public 
    Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
    Station, Unit No. 1, Fairfield County, South Carolina
    
        Date of application for amendment: April 16, 1996, as supplemented 
    July 25, 1996
        Brief description of amendment: The amendment permits 
    implementation of 10 CFR Part 50, Appendix J, Option B, ``Performance-
    Based Requirements.''
        Date of issuance: October 2, 1996
        Effective date: October 2, 1996
        Amendment No.: 135
        Facility Operating License No. NPF-12: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 3, 1996 (61 FR 
    34898) The July 25, 1996, supplement provides clarifying information 
    and did not change the scope of the initial notice. The Commission's 
    related evaluation of the amendment is contained in a Safety Evaluation 
    dated October 2, 1996.No significant hazards consideration comments 
    received: No
        Local Public Document Room location: Fairfield County Library, 300 
    Washington Street, Winnsboro, SC 29180
    
    Southern California Edison Company, et al, Docket No. 50-206, San 
    Onofre Nuclear Generating Station, Unit No. 1, San Diego County, 
    California
    
        Date of application for amendment: March 13, 1996
        Brief description of amendment: The change revises the San Onofre 
    Unit 1 License Condition 2.D. This change eliminates a reporting 
    requirement that is redundant to reporting requirements in 10 CFR 50.72 
    and 50.73. Additionally, the amendment makes administrative and 
    editorial changes to the Permanently Defueled Technical Specifications.
        Date of issuance: October 3, 1996
    
    [[Page 55046]]
    
        Effective date: October 3, 1996
        Amendment No.: 158
        Facility Operating License No. DPR-13: The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40028) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 3, 1996.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Science Library, University of 
    California, Irvine, California 92713
    
    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: December 6, 1995, as 
    supplemented by letters dated August 30, 1996, and September 20, 1996
        Brief description of amendments: These amendments revise Technical 
    Specifications (TS) Section 4.3 ``Fuel Storage'' to allow fuel 
    assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/
    o) to be stored in both the spent fuel racks and the new fuel racks. 
    Additionally, TS Section 3.7.18 ``Spent Fuel Assembly Storage,'' 
    Figures 3.7.18-1 ``Unit 1 Fuel Minimum Burnup vs. Initial Enrichment 
    for Region II Racks,'' and 3.7.18-2 ``Units 2 and 3 Fuel Minimum Burnup 
    vs. Initial Enrichment for Region II Racks,'' are being revised and 
    relabeled.
        Date of issuance: October 3, 1996
        Effective date: October 3, 1996, to be implemented within 30 days 
    as of the date of issuance.
        Amendment Nos.: Unit 2 - 131; Unit 3 - 120
        Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: April 10, 1996 (61 FR 
    15997) The August 30, 1996, and September 20, 1996, letters provided 
    additional clarifying information and did not change the initial no 
    significant hazards consideration determination.The Commission's 
    related evaluation of the amendments is contained in a Safety 
    Evaluation dated October 3, 1996.No significant hazards consideration 
    comments received: No.
        Temporary Local Public Document Room location: Science Library, 
    University of California, P. O. Box 19557, Irvine, California 92713
    
    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 31, 1996 (TXX-96433)
        Brief description of amendments: The amendments revised core safety 
    limit curves (Technical Specification (TS) Figure 2.1-1a) and new N-16 
    setpoint values and parameters (TS Table 2.1-1) for Unit 1, and 
    reference to topical report RXE-95-001-P as an approved methodology for 
    small break loss of coolant accident analysis for Units 1 and 2.
        Date of issuance: September 30, 1996
        Effective date: September 30, 1996, to be implemented within 30 
    days
        Amendment Nos.: 52 and 38
        Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44362) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated September 30, 1996.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
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of application for amendment: April 12, 1996, as supplemented 
    by letters dated August 2, 1996, August 19, 1996, and September 5, 
    1996.
        Brief description of amendment: The amendment revises the Technical 
    Specifications to address the installation of laser welded tube sleeves 
    in the Callaway Plant steam generators.
        Date of issuance: October 1, 1996
        Effective date: October 1, 1996, and will be implemented within 30 
    days of the date of issuance.
        Amendment No.: 116
        Facility Operating License No. NPF-30: The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 8, 1996 (61 FR 
    20857) The August 2, 1996, August 19, 1996, and September 5, 1996, 
    supplemental letters provided clarifying information and did not change 
    the original no significant hazards consideration determination. The 
    Commission's related evaluation of the amendment is contained in a 
    Safety Evaluationdated October 1, 1996.No significant hazards 
    consideration comments received: No.
        Local Public Document Room location: Callaway County Public 
    Library, 710 Court Street, Fulton, Missouri 65251.
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of application for amendment: April 17, 1996, as supplemented 
    by letters dated July 15, 1996, July 31, 1996, and August 28, 1996.
        Brief description of amendment: The amendment would change 
    Technical Specification (TS) 3/4.3 to support a future modification to 
    replace existing digital portions of the main steam and feedwater 
    isolation system (MSFIS) with digital processor equipment and would 
    authorize revision of the FSAR to include a description of the MSFIS 
    modification.
        Date of issuance: October 1, 1996
        Effective date: October 1, 1996, to be implemented prior to startup 
    from the Callaway Plant Refuel 8.
        Amendment No.: 117
        Facility Operating License No. NPF-30: The amendment revised the 
    Technical Specifications and the Final Safety Analysis Report.
        Date of initial notice in Federal Register: June 5, 1996 (61 FR 
    28619) The July 15, 1996, July 31, 1996 and August 28, 1996 
    supplemental letters provided additional clarifying information and did 
    not change the staff's original no significant hazards consideration 
    determination. The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 1, 1996.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Callaway County Public 
    Library, 710 Court Street, Fulton, Missouri 65251
    
    Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
    Vermont Yankee Nuclear Power Station, Vernon, Vermont
    
        Date of application for amendment: April 4, 1996
        Brief description of amendment: The amendment revises the Technical 
    Specifications regarding the surveillance requirement for control rod 
    over-travel by moving the specific testing methodology to licensee 
    administratively controlled documents. Specifically, the amendment 
    removes the requirement in Specification 4.3.B.1(b) to verify prior to 
    coupling that the over-travel indicating light is working properly by 
    withdrawing an uncoupled control rod drive to the over-travel position.
    
    [[Page 55047]]
    
        Date of issuance: September 30, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days.
        Amendment No.: 149
        Facility Operating License No. DPR-28: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 8, 1996 (61 FR 
    20860) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 30, 1996.No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Brooks Memorial Library, 224 
    Main Street, Brattleboro, VT 05301
    
    Washington Public Power Supply System, Docket No. 50-397, Nuclear 
    Project No. 2, Benton County, Washington
    
        Date of application for amendment: August 9, 1996
        Brief description of amendment: The amendment changes the 
    operations manager qualification requirements to allow either of two 
    alternatives (having held a senior reactor operator's license or having 
    been certified for equivalent senior reactor operator knowledge) to the 
    requirement for the operations manager to hold a senior reactor 
    operator's license.
        Date of issuance: October 1, 1996
        Effective date: October 1, 1996
        Amendment No.: 148
        Facility Operating License No. NPF-21: The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: August 28, 1996 (61 FR 
    44350) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated October 1, 1996.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Richland Public Library, 955 
    Northgate Street, Richland, Washington 99352
    
    Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
    Nuclear Power Plant, Kewaunee County, Wisconsin
    
        Date of application for amendment: July 3, 1996, as supplemented on 
    July 23, August 28, and September 16, 1996
        Brief description of amendment: The amendment revises Kewaunee 
    Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator 
    Tubes,'' and its associated basis, by revising the acceptance criteria 
    for indications of tube degradation occurring in the tubesheet crevice 
    region.
        Date of issuance: October 2, 1996
        Effective date: October 2, 1996
        Amendment No.: 129
        Facility Operating License No. DPR-43: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40031) The July 23, August 28, and September 16, 1996, submittals 
    provided clarifying information that did not change the initial 
    proposed no significant hazards consideration determination.The 
    Commission's related evaluation of the amendment is contained in a 
    Safety Evaluation dated October 2, 1996.No significant hazards 
    consideration comments received: No.
        Local Public Document Room location: University of Wisconsin, 
    Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    
    Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
    Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
    Manitowoc County, Wisconsin
    
        Date of application for amendments: May 29, 1996, as supplemented 
    August 20, 1996
        Brief description of amendments: These amendments revise Technical 
    Specification (TS) Section 15.4.4, ``Containment Tests,'' to 
    incorporate the provisions of 10 CFR Part 50, Appendix J, ``Primary 
    Reactor Containment Leakage Testing for Water-Cooled Power Reactors,'' 
    Option B. Revisions have also been made to TS Sections 15.1, 
    ``Definitions,'' 15.3.6, ``Containment System,'' and 15.6, 
    ``Administrative Controls,'' to support the proposed changes to Section 
    15.4.4.
        Date of issuance: October 9, 1996
        Effective date: October 9, 1996, to be implemented within 45 days.
        Amendment Nos.: Unit 1 - 169 and Unit 2 - 173
        Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: July 3, 1996 (61 FR 
    34901) The supplemental information did not affect the staff's initial 
    no significant hazards consideration determination.The Commission's 
    related evaluation of the amendments is contained in a Safety 
    Evaluation dated October 9, 1996.No significant hazards consideration 
    comments received: No
        Local Public Document Room location: Joseph P. Mann Library, 1516 
    Sixteenth Street, Two Rivers, Wisconsin 54241
    
    Notice Of Issuance Of Amendments To Facility Operating Licenses And 
    Final Determination Of No Significant Hazards Consideration And 
    Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
    Circumstances)
    
        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 for the 
    amendment 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.
        Because of exigent or emergency circumstances associated with the 
    date the amendment was needed, there was not time for the Commission to 
    publish, for public comment before issuance, its usual 30-day Notice of 
    Consideration of Issuance of Amendment, Proposed No Significant Hazards 
    Consideration Determination, and Opportunity for a Hearing.
        For exigent circumstances, the Commission has either issued a 
    Federal Register notice providing opportunity for public comment or has 
    used local media to provide notice to the public in the area 
    surrounding a licensee's facility of the licensee's application and of 
    the Commission's proposed determination of no significant hazards 
    consideration. The Commission has provided a reasonable opportunity for 
    the public to comment, using its best efforts to make available to the 
    public means of communication for the public to respond quickly, and in 
    the case of telephone comments, the comments have been recorded or 
    transcribed as appropriate and the licensee has been informed of the 
    public comments.
        In circumstances where failure to act in a timely way would have 
    resulted, for example, in derating or shutdown of a nuclear power plant 
    or in prevention of either resumption of operation or of increase in 
    power output up to the plant's licensed power level, the Commission may 
    not have had an opportunity to provide for public comment on its no 
    significant hazards consideration determination. In such case, the 
    license amendment has been issued without opportunity for comment. If 
    there has been some time for public comment but less than 30 days, the 
    Commission may provide an
    
    [[Page 55048]]
    
    opportunity for public comment. If comments have been requested, it is 
    so stated. In either event, the State has been consulted by telephone 
    whenever possible.
        Under its regulations, the Commission may issue and make an 
    amendment immediately effective, notwithstanding the pendency before it 
    of a request for a hearing from any person, in advance of the holding 
    and completion of any required hearing, where it has determined that no 
    significant hazards consideration is involved.
        The Commission has applied the standards of 10 CFR 50.92 and has 
    made a final determination that the amendment involves no significant 
    hazards consideration. The basis for this determination is contained in 
    the documents related to this action. Accordingly, the amendments have 
    been issued and made effective 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 
    application for amendment, (2) the amendment to Facility Operating 
    License, 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 room for the particular facility involved.
        The Commission is also offering an opportunity for a hearing with 
    respect to the issuance of the amendment. By November 22, 1996, 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. Since the Commission has made a final determination 
    that the amendment involves no significant hazards consideration, if a 
    hearing is requested, it will not stay the effectiveness of the 
    amendment. Any hearing held would take place while the amendment is in 
    effect.
        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 20555-001, Attention: Docketing and Services 
    Branch, or may be delivered to the Commission's Public Document Room, 
    the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
    date. Where petitions are filed during the last 10 days of the notice 
    period, it is requested that the petitioner promptly so inform the 
    Commission by a toll-free telephone call to Western Union at 1-(800) 
    248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
    should be given Datagram Identification Number N1023 and the following 
    message addressed to (Project Director): petitioner's name and 
    telephone number, date petition was mailed, plant name, and publication 
    date and page number of this Federal Register notice. A copy of the 
    petition should also be sent to the Office of the General Counsel, U.S. 
    Nuclear Regulatory Commission, Washington, DC 20555-001, 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 the 
    factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    
    [[Page 55049]]
    
    Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
    Nuclear Station, Units 1 and 2, York County, South Carolina
    
        Date of application for amendments: September 21, 1996
        Brief description of amendments: The amendments approve changes to 
    the Updated Final Analysis Report (UFSAR), and require that the changes 
    be submitted with the next update of the UFSAR pursuant to 10 CFR 
    50.71(e). The associated Safety Evaluation delineates the staff's 
    review and findings, including finding that the as-built condition of 
    the subject power system protective devices is acceptable as-is.
        Date of issuance: September 28, 1996
        Effective date: September 28, 1996
        Amendment Nos.: 153 and 145
        Facility Operating License Nos. NPF-35 and NPF-52: The amendments 
    revised the Updated Final Safety Analysis Report. Public comments 
    requested as to proposed no significant hazards consideration: Yes. The 
    NRC staff published a public notice of the proposed amendments, issued 
    a proposed finding of no significant hazards consideration, and 
    requested that any comments on the proposed no significant hazards 
    consideration be provided to the staff no later than 5:00 p.m., 
    September 28, 1996. The notice was published in ``The Herald'' of Rock 
    Hill, South Carolina, from September 25 through 27, 1996. No comments 
    have been received.
        The Commission's related evaluation of the amendments, finding of 
    exigent circumstances, consultation with the State of South Carolina, 
    and final determination of no significant hazards consideration are 
    contained in a Safety Evaluation dated September 28, 1996.
        Local Public Document Room location: York County Library, 138 East 
    Black Street, Rock Hill, South Carolina 29730
        Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
    South Church Street, Charlotte, North Carolina 28242
        NRC Project Director: Herbert N. Berkow
    
    PECO Energy Company, Public Service Electric and Gas Company, 
    Delmarva Power and Light Company, and Atlantic City Electric 
    Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 
    No. 2, York County, Pennsylvania
    
        Date of application for amendment: March 25, 1996 as supplemented 
    by letters dated August 23, 1996 and September 27, 1996.
        Brief description of amendment: The amendment revises Peach Bottom 
    Technical Specification 2.1.1.2 safety limit minimum critical power 
    ratios to be consistent with the use of GE-13 fuel in the Unit 2 core 
    for operating cycle 12.
        Date of issuance: September 27, 1996
        Effective date: As of date of issuance
        Amendment No.: 217
        Facility Operating License No. DPR-44: Amendment revised the 
    Technical Specifications.Public comments requested as to proposed no 
    significant hazards consideration: Yes (61 FR 45997). That notice 
    provided an opportunity to submit comments on the Commission's proposed 
    no significant hazards consideration determination. No comments have 
    been received. The notice also provided an opportunity to request a 
    hearing by September 30, 1996, but indicated that if the Commission 
    makes a final no significant hazards consideration determination any 
    such hearing would take place after issuance of the amendment.The 
    Commission's related evaluation of the amendment, finding of exigent 
    circumstances, and final no significant hazards consideration 
    determination are contained in a Safety Evaluation dated September 27, 
    1996.
        Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice 
    President and General Counsel, PECO Energy Company, 2301 Market Street, 
    Philadelphia, Pennsylvania 19101
        Local Public Document Room location: Government Publications 
    Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
    Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
    PA 17105.
        Dated at Rockville, Maryland, this 16th day of October 1996.
        For the Nuclear Regulatory Commission
    John A. Zwolinski,
    Acting Director, Division of Reactor Projects - I/II, Office of Nuclear 
    Reactor Regulation
    [FR Doc. 96-27025 Filed 10-22-96; 8:45 am]
    BILLING CODE 7590-O1-F
    
    
    

Document Information

Published:
10/23/1996
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
96-27025
Dates:
October 3, 1996
Pages:
55028-55049 (22 pages)
PDF File:
96-27025.pdf