X96-10925. Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations  

  • [Federal Register Volume 61, Number 187 (Wednesday, September 25, 1996)]
    [Notices]
    [Pages 50338-50351]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: X96-10925]
    
    
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    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
    
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    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 August 30, 1996, through September 13, 1996. 
    The last biweekly notice was published on September 11, 1996.
    
    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 October 25, 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.
        If a hearing is requested, the Commission will make a final 
    determination on the issue of no significant hazards consideration. The
    
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    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.
    
    Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
    Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
    STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
    Will County, Illinois
    
        Date of amendment request: August 2, 1996
        Description of amendment request: The proposed amendment would 
    eliminate from the licenses the requirement to conduct corrosion 
    testing for the laser welded steam generator sleeves.
        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.
        This proposed change only involves deleting laboratory testing 
    requirements designed to demonstrate service life of laser welded 
    sleeved tubes in the presence of a crevice. Current inspection 
    requirements ensure that premature degradation is identified and 
    that tubes containing degraded sleeve joints are plugged. 
    Operational primary-to-secondary leakage limits ensure that 
    appropriate action is taken if sleeve degradation results in 
    leakage. These actions will ensure that offsite dose will be 
    maintained within a small percentage of 10 CFR 100 limits. Failure 
    of a sleeve joint is bounded by the Steam Generator Tube Rupture 
    event evaluated in the [Updated Final Safety Analysis Report] UFSAR. 
    Therefore, the laboratory testing to determine service life of 
    sleeved tube joints in the presence of a crevice does not provide 
    any further useful data. The change does not result in the 
    installation of any new equipment, and no existing equipment is 
    modified.
        Therefore, this proposed change 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.
        This proposed change only addresses deleting the laboratory 
    testing requirements designed to demonstrate service life of sleeved 
    tubes in the presence of a crevice. Sleeved tubes will continue to 
    be inspected and plugged in accordance with existing requirements 
    which are sufficient to ensure detection and repair of degraded 
    tubes. Premature degradation of tubes is addressed through primary-
    to-secondary leakage monitoring and leakage limits. No new equipment 
    is being installed and no existing equipment is being modified by 
    this proposed change. Also, no new system configurations will be 
    introduced as a result of this proposed change. Therefore, no new or 
    different failure modes are being introduced by deleting the 
    laboratory testing.
        Thus, this proposed change does 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.
        This proposed change only involves deleting laboratory testing 
    requirements designed to demonstrate service life of sleeved tubes 
    in the presence of a crevice. Sleeve integrity will be monitored 
    during the operating cycle through the current primary-to-secondary 
    leakage monitoring program. In the event of premature degradation of 
    a sleeve joint that results in tube leakage, plant shutdown will 
    occur as required by Technical Specifications and administrative 
    requirements in accordance with approved plant procedures. Sleeved 
    tubes will be monitored for degradation in accordance with the 
    existing inservice inspection requirements which monitors a minimum 
    20 percent random sleeve sample size. Any tubes with defective 
    sleeve joints will be plugged as required by Technical 
    Specifications. Service life of sleeved tubes in the presence of a 
    crevice, as predicted by laboratory testing, does not affect the 
    margin of safety of the plant. 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 
    requested amendments involve no significant hazards consideration.
        Local Public Document Room location: For Byron, the Byron Public 
    Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
    for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
    Wilmington, Illinois 60481.
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60603
        NRC Project Director: Robert A. Capra
    
    Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
    County Station, Units 1 and 2, LaSalle County, Illinois
    
        Date of amendment request: July 15, 1996
        Description of amendment request: The proposed amendments would 
    revise Technical Specifications (TS) and associated Bases to relocate 
    the fire protection program elements from the TS to the Fire Protection 
    Program. The affected TS sections are 3/4.3.7.9, ``Fire Detection 
    Instrumentation;'' 3/4.7.5, ``Fire Suppression Systems;'' 3/4.7.6, 
    ``Fire Rated Assemblies;'' and 6.1.C.4,
    
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    ``Fire Brigade Staffing.'' In addition, the amendments revise the 
    Operating License to replace existing fire protection license 
    conditions with the NRC's standard fire protection license condition. 
    These changes are made in accordance with the guidance provided in 
    Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
    Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
    from Technical Specifications.'' Also, the May 19, 1995, proposed 
    revision to remove the fire protection requirements from the TS (60 FR 
    35067) is withdrawn.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        1) Involve a significant increase in the probability or 
    consequences of an accident previously evaluated because:
        This amendment request does not involve any actual changes to 
    the fire protection systems at the station. It involves an 
    administrative change which relocates the control of the Fire 
    Protection Program from each unit's operating license and technical 
    specifications to the station Fire Protection Program, as suggested 
    in Generic Letters 86-10 and 88-12. Therefore, the relocation of 
    these controls does not affect the assumptions for any of the 
    accident analysis contained in Chapter 15 of the [Updated Final 
    Safety Analysis Report] UFSAR.
        The Fire Protection Technical Specifications which are to be 
    relocated to the Fire Protection Program will be controlled by the 
    proposed fire protection license condition and 10CFR 50.59. These 
    controls ensure that the requested changes maintain the same level 
    of control for the Fire Protection Program as that which currently 
    exists in the Technical Specifications. Therefore, this change is 
    administrative in nature and does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        2) Create the possibility of a new or different kind of accident 
    from any accident previously evaluated because:
        This amendment request does not involve any physical changes to 
    the fire protection systems or reduce the level of control of the 
    Fire Protection Program. It therefore does not create the 
    possibility of a new or different type of accident than any 
    previously described in the UFSAR.
        3) Involve a significant reduction in the margin of safety 
    because:
        The same level of control which is currently applied to the Fire 
    Protection Program by the limiting conditions for operation and the 
    surveillance requirements of the technical specifications will be 
    included in the controls applied by the unit licenses and the Fire 
    Protection Program. Therefore, the margin of safety as defined in 
    the technical specification bases will not be reduced by this 
    proposed amendment.
        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 
    requested amendments involve no significant hazards consideration.
        Local Public Document Room location: Jacobs Memorial Library, 
    Illinois Valley Community College, Oglesby, Illinois 61348.
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60603
        NRC Project Director: Robert A. Capra
    
    Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
    Nuclear Power Station, Units 1 and 2, Lake County, Illinois
    
        Date of amendment request: July 26, 1996, and supplemented on 
    September 3, 1996
        Description of amendment request: The proposed amendments would 
    allow licensee control of the reactor coolant system (RCS) pressure and 
    temperature (P/T) limits for heatup, cooldown, low temperature 
    operation and hydrostatic testing. They would also revise the reactor 
    vessel material surveillance program specimen withdrawal schedule such 
    that the Unit 2 removal of capsule X is delayed until 19 Effective Full 
    Power Years (EFPY). This change affects the schedule for withdrawing 
    surveillance capsules from the reactor vessel for testing to measure 
    the impact of neutron irradiation of the vessel material and is 
    required by Section III.B.3 of 10 CFR Part 50, Appendix H, ``Reactor 
    Vessel Material Surveillance Program Requirements.'' The schedule must 
    be approved by the Nuclear Regulator Commission (NRC) before 
    implementation.
        Based on input from the Babcock and Wilcox Owners Group Reactor 
    Vessel Working Group, the data from Zion, Unit 2, capsule X would be 
    more useful in the overall Master Integrated Reactor Vessel 
    Surveillance Program (MIRVP) context if irradiated to the ASTM E185-82 
    maximum of twice the peak End Of Life (EOL) vessel fluence, because 
    data at higher fluences is needed to characterize irradiation behavior 
    at the higher EOL fluences characteristic of other non-Commonwealth 
    Edison MIRVP vessels. For this reason, the licensee is proposing 
    withdrawing and testing Zion, Unit 2, capsule X at 19 EFPY, which is 
    currently estimated to occur at refueling outage Z2R18, in the year 
    2002.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        1. Does the change involve a significant increase in the 
    probability or consequences of an accident previously evaluated?
        The proposed change revises the 10 CFR 50, Appendix H reactor 
    vessel material specimen withdrawal schedule. Neither the specimens, 
    nor the process of withdrawal of the specimens, are considered as 
    initiators for any previously evaluated accident. Further, data at 
    all fluence levels of current interest based on ASTM E185-82 has 
    already been obtained from seven Zion Unit 1 and 2 capsules which 
    have been tested, and the existing evaluations show the reactor 
    vessel fracture toughness properties to be as expected, and 
    providing the required safety margin. Extending the time for 
    withdrawal of the specimen does not adversely affect the pressure 
    and temperature limit curves for the reactor vessel. Regulatory 
    Guide 1.99, Rev. 2, was used to prepare the conservative pressure 
    and temperature limit curves which continue to be requirements.
        Additionally, Zion Station participates in the B&W Owners Group 
    Reactor Vessel Working Group designed to significantly increase the 
    amount of PWR surveillance data. Under this Working Group, Zion 
    Station data contributes to the overall understanding of reactor 
    vessel material irradiation behavior at high EOL fluences, and 
    obtains the benefit of data from other plants. This program 
    complements the Zion Station program so that postponement of the 
    specimen withdrawal will have minimal impact on the understanding of 
    the irradiation effects on the Zion Station reactor vessel. 
    Therefore, this change does not involve a significant increase in 
    the probability or consequences of an accident previously evaluated.
        The proposed revision to the specimen withdrawal schedule does 
    not change the system operation or design, and therefore, does not 
    change the response of any required structures, systems or 
    components in the mitigation of any evaluated accident. As such, 
    this change does not involve a significant increase in the 
    consequences of an accident previously evaluated.
        The proposed change relocates the RCS P/T, LTOP [low-temperature 
    overpressure protection] limitations, and supporting information 
    from the Technical Specifications to Licensee control, specifically 
    a Pressure Temperature Limits Report (PTLR). Compliance with these 
    limitations will continue to be required by the Technical 
    Specifications, however the limitations themselves will be relocated 
    to a Licensee controlled document. Changes to these limitations will 
    be controlled by Section 5.6.6 of the Technical Specifications. 
    Changes to the RCS P/T limits can only be made in accordance with 
    the approved methodologies listed in the Technical Specifications 
    which will, in combination with the limitations that continue to be
    
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    imposed by the Technical Specifications, continue to assure the 
    function of the reactor vessel as a pressure boundary. Revisions to 
    the LTOP limits can only be made in accordance with the approved 
    methodologies listed in the Technical Specifications, with any 
    resulting setpoint changes controlled through a process which 
    utilizes 10 CFR 50.59. Therefore, 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 change does not necessitate a physical alteration 
    of the plant (no new or different equipment will be installed). The 
    proposed revision to the specimen withdrawal schedule does not 
    change the system operation or design, and therefore, does not 
    introduce any new failure mechanisms. The proposed specimen 
    withdrawal schedule continues to provide the required data for 
    subsequent reactor vessel evaluations, and previous data has 
    confirmed the confidence in the integrity of the reactor vessel well 
    beyond the completion of the evaluations following the proposed 
    withdrawal. Therefore, this revision to the withdrawal schedule does 
    not create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        The proposed change does not necessitate a physical alteration 
    of the plant (no new or different equipment will be installed). The 
    Technical Specifications will continue to retain requirements to 
    maintain the RCS within acceptable operational limitations and to 
    assure operability of the LTOP system. As such, the Technical 
    Specifications will continue to require compliance with these 
    limitations. Thus, this change does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        3. Does this change involve a significant reduction in a margin 
    of safety?
        The proposed change to the specimen withdrawal schedule will not 
    result in a significant reduction in a margin of safety because it 
    has no impact on any safety analysis assumptions. Additionally, data 
    at all fluence levels of current interest based on ASTM E185-82 has 
    already been obtained with the seven Zion Unit 1 and 2 capsules 
    which have been tested, and the existing evaluations show the 
    reactor vessel fracture toughness properties to be as expected, and 
    providing the required safety margin. The current pressure and 
    temperature limits are conservative and also provide sufficient 
    margin to ensure the integrity of the reactor vessel. The proposed 
    change to the withdrawal schedule does not adversely impact these 
    curves. Therefore, this change does not involve a significant 
    reduction in a margin of safety.
        The proposed change will not result in a significant reduction 
    in a margin of safety because it has no impact on any safety 
    analysis assumptions. Any future changes to the RCS P/T, LTOP 
    limits, or supporting information must be performed in accordance 
    with approved NRC methodologies, and compliance with the limitations 
    relocated to the PTLR will continue to be required by the Technical 
    Specifications. Additionally, any revision to the LTOP limits which 
    result in setpoint changes will be controlled through a process 
    which utilizes 10 CFR 50.59. Therefore, this 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 
    requested amendments involve no significant hazards consideration.
        Local Public Document Room location: Waukegan Public Library, 128 
    N. County Street, Waukegan, Illinois 60085
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60603
        NRC Project Director: Robert A. Capra
    
    Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
    Michigan
    
        Date of amendment request: September 5, 1996 (NRC-96-0075)
        Description of amendment request: The proposed amendment would 
    revise Technical Specification (TS) sections 2.1.2 and 3.4.1.1 to 
    incorporate cycle-specific safety limit minimum critical power ratios 
    (SLMCPRs) for the core that will be loaded during the upcoming 
    refueling outage expected to commence in November 1996.
        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.
        The derivation of the revised SLMCPRs for Fermi 2 for 
    incorporation into the TS, and its use to determine cycle-specific 
    thermal limits, have been performed using NRC-approved methods. 
    Additionally, interim implementing procedures, which incorporate 
    cycle-specific parameters, have been used which result in a more 
    restrictive value for the SLMCPR. These calculations do not change 
    the method of operating the plant and have no effect on the 
    probability of an accident initiating event or transient. The basis 
    of the MCPR Safety Limit is to ensure that no mechanistic fuel 
    damage is calculated to occur if the limit is not violated. The new 
    SLMCPRs preserve the existing margin to transition boiling and the 
    probability of fuel damage is not increased. Therefore, the proposed 
    TS change does 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 change results from analysis of the Cycle 6 core 
    reload using the same fuel types as previous cycles. These changes 
    do not involve any new method for operating the facility and do not 
    involve any facility modifications. No new initiating events or 
    transients result from these changes. Therefore, the proposed TS 
    change does 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 as defined in the TS Bases will remain the 
    same. The new SLMCPRs are calculated using NRC-approved methods 
    which are in accordance with the current fuel design and licensing 
    criteria. Additionally, interim implementing procedures, which 
    incorporate cycle-specific parameters, have been used. The MCPR 
    Safety Limit remains high enough to ensure that greater than 99.9% 
    of all fuel rods in the core will avoid transition boiling if the 
    limit is not violated, thereby preserving the fuel cladding 
    integrity. Therefore, the proposed TS change 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: 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 Hannon
    
    Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
    Nuclear Station, Unit 1, Claiborne County, Mississippi
    
        Date of amendment request: July 31, 1996, as supplemented by letter 
    dated September 5, 1996. These letters supersede the application 
    submitted in letter dated May 9, 1996, which was noticed in the Federal 
    Register on June 5, 1996 (61 FR 28614).
        Description of amendment request: The amendment request would (1) 
    increase the safety limit minimum critical power ratio (MCPR) for two 
    loop operation and single loop operation to 1.12 and 1.14, 
    respectively, and (2) add a General Electric topical report to the list 
    of documents describing the analytical methods used to determine the 
    core operating limits. The proposed changes are to Section 2.1.1, 
    Reactor
    
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    Core Safety Limits, and Section 5.6.5, Core Operating Limits Report 
    (COLR), respectively, of the Technical Specifications (TSs). This 
    amendment would go into effect in Operating Cycle 9, at the end of the 
    upcoming Refueling Outage 8, and the plant will have a mixed core of 
    Siemens Power Corporation (SPS) 9x9-5 and General Electric (GE) GE11 
    reload fuel. The licensee also proposed changes to the Bases of the TSs 
    associated with the above proposed changes.
        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:
        I. The proposed change does not significantly increase the 
    probability or consequences of an accident previously evaluated.
        The Minimum Critical Power Ratio (MCPR) safety limit is defined 
    in the Bases to Technical Specification 2.1.1 as that limit which 
    ``ensures that during normal operation and during Anticipated 
    Operational Occurrences (AOOs), at least 99.9% of the fuel rods in 
    the core do not experience transition boiling.'' The MCPR safety 
    limit is re-evaluated for each reload and, for GGNS [Operating] 
    Cycle 9, the analyses have concluded that a two-loop MCPR safety 
    limit of 1.12 based on the application of the generic GE MCPR 
    methodology is necessary to ensure that this acceptance criterion is 
    satisfied. For single-loop operation, a MCPR safety limit of 1.14 
    based on the generic GE MCPR methodology was determined to be 
    necessary. Core MCPR operating limits are developed to support the 
    Technical Specification 3.2 requirements and ensure these safety 
    limits are maintained in the event of the worst-case transient. 
    Since the MCPR safety limit will be maintained at all times, 
    operation under the proposed changes will ensure at least 99.9% of 
    the fuel rods in the core do not experience transition boiling. 
    Therefore, The Minimum Critical Power Ratio (MCPR) safety limit 
    change does not affect the probability or consequences of an 
    accident.
        The implementation of GE's GESTAR-II approved methodology has no 
    effect on the probability or consequences of any accidents 
    previously evaluated. One exception to GESTAR is that the mis-
    oriented and mis-located bundle events will continue to be analyzed 
    as accidents subject to the acceptance criteria in the current 
    licensing basis. The design of the GE11 fuel bundles is such that 
    the bundles are not likely to be mis-oriented or mis-located and the 
    normal administrative controls will be in effect for assuring proper 
    orientation and location. Therefore, the probability of a fuel 
    loading error is not increased. This analysis ensures that 
    postulated dose releases will not exceed a small fraction (10 
    percent) of 10 CFR 100 limits.
        Therefore, the consequences of accidents previously evaluated 
    are unchanged.
        II. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The GE11 fuel to be used in [Operating] Cycle 9 is of a design 
    compatible with fuel present in the core and used in the previous 
    cycle. Therefore, the GE11 fuel will not create the possibility of a 
    new or different kind of accident. The proposed changes do not 
    involve any new modes of operation, any changes to setpoints, or any 
    plant modifications. They introduce revised MCPR safety limits that 
    have been proved to be acceptable for Cycle 9 operation. Compliance 
    with the applicable criterion for incipient boiling transition 
    continues to be ensured. The proposed MCPR safety limits do not 
    result in the creation of any new precursors to an accident.
        Therefore, the proposed changes do not create the possibility of 
    a new or different type of accident from any accident previously 
    evaluated.
        III. The proposed change does not involve a significant 
    reduction in a margin of safety.
        The MCPR safety limits have been evaluated to ensure that during 
    normal operation and during AOOs [abnormal operating occurrences], 
    at least 99.9% of the fuel rods in the core do not experience 
    transition boiling. Therefore, the implementation of the proposed 
    changes in the MCPR safety limit ensure there is no reduction in the 
    margin of safety.
        As with the current SPC methodology, GGNS will implement only 
    the NRC-approved revisions to GE's GESTAR methodology. This GE 
    methodology is similar to those SPC reports currently listed in TS 
    5.6.5 and it will be applied in a similar, conservative fashion. One 
    exception to GESTAR is that the mis-oriented and mis-located bundle 
    events will continue to be analyzed as accidents subject to the 
    acceptance criteria in the current licensing basis. This analysis 
    ensures that postulated dose releases will not exceed a small 
    fraction (10 percent) of 10CFR100 [10 CFR Part 100] limits. On this 
    basis, the implementation of this GE methodology 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: Judge George W. Armstrong 
    Library, 220 S. Commerce Street, Natchez, MS 39120
        Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
    Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
        NRC Project Director: William D. Beckner
    
    Gulf States Entergy, 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 1, 1996
        Description of amendment request: The proposed amendment would 
    revise the Technical Specifications (TSs) to incorporate requirements 
    for limiting the time that the hydrogen mixing isolation valves on the 
    drywell are open. The requirements were contained in the old TSs and 
    with the conversion to the Improved Standard Technical Specifications, 
    the requirements were inadvertently changed. The proposed action is to 
    restore requirements to meet the licensing basis for the River Bend 
    Station. The proposed amendment would also change the time from 7 days 
    to 31 days to determine the cumulative time the valves are open.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        The proposed changes in this submittal put the requirements that 
    were in the original Technical Specifications for the Hydrogen 
    Mixing System back into the current Technical Specifications. The 
    changes reenstate into the Technical Specifications limitations that 
    were previously agreed to between River Bend and the Nuclear 
    Regulatory Commission in the FSAR Safety Evaluation Report for the 
    Hydrogen Mixing System.
        The River Bend SER states in Supplement 2, Section 6.2.4, 
    ``Since the applicant has not demonstrated that these valves are 
    capable of closing under accident conditions in the drywell, certain 
    restrictions apply. Technical Specification 3.6.6.2 specifies that 
    in Operating Modes 1 and 2, the total number of hours used should 
    not exceed 5 hours/365 days and in Operating Mode 3 the number of 
    hours should be limited to 90 hours/365 days.'' To date, the 
    hydrogen mixing isolation valves have not been fully demonstrated to 
    be capable of closing under accident conditions in the drywell. The 
    old Standard Technical Specifications (Attachment 2) used at River 
    Bend reflected this condition. When conversion to ITS was made, 
    these requirements were dropped but should not have been. In 
    addition, the requirement to operate the hydrogen mixing system 
    every 92 days during Modes 1, 2, and 3 was added without 
    consideration for the requirements in the River Bend Safety 
    Evaluation Report.
        Consequently, for these proposed change, since the requirements 
    already exist and are being reenstated into the Technical 
    Specifications, this change is administrative in nature. The 
    requirements have remained in place through the SER, but were
    
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    inadvertently removed from the Technical Specifications. This change 
    places the requirements from the SER back into the Technical 
    Specifications.
        In addition, changing the requirement from the old Technical 
    Specifications for determining the cumulative time that the hydrogen 
    mixing inlet and outlet valves are open from every 7 days to every 
    31 days is again administrative in nature, since this only changes 
    the frequency with which a given requirement is tracked 
    administratively. It does not change the actual requirement in any 
    way.
        Consequently, since both of these changes are administrative in 
    nature and only incorporate requirements into the Technical 
    Specifications that already existed in the RBS FSAR Safety 
    Evaluation Report, the changes proposed in this amendment request do 
    not change the probability or consequences of an accident previously 
    evaluated.
        This proposed change does not involve a change to the plant 
    design or operation. As a result, the proposed change does not 
    affect any of the parameters or conditions that could contribute to 
    the initiation of any accidents.
        The changes proposed in this amendment request are 
    administrative in nature and merely add requirements back into the 
    Technical Specifications that were inadvertently deleted during the 
    conversion to ITS. Because of the administrative nature of the 
    proposed changes, it is not possible to create a new or different 
    kind of accident from any accident previously evaluated.
        The proposed changes in this amendment request reenstate 
    requirements into the Technical specifications that are contained 
    present in the RBS FSAR Safety Evaluation Report. These requirements 
    were inadvertently deleted during the conversion to ITS.
        Because of the administrative nature of these Technical 
    Specification changes, there is no change to 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: 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
    
    Houston Lighting & Power Company, City Public Service Board of San 
    Antonio, Central Power and Light Company, City of Austin, Texas, 
    Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
    Matagorda County, Texas
    
        Date of amendment request: August 15, 1996.
        Description of amendment request: The proposed amendments would 
    remove a requirement for performance of a surveillance incorporating a 
    high toxic gas test signal.
        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.
        Analyses were performed to evaluated postulated releases of 
    potentially hazardous chemicals for their impact on Control Room 
    habitability. The latest revision of these analyses shows that none 
    of the potentially hazardous chemicals utilized onsite or in the 
    surrounding 5-mile radius around the South Texas Project pose a 
    credible hazard to the Control Room. Consequently, there is no need 
    to ensure that the Control Room Makeup and Cleanup Filtration System 
    can automatically switch into a recirculation mode of operation by 
    isolating the normal supply and exhaust flow in response to a High 
    Toxic Gas test signal. Therefore, elimination of the unnecessary 
    surveillance has no effect on the probability of an accident or its 
    consequences.
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The Toxic Gas Monitoring System was provided to protect against 
    hazardous toxic gas releases only. Verifying automatic switch into 
    the recirculation mode of operation is no longer necessary since the 
    Toxic Gas Analyzers have been removed. This change does not affect 
    other tests for verification of automatic switching into the 
    recirculation mode of operation. Therefore, the proposed change does 
    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.
        Analyses have shown that none of the chemicals onsite and within 
    a 5-mile radius of the South Texas Project pose a credible hazard to 
    the facility. Automatic switching of the Control Room Makeup and 
    Cleanup Filtration System will continue to be verified using test 
    signals from other sources.
        Based upon this evaluation, the South Texas Project has 
    concluded that these changes do not involve any significant hazards 
    considerations.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    request for amendments involves no significant hazards consideration.
        Local Public Document Room location: Wharton County Junior College, 
    J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
        Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
    Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
        NRC Project Director: William D. Beckner
    
    Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
    No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
    Illinois
    
        Date of amendment request: August 15, 1996
        Description of amendment request: A Federal Register Notice on May 
    22, 1996 (61 FR 25707), stated that revisions were being proposed to 
    Clinton Power Station Technical Specification (TS) 3.3.6.2, ``Secondary 
    Containment Isolation Instrumentation;'' TS 3.3.7.1, ``Control Room 
    Ventilation System Instrumentation;'' TS 3.6.1.2, ``Primary Containment 
    Air Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves;'' TS 
    3.6.4.1, ``Secondary Containment;'' TS 3.6.4.2, ``Secondary Containment 
    Isolation Dampers;'' TS 3.6.4.3, ``Standby Gas Treatment;'' TS 3.7.3, 
    ``Control Room Ventilation;'' and TS 3.7.4, ``Control Room AC System.'' 
    By letter dated August 15, 1996, the licensee revised their proposal to 
    consolidate the above changes under a newly proposed Special Operations 
    LCO (i.e., LCO 3.10.10, ``Single Control Rod Withdrawal - Refueling''). 
    Therefore, the Description of Amendment Request to the TSs has changed 
    as described herein. The Basis for No Significant Hazards Consideration 
    has not changed and is repeated below.
        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 eliminate CORE ALTERATIONS as an 
    applicable condition requiring operability of the primary and 
    secondary containment and control room ventilation system. As stated 
    in the BASES for the associated Technical Specifications, 
    operability of these systems is primarily required for mitigation of 
    the design basis accident - fuel handling accident (DBA-FHA) and 
    design basis accident - loss of coolant accident (DBA-LOCA). The 
    performance of CORE ALTERATIONS alone is neither a
    
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    precursor to, nor a condition during which these DBAs are postulated 
    to occur. The proposed changes only delete CORE ALTERATIONS as an 
    applicable condition for the affected Technical Specifications. All 
    other applicable MODES or specified conditions, including operations 
    with the potential for draining the reactor vessels (OPDRVs) and the 
    movement of irradiated fuel assemblies within the primary or 
    secondary containment, remain unchanged. Further, the limitations 
    placed on the handling of light loads are also unchanged. The 
    Technical Specifications (and the separate requirements imposed on 
    the handling of light loads) will thus continue to require that 
    systems or functions designed to mitigate design-basis/previously 
    evaluated accidents are OPERABLE during the relevant operating MODES 
    or conditions. On the basis of the above, it is concluded that the 
    requested amendment will not increase the probability or 
    consequences of any accident previously evaluated.
        2. The proposed changes do not involve any modification to the 
    plant design or to the operation of plant systems (except to 
    determine when certain analyzed accident-mitigating systems or 
    features are required to be OPERABLE). The failure modes considered 
    for the proposed changes are the same as those previously 
    considered, therefore, it can be concluded that no new failure modes 
    will be created. On this basis, the proposed amendment will not 
    create the possibility of a new or different kind of accident from 
    any accident previously evaluated.
        3. The changes being made to eliminate CORE ALTERATIONS as an 
    applicable condition for which certain LCOs must be met, do not 
    eliminate the requirements for operability of those systems or 
    features assumed to mitigate design-basis or analyzed accidents 
    during the applicable MODES when such systems or features are 
    assumed to be available for performing their mitigating function. 
    The safety margins assumed or established by the accident analyses 
    for those design-basis events (as described in the accident analyses 
    of the Clinton Power Station Updated Final Safety Analysis Report) 
    therefore remain unchanged. Further, the proposed changes do not 
    impact the controls imposed on the handling of light loads 
    (including unirradiated fuel assemblies) for ensuring that such 
    activities cannot result in an event that yields consequences more 
    severe than those calculated for the DBA-FHA. With respect to 
    reactivity concerns during refueling operations (MODE 5), all 
    systems or features required to be OPERABLE for precluding 
    inadvertent criticality and monitoring reactivity changes will 
    continue to be required OPERABLE as per the current Technical 
    Specification requirements. The deletion of CORE ALTERATIONS as an 
    applicable condition only applies to the noted systems which do not 
    contribute to precluding reactivity events. Based on the above, 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 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: Vespasian Warner Public 
    Library, 120 West Johnson Street, Clinton, Illinois 61727
        Attorney for licensee: Leah Manning Stetzner, Vice President, 
    General Counsel, and Corporate Secretary, 500 South 27th Street, 
    Decatur, Illinois 62525
        NRC Project Director: Gail H. Marcus
    
    Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
    Atomic Power Station, Lincoln County, Maine
    
        Date of amendment request: August 12, 1996
        Description of amendment request: The proposed amendment would add 
    an additional circumstance to Exception 2 of Technical Specification 
    (TS) 3.6, Emergency Core Cooling and Containment Spray Systems, during 
    which operation of a service water/component cooling pump subsystem is 
    permitted at reduced flow to flush the service water header or inlet 
    strainer. The Bases for this TS would be augmented to support the 
    additional circumstance of reduced service water flow.
        The proposed amendment would also modify the valve surveillance 
    requirements of TS 4.6.A.1.b, Periodic Testing of ECCS Valves, to 
    provide an exception to surveillance requirements for those locked 
    valves that are inaccessible during power operations or located in a 
    locked high radiation area. The Bases for this TS would be augmented to 
    support the change in surveillance requirements.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration. The NRC staff's analysis is presented below.
        1. The proposed amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        Invocation of the proposed addition to Exception 2 to TS 3.6 would 
    not alter any associated Remedial Action completion time, nor those of 
    TS 3.0.A, Nonconformance with a Limiting Condition for Operation. The 
    evolutions for which this amendment is intended (flushing a heat 
    exchanger inlet strainer or cleaning a service water header that has 
    become fouled)are administratively controlled by procedures that 
    require review and approval by the Plant Operation Review Committee.
        The proposed change to TS 4.6.A.1.b would revise the surveillance 
    requirements for a very limited number of locked manual valves in the 
    emergency core cooling system (ECCS). The purpose of the surveillance 
    requirements is unchanged and is intended to verify that locked valves 
    remain in their correct position. The position of the valves is not 
    changed and the revised surveillance requirements will continue to 
    demonstrate ECCS valve operability.
        Thus, the proposed amendment does not involve a significant 
    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 addition to Exception 2 to TS 3.6 recognizes that 
    service water cleaning and flushing are operations that are required to 
    maintain heat transfer capability and equipment reliability. The 
    proposed amendment does not affect the design of the plant and do not 
    permit operation of the plant outside the currently allowed modes of 
    operation.
        The proposed change to TS 4.6.A.1.b maintains verification of ECCS 
    valve operability, while requiring no changes in system configuration 
    to perform surveillance testing. System functional performance is not 
    adversely affected.
        Thus, the proposed amendment does 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.
        The proposed change to TS 3.6 does not significantly alter the 
    availability or condition of applicable equipment and therefore does 
    not alter the accident analyses or the conclusions associated with 
    that equipment. The proposed change permits service water flow to be 
    reduced below that required for operation of the ECCS in the 
    recirculation mode, for a short time. The time during which flow is 
    reduced and both the mussel control and flushing evolutions are 
    administratively controlled by procedures reviewed and approved by 
    the Plant Operation Review Committee.
        The proposed change to TS 4.6.A.1.b maintains verification of valve 
    operability. Valve position surveillances will continue to be conducted 
    in accordance with plant Technical Specifications to ensure valve 
    operational readiness.
        Thus, there is no significant reduction in a margin of safety.
        Based on this review, it appears that the three standards of 10 
    CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
    
    [[Page 50346]]
    
    determine that the amendment request involves no significant hazards 
    consideration.
        Local Public Document Room location: Wiscasset Public Library, High 
    Street, P.O. Box 367, Wiscasset, ME 04578
        Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
    Power Company, 329 Bath Road, Brunswick, ME 04011 NRC Deputy Director: 
    John A. Zwolinski
    
    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
    Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
    Obispo County, California
    
        Date of amendment requests: June 7, 1996
        Description of amendment requests: The proposed amendments would 
    revise the combined Technical Specifications (TS) for the Diablo Canyon 
    Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications 
    (TS) 3/4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region 
    2,'' and 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool 
    Region 1,'' to allow storage of fuel assemblies in a checkerboard 
    pattern in region 2 of the spent fuel pool.
        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.
        Analysis indicates that allowing fuel storage in a checkerboard
        pattern with empty storage cells in region 2 of the spent fuel
        pool will not result in an inadvertent criticality event. The 
    keff will continue to remain below 0.95 as required to meet the 
    acceptance criteria in the NRC Standard Review Plan, Section 9.1.1.
        Therefore, the proposed changes do 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 change to allow fuel storage in a checkerboard pattern with 
    no minimum burnup requirements in region 2 of the spent fuel pool 
    would designate locations where a fuel assembly could be incorrectly 
    placed. However, the incorrect placement of a fuel assembly has been 
    analyzed and would not cause an inadvertent criticality or any other 
    accident.
        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 change does not involve a significant reduction 
    in a margin of safety.
        The NRC Standard Review Plan, Section 9.1.1, acceptance 
    criterion of a keff of 0.95 provides the margin to criticality. 
    An analysis was performed that concluded that the proposed change to 
    allow fuel storage in spent fuel pool region 2 in a checkerboard 
    pattern meets the acceptance criterion.
        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 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment requests involve no significant hazards consideration.
        Local Public Document Room location: California Polytechnic State 
    University, Robert E. Kennedy Library, Government Documents and Maps 
    Department, San Luis Obispo, California 93407
        Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
    Electric Company, P.O. Box 7442, San Francisco, California 94120
        NRC Project Director: William H. Bateman
    
    Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
    Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
    Alabama
    
        Date of amendment request: June 6, 1996 (TS 372)
        Description of amendment request: The proposed amendment revises 
    Section 6 of the Browns Ferry Nuclear Plant Units 1, 2, and 3 technical 
    specifications. Administrative controls associated with quality 
    assurance are relocated to the licensee's Nuclear Quality Assurance 
    Plan, consistent with Administrative Letter 95-06, and provides 
    revisions that make Section 6 more consistent with the improved 
    Standard Technical Specifications. Additional administrative changes 
    are included to ensure consistent terminology within the 
    specifications, and to update obsolete items such as titles and 
    addresses. The proposed amendment also includes minor editorial 
    changes.
        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 amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The proposed TS change to revise items 1 through 28 above 
    (Section I, Description of the Proposed Change) was evaluated and 
    the proposed TS changes were determined to be administrative in 
    nature. The changes [items 2 through 9, 11, 17 through 21, 23, 26, 
    and 27] involve administrative title changes of TVA management 
    positions, the updating of an NRC mailing address and an NRC 
    regional office title. In addition, certain sections [items 1, 10, 
    12, 13, 24, and 25] are being relocated into other licensee 
    documents for which those provisions are adequately controlled by 
    regulatory requirements. [Items 14, 15, 16, 22, and 28 are editorial 
    changes.] These changes do not affect any of the design basis 
    accidents. They do not involve 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 TS change to revise items 1 through 28 above 
    (Section I, Description of the Proposed Change) was evaluated and 
    the proposed TS changes were determined to be administrative in 
    nature. The changes involve administrative title changes of TVA 
    management positions, the updating of an NRC mailing address and an 
    NRC regional office title. In addition, certain sections are being 
    relocated into other licensee documents for which those provisions 
    are adequately controlled by regulatory requirements. These changes 
    do not affect any of the design basis accidents. No modifications to 
    any plant equipment are involved. There are no effects on system 
    interactions made by these changes. They do not create the 
    possibility of a new or different kind of accident from an accident 
    previously evaluated.
        3. The proposed amendment does not involve a significant 
    reduction in a margin of safety.
        The proposed TS change to revise items 1 through 28 above 
    (Section I, Description of the Proposed Change) was evaluated and 
    the proposed TS changes were determined to be administrative in 
    nature. The changes involve administrative title changes of TVA 
    management positions, the updating of an NRC mailing address and an 
    NRC regional office title. In addition, certain sections are being 
    relocated into other licensee documents for which those provisions 
    are adequately controlled by regulatory requirements. The margin of 
    safety as reported in the basis for the TSs is not reduced. The 
    proposed change is administrative and does not impact any technical 
    information contained in the bases of the TS.
        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: Athens Public Library, South 
    Street, Athens, Alabama 35611
    
    [[Page 50347]]
    
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
    Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
    Alabama
    
        Date of amendment request: August 30, 1996 (TS 380)
        Description of amendment request: The proposed amendment deletes 
    License Condition 2.C.(3) regarding thermal water quality standards 
    from the licenses for the Browns Ferry Nuclear Plant Units 1, 2, and 3.
        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 amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The proposed License Condition change is an adminstrative change 
    and has no relationship to plant safety analyses. Therefore, this 
    change does not increase the frequency of the precursors to design 
    basis events or operational transients analyzed in the BFN [Browns 
    Ferry Nuclear Plant] Final Safety Analysis Report. Likewise, the 
    proposed changes will not increase the 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 License Condition change is an administrative 
    change and has no relationship to plant safety analyses. Thus, the 
    change does not create any type of new accident sequences. Likewise, 
    the proposed amendment does 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 a margin of safety.
        The proposed License Condition change is an administrative 
    change and has no relationship to plant safety analyses. Therefore, 
    the proposed amendment does not involve a reduction in the margin of 
    safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Athens Public Library, South 
    Street, Athens, Alabama 35611
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    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.
    
    Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
    Nuclear Power Station, Units 1 and 2, Lake County, Illinois
    
        Date of amendment request: August 16, 1996
        Description of amendment request: This notice relates to your 
    submittal to remove the uncertainty term from the specified distance 
    and remove the footnote which specifies the time frame it is 
    applicable.
        Date of publication of individual notice in Federal Register: 
    September 11, 1996 (61 FR 47968)
        Expiration date of individual notice: October 11, 1996
        Local Public Document Room location: location: Waukegan Public 
    Library, 128 N. County Street, Waukegan, Illinois 60085.
    
    Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
    Nuclear Power Station, Units 1 and 2, Lake County, Illinois
    
        Date of amendment request: September 3, 1996
        Description of amendment request: This notice relates to your 
    submittal to modify Technical Specification Section 4.3.1.B.4.A.10.a 
    which provides the acceptance criteria for steam generator tube repairs 
    by adding a footnote which references the cleanliness and 
    nondestructive examination requirements as described in CEN-629-P, 
    Revision 00, ``Repair of Westinghouse Series 44 and 51 Steam Generator 
    Tubes Using Leak Tight Sleeves.'' Date of publication of individual 
    notice in Federal Register: September 11, 1996 (61 FR 47966)
        Expiration date of individual notice: October 11, 1996
        Local Public Document Room location: location: Waukegan Public 
    Library, 128 N. County Street, Waukegan, Illinois 60085.
    
    PECO Energy Company, Public Service Electric and Gas Company, 
    Delmarva Power and Light Company, and Atlantic City Electric 
    Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
    Station, Unit Nos. 2 and 3, York County, Pennsylvania
    
        Date of amendment request: March 25, 1996, as supplemented by 
    letter dated August 23, 1996
        Brief description of amendment request: The proposed amendment 
    would revise the safety limit minimum critical power ratios (SLMCPRs) 
    to support use of GE-13 fuel at PBAPS, Units 2 and 3. Date of 
    publication of individual notice in Federal Register: August 30, 1996 
    (61 FR 45997)
        Expiration date of individual notice: September 30, 1996
        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.
    
    Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna 
    Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
    
        Date of amendment request: May 28, 1996, as supplemented by letter 
    dated July 25, 1996
        Brief description of amendment request: The proposed amendment 
    would revise the Minimum Critical Power Ratio safety limit values, 
    adding two references to reflect the use of the ANF-B Critical Power 
    Ratio Correlation and to reflect the use of the ABB Combustion 
    Engineering licensing methodology, with a modification to the 
    associated Bases.
        Date of publication of individual notice in Federal Register: 
    September 9, 1996 (61 FR 47529)
        Expiration date of individual notice: October 9, 1996
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    
    [[Page 50348]]
    
    Notice Of Issuance Of Amendments To Facility Operating Licenses
    
        During the period since publication of the last biweekly notice, 
    the Commission has issued the following amendments. The Commission has 
    determined for each of these amendments that the application complies 
    with the standards and requirements of the Atomic Energy Act of 1954, 
    as amended (the Act), and the Commission's rules and regulations. The 
    Commission has made appropriate findings as required by the Act and the 
    Commission's rules and regulations in 10 CFR Chapter I, which are set 
    forth in the license amendment.
        Notice of Consideration of Issuance of Amendment to Facility 
    Operating License, Proposed No Significant Hazards Consideration 
    Determination, and Opportunity for A Hearing in connection with these 
    actions was published in the Federal Register as indicated.
        Unless otherwise indicated, the Commission has determined that 
    these amendments satisfy the criteria for categorical exclusion in 
    accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
    no environmental impact statement or environmental assessment need be 
    prepared for these amendments. If the Commission has prepared an 
    environmental assessment under the special circumstances provision in 
    10 CFR 51.12(b) and has made a determination based on that assessment, 
    it is so indicated.
        For further details with respect to the action see (1) the 
    applications for amendment, (2) the amendment, and (3) the Commission's 
    related letter, Safety Evaluation and/or Environmental Assessment as 
    indicated. All of these items are available for public inspection at 
    the Commission's Public Document Room, the Gelman Building, 2120 L 
    Street, NW., Washington, DC, and at the local public document rooms for 
    the particular facilities involved.
    
    Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
    Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
    
        Date of application for amendment: January 30, 1996, as 
    supplemented May 20, 1996
        Brief description of amendment: This amendment revises the 
    Technical Specifications (TS) to: (1) add TS 4.6.1.5 to provide 
    criteria for 24-hour full-load testing of the emergency diesel 
    generators (EDGs) to be performed during each refueling outage; (2) 
    revise TS 4.6.1.2 to allow testing of the EDG protective bypasses 
    listed in TS 3.7.1.d to be done independent of the safety injection or 
    loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the 
    EDG protective bypass inspection.
        Date of issuance: September 11, 1996
        Effective date: September 11, 1996
        Amendment No. 174
        Facility Operating License No. DPR-23. Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: February 28, 1996 (61 
    FR 7546) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 11, 1996. The May 20, 
    1996, letter provided clarifying information that did not change the 
    initial proposed no significant hazards consideration determination. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: location: Hartsville Memorial 
    Library, 147 West College Avenue, Hartsville, South Carolina 29550
    
    Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
    Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
    
        Date of application for amendment: December 10, 1995, as 
    supplemented August 1, 1996, and September 4, 1996.
        Brief description of amendment: This amendment revises Technical 
    Specification (TS) Section 3.5.1 and Tables 3.5-2, 3, and 4 concerning 
    the reactor trip system, engineering safety feature actuation system, 
    and isolation function.
        Date of issuance: September 12, 1996Effective date: September 12, 
    1996
        Amendment No. 175
        Facility Operating License No. DPR-23. Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: February 14, 1996 (61 
    FR 5812). The August 1, 1996, and September 4, 1996, submittals 
    provided administrative changes to the TS pages 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 September 12, 1996.No significant hazards 
    consideration comments received: No
        Local Public Document Room location: Hartsville Memorial Library, 
    147 West College Avenue, Hartsville, South Carolina 29550
    
    Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear 
    Station, Unit 1, York County, South Carolina
    
        Date of amendment request: September 30, 1994, as supplemented 
    September 18, 1995, January 19, March 15, May 16, and August 27, 1996
        Description of amendment: The amendment revises the Technical 
    Specifications to reflect the new setpoints, operational parameters, 
    and approved analysis methodologies associated with replacement of the 
    Unit 1 steam generators. The amendment also deletes references to steam 
    generator tube repair methods, which will no longer be applicable after 
    the replacement, and clarifies initial surveillances.
        Date of issuance: August 29, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days
        Amendment No.: 151
        Facility Operating License No. NPF-35: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: April 10, 1996 (61 FR 
    15986) The May 16 and August 27, 1996, letters provided clarifying 
    information that did not change the scope of the September 30, 1994, 
    application and the initial proposed no significant hazards 
    consideration determination. The Commission's related evaluation of the 
    amendments is contained in a Safety Evaluation dated August 29, 1996. 
    No significant hazards consideration comments received: No
        Local Public Document Room location: York County Library, 138 East 
    Black Street, Rock Hill, South Carolina 29730
    
    GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
    Nuclear Generating Station, Ocean County, New Jersey
    
        Date of application for amendment: July 17, 1996, as supplemented 
    August 28, 1996 (TSCR 242, Rev. 2). This application supersedes 
    applications dated February 23 (TSCR 242) and June 19, 1996 (TSCR 242, 
    Rev. 1).
        Brief description of amendment: The amendment changes the Technical 
    Specifications (TS) to allow the implementation of 10 CFR Part 50, 
    Appendix J, Option B.
        Date of Issuance: September 3, 1996
        Effective date: September 3, 1996, to be implemented within 30 days 
    of issuance
        Amendment No.: 186
        Facility Operating License No. DPR-16. Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40019) Supersedes notice dated March 27, 1996 (61 FR 13526). The August 
    28, 1996, supplement provided updated and corrected TS and bases pages. 
    These
    
    [[Page 50349]]
    
    revisions were within the scope of the original application and did not 
    change the staff's initial proposed no significant hazards 
    consideration determination. Therefore renoticing was not warranted. 
    The Commission's related evaluation of this amendment is contained in a 
    Safety Evaluation dated September 3, 1996. No significant hazards 
    consideration comments received: No.
        Local Public Document Room location: location: Ocean County 
    Library, Reference Department, 101 Washington Street, Toms River, NJ 
    08753
        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: February 22, 1996, as 
    supplemented by letter dated July 3, 1996
        Brief description of amendment: The amendment revises the Clinton 
    Power Station Technical Specifications for the drywell to permit bypass 
    testing on a 10-year frequency with increased testing if performance 
    degrades, changes the drywell air lock testing and surveillance 
    requirements, deletes action notes for the drywell air lock and drywell 
    isolation valves when the bypass leakage limit is not met, and deletes 
    the specific leakage limits for the drywell air lock seal.
        Date of issuance: September 4, 1996
        Effective date: September 4, 1996
        Amendment No.: 106
        Facility Operating License No. NPF-62: The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: April 24, 1996 (61 FR 
    18170) The July 3, 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 September 4, 1996. 
    No significant hazards consideration comments received: No
        Local Public Document Room location: location: The Vespasian Warner 
    Public Library, 120 West Johnson Street, Clinton, Illinois 61727
    
    North Atlantic Energy Service Corporation, Docket No. 50-443, 
    Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
    
        Date of amendment request: June 20, 1996
        Description of amendment request: The proposed amendment modifies 
    the Seabrook Station Appendix A Technical Specifications (TSs) for the 
    Electrical Power Systems, Onsite Power Distribution. Specifically, the 
    proposed amendment changes TS 3.8.3.1, Action a., to increase from 8 
    hours to 7 days the allowable time that 480-volt Emergency Bus 
    E64 may be less than fully energized.
        Date of issuance: August 30, 1996
        Effective date: As of date of issuance, to be implemented within 60 
    days.
        Amendment No.: 48
        Facility Operating License No. NPF-86. Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 26, 1996 (61 FR 
    33142) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated August 30, 1996. No significant 
    hazards consideration comments received: No
        Local Public Document Room location:  location: Exeter Public 
    Library, Founders Park, Exeter, NH 03833
    
    Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
    Nuclear Power Station, Unit 1, New London County, Connecticut
    
        Date of application for amendment: April 25, 1996
        Brief description of amendment: The amendment modifies the 
    calibration requirement for the source range monitors and intermediate 
    range monitors by noting that the sensors are excluded.
        Date of issuance: August 19, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days.
        Amendment No.: 96
        Facility Operating License No. DPR-21. Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 19, 1996 (61 FR 
    31183) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated August 19, 1996. No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: 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
    
    Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
    Millstone Nuclear Power Station, Unit No. 2, New London County, 
    Connecticut
    
        Date of application for amendment: March 28, 1996
        Brief description of amendment: The amendment changes Technical 
    Specification 3.7.7, ``Sealed Source Contamination,'' and its Bases 
    that modify the criteria for testing sealed sources for contamination 
    and leakage. The approved changes are consistent with the testing 
    criteria currently used at the Millstone Nuclear Power Station, Unit 
    No. 3, the Haddam Neck Plant, and the Seabrook Station.
        Date of issuance: September 4, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 60 days.
        Amendment No.: 202
        Facility Operating License No. DPR-65: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 8, 1996 (61 FR 
    20853) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 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, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
    Rope Ferry Road, Waterford, CT 06385
    
    Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
    Power Plant, Unit 3, Humboldt County, California
    
        Date of application for amendment: March 13, 1996
        Brief description of amendment: This amendment revised the 
    Technical Specification by incorporating position changes to reflect a 
    proposed plant staff reorganization.
        Date of issuance: September 6, 1996
        Effective date: This license amendment is effective as of the date 
    of its issuance and must be fully implemented no later than 30 days 
    from the date of issuance.
        Amendment No.: 31Facility License No. DPR-7: This amendment revised 
    the Technical Specifications
        Date of initial notice in Federal Register: April 24, 1996 (61 FR 
    18174) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 6, 1996. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Humboldt County Library, 1313 
    3rd Street, Eureka, California 95501
    
    [[Page 50350]]
    
    Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
    Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
    Pennsylvania
    
        Date of application for amendments: February 23, 1996, as 
    supplemented by letter dated June 28, 1996
        Brief description of amendments: These amendments change the 
    Technical Specification Requirement 4.6.2.1d concerning drywell-to-
    suppression chamber bypass testing interval to correspond with the 
    interval for Primary Containment Integrated Leak Rate Testing under 10 
    CFR Part 50, Appendix J, Option B.
        Date of issuance: September 6, 1996
        Effective date: September 6, 1996
        Amendment Nos.: 160 and 131
        Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: April 10, 1996 (61 FR 
    15992) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated September 6, 1996. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    
    Power Authority of The State of New York, Docket No. 50-286, Indian 
    Point Nuclear Generating Unit No. 3, Westchester County, New York
    
        Date of application for amendment: June 21, 1996, as supplemented 
    August 19, 1996, and August 21, 1996.
        Brief description of amendment: The amendment extends the 
    surveillance interval on certain instruments from 18 to 24 months.
        Date of issuance: September 5, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days.
        Amendment No.: 168
        Facility Operating License No. DPR-64: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    49027) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 5, 1996. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: location: White Plains Public 
    Library, 100 Martine Avenue, White Plains, New York 10610.
    
    Public Service Electric & Gas Company, Docket No. 50-354, Hope 
    Creek Generating Station, Salem County, New Jersey
    
        Date of application for amendment: March 6, 1996, as supplemented 
    by letter dated May 30, 1996.
        Brief description of amendment: The amendment changes Technical 
    Specification (TS) 3.8.1, ``A.C. Sources - Operating,'' to decrease the 
    minimum fuel oil storage capacity of the Emergency Diesel Generator 
    Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In addition, 
    footnote ** is deleted from TS 3.8.1.1.b.2. The TS change also adds an 
    Action Statement to address remedial action when a fuel oil transfer 
    pump becomes inoperable.
        Date of issuance: September 10, 1996
        Effective date: As of date of issuance, to be implemented within 90 
    days.
        Amendment No.: 96
        Facility Operating License No. NPF-57: This amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: July 3, 1996 (61 FR 
    34897) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 10, 1996. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: location: Pennsville Public 
    Library, 190 S. Broadway, Pennsville, New Jersey 08070
    
    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: December 22, 1995
        Brief description of amendment: The change revises the San Onofre 
    Unit 1 License Condition to delete a reference to License Condition 
    2.C(4) from License Condition 2.D. This change eliminates a reporting 
    requirement for violations of the physical protection plans that is 
    redundant to reporting requirements in 10 CFR 73.71 and 10 CFR Part 73 
    Appendix G.
        Date of issuance: August 30, 1996
        Effective date: August 30, 1996 and shall be implemented no later 
    than 30 days from August 30, 1996.
        Amendment No.: 157
        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 August 30, 1996. No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Science Library, University of 
    California, Irvine, California 92713
    
    Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
    50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
    County, Alabama
    
        Date of amendments request: June 12, 1996
        Brief description of amendments: The amendments revise the reactor 
    core safety limits, Overtemperature delta T (OTDT) and Overpressure 
    delta T (OPDT) reactor trip setpoints and allowable values, and the 
    power distribution limits associated with implementation of Relaxed 
    Axial Offset Control (RAOC) and FQ surveillance. The amendments 
    also include changes to the Bases associated with these specifications 
    and surveillances.
        Date of issuance: September 3, 1996
        Effective date: As of the date of issuance to be implemented within 
    30 days
        Amendment Nos.: 121 and 113
        Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
    the Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40029) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated September 3, 1996. No 
    significant hazards consideration comments received: No
        Local Public Document Room location: location: Houston-Love 
    Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, 
    Dothan, Alabama 36302
    
    Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
    50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
    County, Alabama
    
        Date of amendments request: June 20, 1996
        Brief description of amendments: The amendments revise the 
    Technical Specifications to reflect the implementation of 10 CFR Part 
    50, Appendix J, Option B.
        Date of issuance: September 3, 1996
        Effective date: As of the date of issuance to be implemented within 
    30 days
        Amendment Nos.: 122 and 114
        Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
    the Technical Specifications.
        Date of initial notice in Federal Register: July 31, 1996 (61 FR 
    40030) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated September 3, 1996. No 
    significant hazards consideration comments received: No
    
    [[Page 50351]]
    
        Local Public Document Room location: location: Houston-Love 
    Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, 
    Dothan, Alabama 36302
    
    Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
    Plant, Unit 1, Rhea County, Tennessee
    
        Date of application for amendment: July 31, 1996
        Brief description of amendment: The amendment revises Technical 
    Specification 3.6.12 to allow a one-time extension of the 3-month 
    surveillance requirement for the ice condenser lower inlet doors.
        Date of issuance: September 9, 1996
        Effective date: As of the date of issuance, to be implemented 
    within 30 days
        Amendment No.: 3
        Facility Operating License No. NPF-90: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: August 8, 1996 (61 FR 
    41431) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 9, 1996. No 
    significant hazards consideration comments received: None
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1001 Broad Street, Chattanooga, TN 37402
    
    Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
    Nuclear Power Plant, Kewaunee County, Wisconsin
    
        Date of application for amendment: February 19, 1996, as 
    supplemented on July 3 and August 26, 1996
        Brief description of amendment: The amendment revises Kewaunee 
    Nuclear Power Plant Technical Specification Section 4.2 and its 
    associated basis by allowing the application of a voltage-based repair 
    limit for the steam generator tube support plate intersections 
    experiencing outside diameter stress corrosion cracking. The repair 
    criteria are based on guidance provided in Generic Letter 95-05, 
    ``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
    affected by Outside Diameter Stress Corrosion Cracking,'' dated August 
    3, 1995, and on associated industry guidance.
        Date of issuance: September 11, 1996
        Effective date: September 11, 1996, and is to be implemented within 
    30 days of the date of issuance.
        Amendment No.: 126
        Facility Operating License No. DPR-43: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: April 10, 1996 (61 FR 
    15999) The July 3 and August 26, 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 
    September 11, 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
    
    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
    Creek Generating Station, Coffey County, Kansas
    
        Date of amendment request: October 24, 1995, and superseded by 
    letter dated May 16, 1996.
        Brief description of amendment: The amendment adopts ASTM D3803-
    1989 as the laboratory testing standard for charcoal samples from the 
    charcoal absorbers in the control room filtration system, control 
    building pressurization system, and the auxiliary/fuel building 
    emergency exhaust system. The output of the heaters in the control 
    building pressurization system is reduced from a nominal 15kW to a 
    nominal 5kW and the acceptance criterion for the testing of the 
    charcoal absorbers is changed.
        Date of issuance: September 4, 1996
        Effective date: September 4, 1996, to be implemented within 120 
    days of the date of issuance.
        Amendment No.: 102
        Facility Operating License No. NPF-42. The amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 5, 1996 (61 FR 
    28622) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated September 4, 1996. No 
    significant hazards consideration comments received: No.
        Local Public Document Room location: locations: Emporia State 
    University, William Allen White Library, 1200 Commercial Street, 
    Emporia, Kansas 66801 and Washburn University School of Law Library, 
    Topeka, Kansas 66621 Dated at Rockville, Maryland, this 18th day of 
    September 1996.
        For the Nuclear Regulatory Commission
    Steven A. Varga,
    Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
    Regulation
    [Doc. 96-24413 Filed 9-24-96; 8:45 am]
    BILLING CODE 7590-01-F
    
    
    

Document Information

Effective Date:
9/11/1996
Published:
09/25/1996
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
X96-10925
Dates:
September 11, 1996
Pages:
50338-50351 (14 pages)
PDF File:
x96-10925.pdf