X95-10830. Biweekly Notice  

  • [Federal Register Volume 60, Number 168 (Wednesday, August 30, 1995)]
    [Notices]
    [Pages 45172-45196]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: X95-10830]
    
    
    
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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 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 4, 1995, through August 18, 1995. The 
    last biweekly notice was published on August 16, 1995 (60 FR 42597).
    
    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 Rules Review and 
    Directives Branch, Division of Freedom of Information and Publications 
    Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
    Washington, DC 20555, 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 September 29, 1995, 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., 
    
    [[Page 45173]]
    Washington, DC and at the local public document room for the particular 
    facility involved. If a request for a hearing or petition for leave to 
    intervene is filed by the above date, the Commission or an Atomic 
    Safety and Licensing Board, designated by the Commission or by the 
    Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
    the request and/or petition; and the Secretary or the designated Atomic 
    Safety and Licensing Board will issue a notice of a hearing or an 
    appropriate order.
        As required by 10 CFR 2.714, a petition for leave to intervene 
    shall set forth with particularity the interest of the petitioner in 
    the proceeding, and how that interest may be affected by the results of 
    the proceeding. The petition should specifically explain the reasons 
    why intervention should be permitted with particular reference to the 
    following factors: (1) the nature of the petitioner's right under the 
    Act to be made a party to the proceeding; (2) the nature and extent of 
    the petitioner's property, financial, or other interest in the 
    proceeding; and (3) the possible effect of any order which may be 
    entered in the proceeding on the petitioner's interest. The petition 
    should also identify the specific aspect(s) of the subject matter of 
    the proceeding as to which petitioner wishes to intervene. Any person 
    who has filed a petition for leave to intervene or who has been 
    admitted as a party may amend the petition without requesting leave of 
    the Board up to 15 days prior to the first prehearing conference 
    scheduled in the proceeding, but such an amended petition must satisfy 
    the specificity requirements described above.
        Not later than 15 days prior to the first prehearing conference 
    scheduled in the proceeding, a petitioner shall file a supplement to 
    the petition to intervene which must include a list of the contentions 
    which are sought to be litigated in the matter. Each contention must 
    consist of a specific statement of the issue of law or fact to be 
    raised or controverted. In addition, the petitioner shall provide a 
    brief explanation of the bases of the contention and a concise 
    statement of the alleged facts or expert opinion which support the 
    contention and on which the petitioner intends to rely in proving the 
    contention at the hearing. The petitioner must also provide references 
    to those specific sources and documents of which the petitioner is 
    aware and on which the petitioner intends to rely to establish those 
    facts or expert opinion. Petitioner must provide sufficient information 
    to show that a genuine dispute exists with the applicant on a material 
    issue of law or fact. Contentions shall be limited to matters within 
    the scope of the amendment under consideration. The contention must be 
    one which, if proven, would entitle the petitioner to relief. A 
    petitioner who fails to file such a supplement which satisfies these 
    requirements with respect to at least one contention will not be 
    permitted to participate as a party.
        Those permitted to intervene become parties to the proceeding, 
    subject to any limitations in the order granting leave to intervene, 
    and have the opportunity to participate fully in the conduct of the 
    hearing, including the opportunity to present evidence and cross-
    examine witnesses.
        If a hearing is requested, the Commission will make a final 
    determination on the issue of no significant hazards consideration. The 
    final determination will serve to decide when the hearing is held.
        If the final determination is that the amendment request involves 
    no significant hazards consideration, the Commission may issue the 
    amendment and make it immediately effective, notwithstanding the 
    request for a hearing. Any hearing held would take place after issuance 
    of the amendment.
        If the final determination is that the amendment request involves a 
    significant hazards consideration, any hearing held would take place 
    before the issuance of any amendment.
        A request for a hearing or a petition for leave to intervene must 
    be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
    Commission, Washington, DC 20555, 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, 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.
    
    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
    50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
    Units Nos. 1, 2, and 3, Maricopa County, Arizona
    
        Date of amendments request: August 3, 1995
        Description of amendments request: The proposed amendment changes 
    would add the analytical method supplement entitled ``Fuel Rod Maximum 
    Allowable Gas Pressure,'' CEN-372-P-A, dated May 1990, and its 
    associated Nuclear Regulatory Commission Safety Evaluation Report, 
    dated April 10, 1990, to the list of analytical methods in TS 6.9.1.10 
    used to determine the PVNGS core operating limits.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        1. The proposed change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        The proposed change does not involve any change to the 
    configuration or method of operation of any plant equipment that is 
    used to mitigate the consequences of an accident. The proposed 
    change adds an NRC approved methodology and its associated Safety 
    Evaluation Report (SER), to the list of analytical methods used to 
    determine the core operating limits. The use of this methodology 
    ensures that the consequences of an accident remain within the 
    limits established by existing analyses. They do not alter any of 
    the assumptions or bounding conditions currently in the UFSAR.
        The U3C6 ECCS performance analysis included the analysis of the 
    impact of the maximum calculated fuel rod gas pressures on the 
    timing of cladding rupture and on the peak cladding temperature. 
    This analysis concluded that the peak cladding temperature for Cycle 
    6 remained below that of the analysis of record and that the peak 
    cladding temperature continued to occur at 
    
    [[Page 45174]]
    low burnup, specifically the burnup corresponding to the maximum 
    initial fuel stored energy.
        In addition to the LOCA analysis a DNB propagation analysis was 
    performed to demonstrate that DNB propagation does not occur during 
    postulated accidents that experience DNB when pressure in a fuel pin 
    is higher than the system pressure. This analysis was performed 
    using the fuel rod strain model described in CEN-372-P-A.
        Based on these analyses, there is no increase in the probability 
    or consequences of an accident previously evaluated.
        2. The proposed change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The proposed change does not involve any change to the 
    configuration or method of operation of any plant equipment that is 
    used to mitigate the consequences of an accident. Accordingly, no 
    new failure modes have been defined for any plant system or 
    component important to safety nor has any new limiting failure been 
    identified as a result of the proposed change. The intent of the 
    proposed change is to utilize a new analytical method to ensure that 
    the consequences of any equipment malfunction remain within the 
    limits of existing analyses resulting in no impact on radiological 
    consequences.
        The impact of the maximum fuel rod gas pressures calculated for 
    U3C6 was evaluated as part of the Cycle 6 ECCS performance analysis. 
    Except for the highest burnup analyzed, the time of cladding rupture 
    decreased as the initial fuel rod gas pressure increased with 
    burnup. However, the peak cladding temperature occurred at the 
    burnup with the maximum initial fuel stored energy. The analysis 
    also determined that the ECCS performance analysis for U3C6 is 
    bounded by that of the reference cycle analysis.
        An evaluation was conducted to ensure that fuel would not 
    experience DNB propagation when the pressure in a fuel pin is higher 
    than the system pressure. DNB was shown not to propagate by 
    demonstrating that the degree of cladding deformation is no more 
    than the limit defined by the fuel rod maximum pressure Topical 
    Report (CEN-372-P-A).
        Therefore, it can be concluded that the proposed change to 
    Section 6.9.1.10 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 adds an NRC approved Topical Report 
    (methodology) and its associated SER, to the list of analytical 
    methods used to determine core operating limits. The use of the new 
    methodology ensures that safety margins are maintained within the 
    results of existing calculations. Since the core operating limits 
    will continue to be established by an NRC approved methodology and 
    will provide adequate core protection, the proposed amendment does 
    not involve a significant reduction in the margin of safety.
        Analyses were conducted to determine the impact of higher fuel 
    rod pressure on ECCS performance and DNB propagation. The results of 
    the analyses show that the effects of higher fuel rod pressure are 
    bounded by previous results.
        The NRC staff has reviewed the licensee's analysis and, based on 
    that review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendments request involve no significant hazards consideration. Local 
    Public Document Room location: Phoenix Public Library, 1221 N. Central 
    Avenue, Phoenix, Arizona 85004
        Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
    and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
    Station 9068, Phoenix, Arizona 85072-3999
        NRC Project Director: William H. Bateman
    
    Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
    Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
    County, Maryland
    
        Date of amendments request: July 13, 1995
        Description of amendments request: The proposed amendments would 
    revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, 
    Technical Specifications (TSs) Section 5.2.1, ``Fuel Assemblies.'' The 
    current TSs only allow fuel that is clad with either zircaloy or ZIRLO. 
    The proposed change would allow the use of cladding material other than 
    zircaloy or ZIRLO with an approved exemption. Thus, the proposed change 
    will eliminate the need for future amendments to allow the use of 
    different cladding material for which the Commission has issued an 
    exemption.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        1. Would not involve a significant increase in the probability 
    or consequences of an accident previously evaluated.
        Calvert Cliffs Technical Specification 5.2.1, Fuel Assemblies, 
    states that fuel rods are clad with either zircaloy or ZIRLO. This 
    reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR [Part] 
    50, Appendix K, which also restrict fuel rod cladding materials to 
    zircaloy or ZIRLO. Baltimore Gas and Electric Company proposes to 
    insert fuel assemblies into Calvert Cliffs Unit 1 which have some 
    fuel rods clad in zirconium alloys that do not meet the definition 
    of zircaloy or ZIRLO for testing purposes and has applied for an 
    exemption to the regulations to allow that change. The proposed 
    change to the Calvert Cliffs Technical Specifications will allow the 
    use of cladding materials that are not zircaloy or ZIRLO with an 
    approved exemption in accordance with 10 CFR 50.12.
        The proposed change to the Unit 1 and Unit 2 Technical 
    Specifications will allow the use of fuel rod cladding materials 
    other than zircaloy or ZIRLO as long as those materials have been 
    approved by an exemption to the regulations. To obtain approval of 
    new cladding materials, 10 CFR 50.12 requires that the applicant 
    show that the proposed exemption is authorized by law, is consistent 
    with the common defense and security, will not present an undue risk 
    to the public health and safety; and is accompanied by special 
    circumstances.
        Under the proposed change, any fuel rod cladding materials that 
    are not zircaloy or ZIRLO must still be approved by the Nuclear 
    Regulatory Commission (NRC) prior to use under 10 CFR 50.12. This 
    change to the Technical Specifications allows the NRC to approve the 
    use of cladding materials that are not either zircaloy or ZIRLO 
    under 10 CFR 50.12 and not require an additional approval under 10 
    CFR 50.90. As such, the proposed change eliminates a duplicative 
    regulatory requirement and would have no effect on the probability 
    or consequences of an accident.
        Therefore, the proposed change does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        2. Would not create the possibility of a new or different type 
    of accident from any accident previously evaluated.
        The proposed change eliminates a duplicated approval requirement 
    and would have no effect on the possibility of a new or different 
    type of accident. The proposed change to the Technical 
    Specifications would allow the NRC to approve the use of fuel rod 
    cladding materials that are not either zircaloy or ZIRLO under 10 
    CFR 50.12 and not require an additional approval under 10 CFR 50.90.
        Therefore, the proposed change does not create the possibility 
    of a new or different type of accident from any accident previously 
    evaluated.
        3. Would not involve a significant reduction in a margin of 
    safety.
        The proposed change eliminates a duplicated approval requirement 
    and will have no effect on the margin of safety. The proposed change 
    to the Technical Specifications would allow the NRC to approve the 
    use of fuel rod cladding materials that are not either zircaloy or 
    ZIRLO under 10 CFR 50.12, and not require an additional approval 
    under 10 CFR 50.90.
        Therefore, the proposed change does not involve a significant 
    reduction in a margin 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 
    amendments request involves no significant hazards consideration.
        Local Public Document Room location: Calvert County Library, Prince 
    Frederick, Maryland 20678
        Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
    Potts and 
    
    [[Page 45175]]
    Trowbridge, 2300 N Street, NW., Washington, DC 20037
        NRC Project Director: Ledyard B. Marsh
    
    Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
    Braidwood Station, Units 1 and 2, Will County, Illinois Docket Nos. 
    STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle 
    County, Illinois Docket Nos. 50-237 and 50-249, Dresden Nuclear 
    Power Station,Units 2 and 3, Grundy County, Illinois Docket Nos. 
    50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle 
    County, Illinois Docket Nos. 50-254 and 50-265, Quad Cities Nuclear 
    Power Station, Units 1 and 2, Rock Island County, Illinois Docket 
    Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2, 
    Lake County, Illinois
    
        Date of application for amendment requests: April 24, 1995
        Description of amendment requests: The licensee proposes to amend 
    Section 6 of the Technical Specifications of all ComEd stations to make 
    the following changes: (1) delete the ``Review, Investigative and Audit 
    Functions'' sections, in their entirety, and relocate these 
    requirements to appropriate sections of the ComEd Quality Assurance 
    Topical Report, (2) change titles to reflect the reorganization of 
    ComEd's Nuclear Operations Division, and (3) miscellaneous 
    administrative and editorial changes.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        A. The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident previously evaluated.
        (1) The proposed relocation of the ``Review, Investigative and 
    Audit Functions'' sections of Technical Specifications to the QA 
    Topical Report does not affect any accident initiators or 
    precursors, and does not change or alter the design assumptions for 
    the systems and components used to mitigate the consequences of an 
    accident.
        The relocation of these sections is consistent with the 
    recommended changes specified in the October 25, 1993 letter from W. 
    T. Russell (USNRC) to the Chairpersons of the Owner Groups' 
    Technical Specifications Committees, entitled, ``Content of Standard 
    Technical Specifications, Section 5.0, Administrative Controls''.
        Relocating these requirements to the QA Topical Report will 
    continue to ensure that proposed future changes to these 
    requirements will receive proper regulatory oversight. NRC review of 
    the Quality Assurance Program is governed by 10CFR50.54. 
    10CFR50.54(a)(3) states: ``Changes to the quality assurance program 
    description that do not reduce the commitments must be submitted to 
    the NRC in accordance with the requirements of 50.71. Changes to the 
    quality assurance program description that do reduce the commitments 
    must be submitted to NRC and receive NRC approval prior to 
    implementation, ...'' Based on these 10CFR50.54 requirements, 
    appropriate licensee and regulatory control of the requirements in 
    the subject relocated Technical Specification sections will be 
    maintained.
        (2) The proposed title and organizational changes to Section 6 
    of Technical Specifications do not affect any accident initiators or 
    precursors and do not change or alter the design assumptions for the 
    systems or components used to mitigate the consequences of an 
    accident.
        Commonwealth Edison's organizational changes allow for increased 
    senior management attention and oversight of station activities. 
    Position titles and associated responsibilities have changed to 
    increase the company's efficiency in the management of its nuclear 
    stations. These administrative changes do not reduce any 
    requirements or commitments. The proposed changes enhance the 
    administrative controls necessary to ensure safe plant operation.
        (3) Other proposed administrative/editorial changes simply make 
    corrections or provide needed clarification prompted by the 
    reorganization. These changes provide consistency with station 
    procedures, programs, other Technical Specifications, and Standard 
    Technical Specifications. They are administrative in nature and do 
    not impact any accident previously evaluated in the UFSAR.
        In conclusion, none of the proposed changes involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        B. The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        (1) The proposed relocation of the ``Review, Investigative and 
    Audit Functions'' sections of Technical Specifications to the QA 
    Topical Report does not affect the design or operation of any 
    system, structure, or component in the plant. There are no changes 
    to parameters governing plant operation and no new or different type 
    of equipment will be installed that could give rise to a new or 
    different kind of accident that was previously evaluated.
        The proposed changes are considered to be administrative or 
    programmatic in nature and do not affect equipment or components 
    that could initiate an accident. All administrative commitments 
    being relocated to the QA Topical Report will continue to receive 
    appropriate regulatory oversight pursuant to 10CFR50.54.
        (2) The proposed title and organization changes do not affect 
    the design or operation of any system, structure, or component in 
    the plant. There are no changes to parameters governing plant 
    operation; no new or different type of equipment will be installed. 
    The proposed changes are considered to be administrative changes 
    that will enhance the performance of organizations responsible for 
    the safe operation of the plant to respond to plant transients or 
    emergencies. All responsibilities described in Technical 
    Specifications for management activities will continue to be 
    performed by qualified individuals.
        (3) All other proposed changes are administrative in nature and 
    do not create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        In conclusion, the proposed changes do not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        C. The proposed changes do not involve a significant reduction 
    in a margin of safety.
        (1) The proposed changes are administrative or programmatic in 
    nature and do not affect the margin of safety for any safety 
    parameters and setpoints addressed in Technical Specifications. The 
    assumptions, initial conditions and methodologies used in the 
    accident analyses remain unchanged, therefore, accident analyses 
    results are not impacted.
        Placing these requirements in QA Topical Report will continue to 
    ensure that proposed future changes to these requirements will 
    receive proper regulatory oversight pursuant to 10CFR50.54.
        (2) The proposed title and organizational changes are 
    administrative in nature and do not affect the margin of safety for 
    any Technical Specification. The initial conditions and 
    methodologies used in the accident analyses remain unchanged, 
    therefore, accident analyses results are not impacted.
        (3) All other proposed changes are administrative in nature and 
    have no impact on the margin of safety for any Technical 
    Specification.
        In conclusion, the proposed changes do not involve a significant 
    reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: for Braidwood, the Wilmington 
    Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481; for 
    Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box 
    434, Byron, Illinois 61010; for Dresden, Morris Area Public Library 
    District, 604 Liberty Street, Morris, Illinois 60450; for LaSalle, 
    Jacobs Memorial Library, Illinois Valley Community College, Oglesby, 
    Illinois 61348; for Quad Cities, Dixon Public Library, 221 Hennepin 
    Avenue, Dixon, Illinois 61021; for Zion, Waukegan Public Library, 128 
    N. County Street, Waukegan, Illinois 60085
    
    [[Page 45176]]
    
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60690
        NRC Project Director: Robert A. Capra
    
    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 Docket Nos. 50-373 and 50-374, LaSalle County 
    Station, Units 1 and 2, LaSalle County, Illinois
    
        Date of amendment request: June 8, 1995
        Description of amendment request: The proposed amendments would 
    revise Technical Specifications Section 3/4.8, Electrical Power 
    Systems, and the associated Bases for LaSalle County, Byron, and 
    Braidwood Stations. The proposed changes revise surveillance and 
    administrative requirements associated with emergency diesel generators 
    (EDGs) in accordance with the guidance of NRC Generic Letter 94-01, 
    ``Removal of Accelerated Testing and Special Reporting Requirements for 
    Emergency Diesel Generators,'' Generic Letter 93-05, ``Line-Item 
    Technical Specifications Improvements to Reduce Surveillance 
    Requirements for Testing During Power Operation,'' and Regulatory Guide 
    (RG) 1.9, ``Selection, Design, Qualification, and Testing of Emergency 
    Diesel Generator Units Used as Class 1E Onsite Electric Power Systems 
    at Nuclear Power Plants.'' The proposed changes include: (1) 
    eliminating increased testing requirements for EDGs, (2) eliminating 
    special reporting requirements for EDGs, (3) eliminating the semi-
    annual fast load test and replacing it with a requirement to load EDGs 
    semi-annually in accordance with the vendor recommendations for all 
    test purposes other than the refueling outage Loss of Offsite Power 
    (LOOP) tests, (4) de-coupling the 24-hour endurance run and the LOOP/
    loss-of-coolant (LOCA) (LOOP only for LaSalle) sequencing requirements 
    for the hot start test, (5) removing RG 1.108 references to testing 
    requirements, (6) eliminating testing requirements when an EDG becomes 
    inoperable due to an inoperable support system, an independently 
    testable component, or preplanned maintenance or testing, or if there 
    is not a potential common mode failure for the remaining diesel 
    generator, (7) deleting the requirement for inspecting the EDGs in 
    accordance with procedures prepared in conjunction with its 
    manufacturer's recommendations, and (8) making editorial changes.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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:
        The proposed changes do not affect accident initiators or 
    precursors and do not alter the design assumptions affecting the 
    ability of the EDGs to mitigate the consequences of an accident.
        Deleting the special reporting requirements from the Technical 
    Specifications is administrative. ComEd will continue to notify the 
    Commission of significant EDG failures in accordance with 10 CFR 
    50.72 and 50.73 criteria.
        Excessive testing requirements have proven to be a contributor 
    to increased equipment degradation. Removing inappropriate and 
    redundant requirements increases EDG reliability and enhances the 
    ability of EDGs to mitigate the consequences of an accident. 
    Implementing ComEd's alternative to the maintenance rule for the 
    EDGs provides additional assurance that high EDG performance will be 
    maintained.
        EDG equipment degradation will be reduced by eliminating the 
    semi-annual fast load test for EDGs in accordance with the vendor 
    recommendations for test purposes other than the refueling outage 
    Loss of Offsite Power (LOOP) tests. This improves EDG reliability 
    and availability and further enhances their ability to mitigate the 
    consequences of an accident. The LOOP test would still be performed 
    to provide assurance that the EDG is capable of responding to a LOOP 
    as assumed in the accident analyses.
        De-coupling the 24 hour endurance test and the LOOP/LOCA (for 
    LaSalle, LOOP) sequencing test requirements for the hot start test 
    has no effect on accident mitigation. Demonstrating diesel generator 
    hot restart capability without loading the engine does not 
    invalidate or reduce the effectiveness of the hot restart test. The 
    hot restart test can be conducted in any plant condition since its 
    performance at power will have no adverse effect on plant 
    operations.
        The proposed editorial changes are administrative in nature. 
    They improve readability and provide consistency with current 
    industry guidance.
        Therefore, the proposed changes do not involve an increase in 
    the probability or consequences of an accident previously evaluated.
        2) Create the possibility of a new or different kind of accident 
    from any accident previously evaluated:
        The proposed changes do not alter the ability of the EDGs to 
    perform their intended function to mitigate the consequences of an 
    initiating event within the acceptance limits assumed in plant 
    safety analyses. The proposed changes have no impact on component or 
    system interactions, or the plant design basis.
        Instrumentation setpoints, starting, sequencing, and loading 
    functions associated with EDGs are not affected by the proposed 
    changes. Furthermore, combining the alternate EDG system maintenance 
    rule implementation program with the proposed amendment will enhance 
    both the availability and the performance of the EDGS.
        Therefore, there is not a potential for creating the possibility 
    of a new or different type of accident from any accident previously 
    evaluated.
        3) Involve a significant reduction in a margin of safety:
        The proposed changes do not increase the probability or 
    consequences of an accident, and there is no impact on equipment 
    design or operation. The proposed changes do not affect the results 
    of accident and transient analyses. Plant and system response to an 
    initiating event will remain in compliance within the assumptions of 
    safety analyses. There is no associated change to the type, amount, 
    or control of radioactive effluents, nor is there an associated 
    increase in individual or cumulative occupational radiation 
    exposure. There is no effect upon the capabilities of the associated 
    systems to perform their intended functions within the allowed 
    response times assumed in safety analyses.
        The proposed changes are compatible with plant operating 
    experience and are consistent with the guidance provided in NUREG-
    1366, Generic Letters 93-05 and 94-01, and Regulatory Guide 1.9. In 
    two instances ComEd's proposed changes deviate from these guidance 
    documents. However, the changes are consistent with the intent of 
    the documents or other NRC guidance documents. Eliminating excessive 
    testing requirements can improve safety by reducing challenges to 
    plant systems and reducing equipment wear and degradation. While the 
    proposed changes affect surveillance intervals; there are no changes 
    to the methods used to perform the surveillances.
        EDG reliability and availability will be improved by the 
    proposed changes. The surveillances will continue to demonstrate the 
    ability of the EDGs to perform their intended function of providing 
    electrical power to the emergency safety systems needed to mitigate 
    design basis transients. No margin of safety is reduced.
        Guidance has been provided in ``Final Procedures and Standards 
    on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
    for the application of standards to license change requests for 
    determination of the existence of significant hazards 
    considerations. This document provides examples of amendments which 
    are and are not considered likely to involve significant hazards 
    considerations. These proposed amendments most closely fit the 
    example of a change which may either result in some increase to the 
    probability or consequences of a previously analyzed accident or may 
    reduce in some way a safety margin, but where the results of the 
    change are clearly within all acceptance criteria with respect to 
    the system or component specified in the standard review plan.
        This proposed amendment does not involve a significant 
    relaxation of the criteria used to establish safety limits, a 
    significant 
    
    [[Page 45177]]
    relaxation of the bases for the limiting safety system settings, or a 
    significant relaxation of the bases for the limiting conditions for 
    operations. The proposed change does not reduce the margin of safety 
    as defined in the basis for any Technical Specification.
        Therefore, based on the guidance provided in the Federal 
    Register and the criteria established in 10 CFR 50.92(c), ComEd has 
    concluded that the proposed change does not constitute a significant 
    hazards consideration.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    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; for LaSalle, 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. 50-237 and 50-249, Dresden 
    Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
    Units 1 and 2, Rock Island County, Illinois
    
        Date of application for amendment requests:  August 30, 1994, as 
    supplemented August 4, 1995.
        Description of amendment requests: As a result of findings by a 
    Diagnostic Evaluation Team inspection performed by the NRC staff at the 
    Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
    (ComEd, the licensee) made a decision that both the Dresden Nuclear 
    Power Station and sister site Quad Cities Nuclear Power Station needed 
    attention focused on the existing custom Technical Specifications (TS) 
    used.
        The licensee made the decision to initiate a Technical 
    Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
    The licensee evaluated the current TS for both Dresden and Quad Cities 
    against the Standard Technical Specifications (STS) contained in NUREG-
    0123, ``Standard Technical Specifications General Electric Plants BWR/
    4.'' The licensee's evaluation identified numerous potential 
    improvements such as clarifying requirements, changing TS to make them 
    more understandable and to eliminate interpretation, and deleting 
    requirements that are no longer considered current with industry 
    practice. As a result of the evaluation, ComEd has elected to upgrade 
    both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
        The TSUP for Dresden and Quad Cities is not a complete adaption of 
    the STS. The TSUP focuses on (1) integrating additional information 
    such as equipment operability requirements during shutdown conditions, 
    (2) clarifying requirements such as limiting conditions for operation 
    and action statements utilizing STS terminology, (3) deleting 
    superseded requirements and modifications to the TS based on the 
    licensee's responses to Generic Letters (GL), and (4) relocating 
    specific items to more appropriate TS locations.
        The August 30, 1994, and August 4, 1995, applications proposed to 
    upgrade only Section 3/4.2 (Instrumentation) of the Dresden and Quad 
    Cities TS.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        1) The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident previously evaluated 
    because:
        In general, the proposed amendment represents the conversion of 
    current requirements to a more generic format, or the addition of 
    requirements which are based on the current safety analysis. 
    Implementation of these changes will provide increased reliability 
    of equipment assumed to operate in the current safety analysis, or 
    provide continued assurance that specified parameters remain within 
    their acceptance limits, and as such, will not significantly 
    increase the probability or consequences of a previously evaluated 
    accident.
        Some of the proposed changes to the current Technical 
    Specifications (CTS) represent minor curtailments of the current 
    requirements which are based on generic guidance or previously 
    approved provisions for other stations. The proposed amendment for 
    Dresden and Quad Cities Station's Technical Specification Section 3/
    4.2 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard 
    Technical Specifications General Electric Plants BWR/4) guidance or 
    NRC accepted changes at later operating BWR plants. Any deviations 
    from BWR-STS and CTS requirements do not significantly increase the 
    probability or consequences of any previously evaluated accident for 
    Dresden and Quad Cities Station. These proposed changes are 
    consistent with the current safety analyses and have been previously 
    determined to represent sufficient requirements for the assurance 
    and reliability of equipment assumed to operate in the safety 
    analysis, or provide continued assurance that specified parameters 
    remain within their acceptance limits. As such, these changes will 
    not significantly increase the probability or consequences of a 
    previously evaluated accident.
        The associated systems that make up the Instrumentation Systems 
    are not assumed in any safety analysis to initiate any accident 
    sequence for both Dresden and Quad Cities Stations; therefore, the 
    probability of any accident previously evaluated is not increased by 
    the proposed amendment. In addition, the proposed surveillance 
    requirements for the proposed amendments to these systems are 
    generally more prescriptive than the current requirements specified 
    within the Technical Specifications. These more prescriptive 
    surveillance requirements increase the probability that the 
    Instrumentation Systems will perform their intended functions. 
    Therefore, the proposed TS will improve the reliability and 
    availability of all affected systems and reduce the consequences of 
    any accident previously evaluated.
        2. Create the possibility of a new or different kind of accident 
    from any previously evaluated because:
        In general, the proposed amendment represents the conversion of 
    current requirements to a more generic format, or the addition of 
    requirements which are based on the current safety analysis. Others 
    represent minor curtailments of the current requirements which are 
    based on generic guidance or previously approved provisions for 
    other stations. These changes do not involve revisions to the design 
    of the station, other than technically valid trip setpoint changes. 
    Some of the changes may involve revision in the operation of the 
    station; however, these changes provide additional restrictions 
    which are in accordance with the current safety analyses, or are to 
    provide for additional testing or surveillances which will not 
    introduce new failure mechanisms beyond those already considered in 
    the current safety analyses. Therefore, these changes will not 
    create the possibility of a new or different kind of accident from 
    any accident previously evaluated.
        The proposed amendment for Dresden and Quad Cities Station's 
    Technical Specification Section 3/4.2 is based on BWR-STS guidelines 
    or NRC accepted changes at later operating BWR plants. The proposed 
    amendment has been reviewed for acceptability at the Dresden and 
    Quad Cities Nuclear Power Stations considering similarity of system 
    or component design versus the BWR-STS or later operating BWRs. Any 
    deviations from BWR-STS or CTS requirements do not create the 
    possibility of a new or different kind of accident than previously 
    evaluated for Dresden and Quad Cities Stations. No new modes of 
    operation are introduced by the proposed changes. Various 
    surveillance requirements are changed to reflect improvements in 
    technique, frequency of performance or operating experience at later 
    plants. Proposed changes to action statements in many places add 
    requirements that are not in the present technical specifications or 
    adopt 
    
    [[Page 45178]]
    requirements that have been used at other operating BWRs with designs 
    similar to Dresden and Quad Cities. The proposed changes maintain at 
    least the present level of operability. Therefore, the proposed 
    changes do not create the possibility of a new or different kind of 
    accident from any previously evaluated.
        The associated systems that make up the Instrumentation Systems 
    are not assumed in any safety analysis to initiate any accident 
    sequence for Dresden or Quad Cities Stations. In addition, the 
    proposed surveillance requirements for affected systems associated 
    with the Instrumentation Systems are generally more prescriptive 
    than the current requirements specified within the Technical 
    Specifications; therefore, the proposed changes do not create the 
    possibility of a new or different kind of accident from any 
    previously evaluated.
        3. Involve a significant reduction in the margin of safety 
    because:
        In general, the proposed amendment represents the conversion of 
    current requirements to a more generic format, or the addition of 
    requirements which are based on the current safety analysis. Others 
    represent minor curtailments of the current requirements which are 
    based on generic guidance or previously approved provisions for 
    other stations. Some of the later individual items may introduce 
    minor reductions in the margin of safety when compared to the 
    current requirements. However, other individual changes are the 
    adoption of new requirements which will provide significant 
    enhancement of the reliability of the equipment assumed to operate 
    in the safety analysis, or provide enhanced assurance that specified 
    parameters remain within their acceptance limits. These enhancements 
    compensate for the individual minor reductions, such that taken 
    together, the proposed changes will not significantly reduce the 
    margin of safety.
        The proposed amendment to Technical Specification Section 3/4.2 
    implements present requirements in accordance with the guidelines 
    set forth in the BWR-STS. Any deviations from BWR-STS and CTS 
    requirements do not significantly reduce the margin of safety for 
    Dresden and Quad Cities Stations. The proposed changes are intended 
    to improve readability, usability, and the understanding of 
    technical specification requirements while maintaining acceptable 
    levels of safe operation. The proposed changes have been evaluated 
    and found to be acceptable for use at Dresden and Quad Cities based 
    on system design, safety analysis requirements and operational 
    performance. Since the proposed changes are based on NRC accepted 
    provisions at other operating plants that are applicable at Dresden 
    and Quad Cities and maintain necessary levels of system or component 
    readability, the proposed changes do not involve a significant 
    reduction in the margin of safety.
        The proposed amendment for Dresden and Quad Cities Stations will 
    not reduce the availability of systems associated with the 
    Instrumentation Systems when required to mitigate accident 
    conditions; therefore, 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 requests involve no significant hazards consideration.
        Local Public Document Room location: for Dresden, Morris Public 
    Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
    Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60690
        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: March 8, 1995, as supplemented June 1, 
    1995
        Description of amendment request: The proposed amendments would 
    revise the secondary undervoltage setpoint.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 an increase in the 
    probability of occurrence or consequences of any accident previously 
    evaluated.
        The proposed amendment does not change the fundamental function 
    or capability of the Secondary Undervoltage protection as described 
    in UFSAR section 8.3. Inadvertent or spurious operation of the 
    Secondary Undervoltage protection function will initiate loading of 
    the safe shutdown loads on the diesel generators and is not assumed 
    to initiate an accident. The proposed Secondary Undervoltage 
    setpoints are low enough to prevent spurious actuations given the 
    expected off site grid voltages.
        This change does not affect the initiators or precursors of any 
    accident previously evaluated. This change will not increase the 
    likelihood that a transient initiating event will occur because 
    transients are initiated by equipment malfunction and/or 
    catastrophic system failure. The change in setpoints for the 
    Secondary Undervoltage protection system does involve some changes 
    to existing plant equipment (such as transformer tap changes and 
    Circulating Water pump excitation circuit changes). However, all 
    changes to existing plant equipment have been or will be evaluated 
    in accordance with the requirements of 10CFR50.59 prior to 
    installation, to determine that no unreviewed safety questions exist 
    with regard to the plant changes.
        Since any design changes have been or will be determined to be 
    acceptable per 10CFR50.59 prior to installation and no new plant 
    equipment will be installed, the probability of occurrence of 
    accidents previously evaluated will not increase.
        With Zion Station's new Auxiliary Power System configuration and 
    the proposed Secondary Undervoltage setpoints, the probability of a 
    Loss of Off-Site Power (LOOP) is actually reduced since the original 
    Auxiliary Power System configuration and Secondary Undervoltage 
    setpoints required a higher grid voltage to ensure that safety 
    related loads would be powered from Off-Site power sources during a 
    design basis accident.
        The consequences of accidents previously evaluated are not 
    increased. The proposed change does not affect the required level of 
    availability or systems required to mitigate the accidents 
    considered in the Analyses. Administrative controls will be in place 
    to ensure that the installed setpoints are low enough to ensure that 
    the Emergency Diesel Generators are not unnecessarily challenged. 
    The proposed changes will increase the level of confidence that the 
    ESF equipment will be capable of starting and operating during a 
    design basis accident with degraded off-site grid voltage. The 
    increase in the level of confidence is the result of the more 
    rigorous methodology used to determine limited ESF bus voltages, 
    given the minimum expected off-site AC voltage. Based on the 
    previous discussion, it is determined that there will be no 
    significant increase in the consequences of any accident previously 
    evaluated.
        2. The proposed amendment does not create the possibility of a 
    new or different kind of accident from any previously analyzed.
        The proposed Secondary Undervoltage setpoint change does not 
    change the design of the Secondary Undervoltage protection system or 
    its function to protect against degraded offsite power. Actuation of 
    the Secondary Undervoltage protection system will initiate a 
    sequence of events that will start the Emergency Diesel Generator 
    (EDG) for the associated ESF bus, strip all loads from the bus, open 
    all feed breakers to the bus, close the Emergency feed breaker (thus 
    energizing the bus from the EDG), and initiate sequenced starting of 
    the Safe Shutdown equipment supplied by the bus, including a Service 
    Water pump, Component Cooling Water pump, Auxiliary Feedwater pump, 
    and Reactor Containment Fan Cooler(s), as applicable.
        The proposed change does not involve the addition of any new or 
    different types of equipment, nor does it involve the operation of 
    equipment required for safe operation of the facility in a manner 
    different from those addressed in the Final Safety Analysis Report. 
    No safety related equipment or function will be altered as a result 
    of this proposed change. Because no new failure modes are 
    introduced, the proposed amendment does not create a new or 
    different kind of accident from any previously analyzed in the 
    UFSAR.
        Based on the above discussion, the proposed amendment does not 
    create a new or different kind of accident from any previously 
    analyzed in the UFSAR.
        3. The proposed changes do not involve a significant reduction 
    in a margin of safety.
    
    [[Page 45179]]
    
        The proposed amendment will allow the Secondary Undervoltage 
    setpoint to be conservatively established based on new engineering 
    calculations which consider the lowest expected offsite grid voltage 
    and operation of all required ESF equipment under design basis 
    accident loading conditions.
        The proposed Secondary Undervoltage setpoints will provide 
    increased confidence that adequate bus voltage will be available to 
    support starting and operation of all required ESF loads. The 
    proposed setpoint includes worst case instrument error to ensure 
    that the lowest possible voltage will not be lower than the degraded 
    voltage analytical limits. Additionally, the proposed setpoints are 
    low enough to prevent spurious actuations due to expected 
    fluctuations in the grid voltage. The new setpoints are based on a 
    minimum expected grid voltage of 343 kV, with added margin. The 
    proposed changes will provide an increase in the level of protection 
    that currently exists and will ensure the margin of safety is 
    adequately maintained.
        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
    
    Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
    Neck Plant, Middlesex County, Connecticut
    
        Date of amendment request: August 3, 1995
        Description of amendment request: The proposed amendment will add 
    an one-time footnote to Technical Specification (TS) Section 3/4.7.12, 
    ``Ultimate Heat Sink,'' to increase the allowed outage time from 6 
    hours to 18 hours for the months of August and September. In addition, 
    also for the months of August and September, the maximum service water 
    limit will be elevated from 90 deg.F to 95 deg.F.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 analyzed.
        The proposed addition of a 12 hour time period to monitor the 
    ultimate heat sink temperature to the Technical Specification 
    Limiting Condition for Operation action statements does not involve 
    an increase in the probability of an accident previously evaluated. 
    The probability of an accident previously evaluated is not increased 
    by a short-term increase in the ultimate heat sink temperature. An 
    evaluation has been performed that safe shutdown will be achieved 
    and maintained for a loss of normal AC power event with the 
    additional consideration of a single failure with service water 
    inlet temperatures as high as 95 deg.F. In addition, an evaluation 
    of the credible FSAR Chapter 15 events with AC power available and 
    no isolation of non-essential service water loads has been performed 
    that demonstrates that safe shutdown will be achieved and 
    maintained. There has been no significant increase in the 
    consequences of these events previously evaluated.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        The proposed technical specification change does not create the 
    possibility of a new or different kind of accident previously 
    analyzed. The addition of a 12 hour time period to monitor the 
    ultimate heat sink temperature increases the amount of time that is 
    allowed for the plant to be in Hot Standby from 6 to 18 hours should 
    the ultimate heat sink temperature increase above 90 deg.F. This 
    extension of the time allowed for the plant to be in Hot Standby 
    does not change the plant configuration. As such, the change does 
    not create the possibility of a new or different kind of accident 
    previously evaluated.
        3. Involve a significant reduction in the margin of safety.
        The proposed technical specification change does not involve a 
    significant reduction in the margin of safety. The addition of a 12 
    hour time period to monitor the ultimate heat sink temperature 
    increases the time required for the plant to be in Hot Standby from 
    6 to 18 hours should the ultimate heat sink temperature exceed 
    90 deg.F. An evaluation has been performed to demonstrate that the 
    risk significance associated with the increased action time is very 
    low. In addition, safe shutdown capability has been demonstrated for 
    service water inlet temperatures as high as 95 deg.F.
        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: Russell Library, 123 Broad 
    Street, Middletown, CT 06457
        Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
    Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
    CT 06141-0270
        NRC Project Director: Phillip F. McKee
    
    Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
    Atomic Power Station, Lincoln County, Maine
    
        Date of amendment request: May 5, 1995
        Description of amendment request: The proposed amendment would 
    change the surveillance frequency of radiation area, and effluent and 
    process monitors from monthly to quarterly; and the required frequency 
    for minimum exercise of control element assemblies also from monthly to 
    quarterly.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration. The 
    NRC staff has reviewed the licensee's analysis against the standards of 
    10 CFR 50.92(c). The staff's review is presented below:
        1. The proposed change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated. Extending surveillance test intervals as proposed will 
    reduce the probability of inadvertent reactor scrams and ensuing 
    challenges to safety systems. This is accomplished by reducing the 
    occasions and thus the total time that the subject systems are 
    removed from their ``normal'' configuration and placed into the 
    required ``test'' configuration. In addition, the probability of 
    test-induced failures, or failures caused by human error, is 
    likewise decreased. Thus, the 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.
        Extending surveillance test intervals as proposed will not 
    require installation of any new or different equipment, and will not 
    alter or otherwise modify existing plant equipment. Thus, 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.
        Independent research has found that equipment failures and 
    personnel errors during several types of surveillance tests caused a 
    significant number of reactor scrams and attendant unnecessary 
    challenges to safety equipment. The results of this research have 
    been corroborated by the licensee's plant specific operating 
    experience. The licensee concludes that the reduced test intervals 
    proposed in this amendment remain sufficient to ensure known 
    phenomena, such as instrument setpoint drift and random hidden 
    failures, remain within the assumptions of the safety analysis. 
    Thus, the proposed change does not involve a significant reduction 
    in a margin of safety.
        Based on this review, it appears that the three standards of 10 CFR 
    50.92(c) 
    
    [[Page 45180]]
    are satisfied. Therefore, the NRC staff proposes to 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 Project Director: Phillip F. McKee
    
    North Atlantic Energy Service Corporation, Docket No. 50-443, 
    Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
    
        Date of amendment request: July 24, 1995.
        Description of amendment request: The proposed amendment would 
    delete Table 3.4-1, ``Reactor Coolant System Pressure Isolation 
    Valves'' from the Seabrook Station, Unit No. 1 Technical Specification 
    section 3.4.6.2. Reference to Table 3.4-1 also would be deleted from 
    Limiting Condition for Operation 3.4.6.2 f and from Surveillance 
    Requirement 4.4.6.2.2. The information contained in Table 3.4-1 would 
    be relocated to the Technical Requirements Manual.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration. The 
    NRC staff has reviewed the licensee's analysis against the standards of 
    10 CFR 50.92(c). The NRC staff's review is presented below.
        A. The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident previously evaluated 
    (10 CFR 50.92(c)(1)) because they do not in any way alter the 
    operability or surveillance requirements for pressure isolation 
    valves. The proposed changes merely delete a listing of valves which 
    are designated as pressure isolation valves in accordance with the 
    definition provided in 10 CFR Part 50. Therefore, neither the 
    probability nor consequences of previously evaluated accidents are 
    affected.
        B. The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously evaluated 
    (10 CFR 50.92(c)(2)) because they do not affect in any way the 
    manner by which the facility is operated or make any changes in 
    structures, systems, or components which could affect the 
    operational characteristics of the facility.
        C. The proposed changes do not involve a significant reduction 
    in a margin of safety (10 CFR 50.92(c)(3)) because the proposed 
    changes do not affect the operability requirements or surveillance 
    testing of any pressure isolation valve and do not affect in any way 
    the manner by which the facility is operated or involve equipment or 
    features which affect the operational characteristics of the 
    facility.
        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: Exeter Public Library, 
    Founders Park, Exeter, NH 03833
        Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
    Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
        NRC Project Director: Phillip F. McKee
    
    Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
    Millstone Nuclear Power Station, Unit 1, New London County, 
    Connecticut
    
        Date of amendment request: July 28, 1995
        Description of amendment request: The proposed amendment adds 
    Technical Specifications (TS) to Section 3.10, Refueling and Spent Fuel 
    Handling. Specifically, the proposed TS (with applicability, action, 
    and surveillance requirements) will require that: (1) the reactor be 
    subcritical for at least 100 hours before the start of reactor 
    refueling operations, (2) the spent fuel pool bulk temperature be 
    maintained less than or equal to 140 deg.F, and (3) two trains of 
    shutdown cooling be operable during reactor refueling operations. In 
    support of the request, NNECO proposes to: (1) use the ORIGEN2 code to 
    more accurately predict decay heat loads from the spent fuel, (2) use 
    the ONEPOOL code to credit the effect of evaporative cooling on the 
    spent fuel pool bulk temperature, and (3) take credit for both trains 
    of shutdown cooling to assist the spent fuel pool cooling system during 
    refueling outages. In addition, the proposed amendment modifies the 
    table of contents and associated Bases section to reflect the changes.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        NNECO has reviewed the proposed changes in accordance with 
    10CFR50.92 and concluded that the changes do not involve a 
    significant hazards consideration (SHC). The basis for this 
    conclusion is that the three criteria of 10CFR50.92(c) are not 
    compromised. The proposed changes do not involve an SHC because the 
    changes would not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously analyzed.
        The proposed license amendment will allow NNECO to use the 
    shutdown cooling system (SCS) to assist the spent fuel pool cooling 
    (SFPC) system to cool the spent fuel pool during refueling outages. 
    This amendment request does not affect: the number of spent fuel 
    bundles allowed in the spent fuel pool, spent fuel pool criticality 
    analysis, structural analysis of the spent fuel pool, or 
    radiological release scenarios.
        The proposed license amendment also allows NNECO to use ORIGEN2 
    and ONEPOOL codes. The ORIGEN2 code more accurately predicts decay 
    heat loads from the spent fuel in the spent fuel pool. The ONEPOOL 
    code credits the effect of evaporative cooling on the spent fuel 
    pool bulk temperature. The use of these codes will improve the 
    accuracy of predicting spent fuel pool bulk temperatures during 
    normal and abnormal refueling scenarios.
        The use of the SCS to assist the SFPC system to cool the spent 
    fuel pool will allow the movement of spent fuel to begin 100 hours 
    after reactor shutdown. The existing accident analysis for a dropped 
    spent fuel bundle during refueling bounds this situation as the 
    analysis assumed a decay time of 24 hours.
        The three new proposed technical specifications will provide 
    sufficient controls on the movement of spent fuel into the spent 
    fuel pool, bulk temperature of the spent fuel pool and operability 
    of the shutdown cooling system to operate within analysis 
    assumptions during refueling operations at Millstone Unit No. 1.
        Therefore, based on the above, the use of the SCS to assist the 
    SFPC system to cool the spent fuel pool during refueling outages, 
    the use of the ORIGEN2 code, the use of the ONEPOOL code, and the 
    addition of three technical specifications will not involve a 
    significant increase in the probability or consequences of an 
    accident previously analyzed.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        The proposed license amendment to use the SCS to assist the SFPC 
    system to cool the spent fuel pool will allow SCS train B to cool 
    the spent fuel pool in a method similar to train A.
        The proposed license amendment to use ORIGEN2 and ONEPOOL codes 
    to predict spent fuel pool bulk temperatures will increase the 
    accuracy of analyzing normal and abnormal refueling scenarios.
        The three new proposed technical specifications will 
    sufficiently control refueling operations to support analyzed 
    accident scenarios.
        Therefore, the use of the SCS to assist the SFPC system to cool 
    the spent fuel pool, the use of the ORIGEN2 code, the use of ONEPOOL 
    code and the addition of three technical specifications do not 
    create the possibility of a new or different kind of accident from 
    any previously analyzed.
        3. Involve a significant reduction in the margin of safety.
        The proposed license amendment to use the SCS to assist the SFPC 
    system to cool the spent fuel pool will allow the crediting of the 
    SCS and SFPC system to remove heat from 
    
    [[Page 45181]]
    the spent fuel pool during normal refueling scenarios. The analysis 
    demonstrates that this cooling configuration will maintain the spent 
    fuel pool bulk temperature below the pool design limit of 140 deg.F 
    with a postulated single active failure.
        The addition of the train B SCS cross-tie does not adversely 
    affect the existing design basis of the SCS to remove sensible and 
    decay heat from the reactor water, cool it from 280 deg.F to 
    125 deg.F within 24 hours, and to maintain the reactor water at 
    125 deg.F.
        The proposed license amendment to use ORIGEN2 and ONEPOOL codes 
    will improve the accuracy of predicting spent fuel pool bulk 
    temperatures during normal and abnormal refueling scenarios.
        The thermal hydraulic analysis most limiting time to boil 
    calculation of 5.4 hours for loss of all forced cooling to the spent 
    fuel pool is consistent with assumed operator response times for 
    similar scenarios.
        The three new proposed technical specifications will ensure that 
    the margin of safety established by engineering analysis of 
    refueling operations is maintained.
        Therefore, based on the above, the use of the SCS to assist the 
    SFPC system to cool the spent fuel pool, the use of the ORIGEN2 
    code, the use of the ONEPOOL code, and the addition of three 
    technical specifications does not involve a significant reduction in 
    the margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Learning Resources Center, 
    Three Rivers Community-Technical College, 574 New London Turnpike, 
    Norwich, CT 06360
        Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
    Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
    CT 06141-0270
        NRC Project Director: Phillip F. McKee
    
    Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-
    336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, 
    and 3, New London, Connecticut
    
        Date of amendment request: August 4, 1995
        Description of amendment request: The proposed license amendments 
    will modify the Administrative Controls Section (Section 6) of the 
    Millstone Unit Nos. 1, 2, and 3 Technical Specifications to allow the 
    Plant Operations Review Committee (PORC) and Site Operations Review 
    Committee (SORC) to direct its efforts in the review of more critical 
    safety matters which affect day-to-day operation. This will be 
    accomplished by the establishment of a Station Qualified Reviewer 
    Program (SQRP) and the reassignment of certain procedure approvals to 
    designated managers in lieu of approval by PORC/SORC.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration 
    (SHC), which is presented below:
        ...These proposed changes do not involve an SHC because the 
    changes do not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        These changes are administrative in nature. They do not involve 
    any modifications to plant systems and do not alter the method of 
    operation of any plant equipment. The change involves the 
    establishment of a SQRP for the review of plant procedures, programs 
    or changes thereto that do not involve a 10CFR50.59 evaluation.
        Implementing a SQRP will not result in a degradation of the 
    current level of procedure review. PORC/SORC will retain the 
    responsibility for reviewing any document for which a 10CFR50.59 
    evaluation is required. Personnel selected to be SQRs [Station 
    Qualified Reviewers] will possess the technical experience and 
    expertise to provide a thorough technical review as required by 
    plant procedures. These personnel, and the managers authorized to 
    approve these procedures, will be designated in writing by the Unit 
    Director or the Senior Vice President - Millstone Station. 
    Procedures or classes of procedures that can be reviewed per the 
    SQRP will be specified in writing by the Unit Director or the Senior 
    Vice President - Millstone Station. Procedures will receive an 
    appropriate cross-disciplinary review when necessary.
        2. Create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        The proposed technical specification changes do not change the 
    design or function of any plant structure, system, or component, nor 
    do they introduce any new failure modes. As stated above, the 
    implementation of a SQRP will not degrade the quality of plant 
    procedures.
        There are no modifications to plant structures, systems, or 
    components associated with these proposed changes, and the operation 
    of plant equipment and systems remain unchanged. Since the changes 
    proposed in this license amendment request do not revise existing 
    plant structures, systems, or components, do not change the manner 
    in which the plant is operated and, do not change the manner in 
    which the plant will respond to any design basis accidents, the 
    proposed changes do not create the possibility of a new or different 
    kind of accident from any previously analyzed.
        3. Involve a significant reduction in a margin of safety.
        The changes proposed in this proposed license amendment request 
    do not affect the ability of any system to perform its safety-
    related function. As described above, these proposed changes are 
    administrative in nature. They do not change any plant operating 
    parameters or design features and do not reduce the level of 
    effectiveness of any existing administrative controls. The proposed 
    change will not result in changes to the bases for any technical 
    specification. The establishment of the SQRP will continue to 
    provide for the adequate review of procedures. In addition, another 
    direct benefit of this program is that the amount of material 
    presented to PORC/SORC will decrease. The reduction in the amount of 
    material presented to PORC/SORC for review will allow the PORC/SORC 
    to focus on safety significant issues. Since none of the assumptions 
    in the technical specifications bases are affected by the changes 
    presented in this license amendment request, the margin of safety 
    which exists in the current technical specifications is not reduced.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Learning Resources Center, 
    Three Rivers Community-Technical College, 574 New London Turnpike, 
    Norwich, CT 06360
        Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
    Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
    CT 06141-0270
        NRC Project Director: Phillip F. McKee
    
    Northern States Power Company, Docket No. 50-263, Monticello 
    Nuclear Generating Plant, Wright County, Minnesota
    
        Date of amendment request: June 22, 1995
        Description of amendment request: The proposed changes modify the 
    facility requirements for thermal-hydraulic instability avoidance and 
    protection to address concerns over reactor fuel performance during 
    instability events. Changes are proposed to the Technical 
    Specifications to utilize the flow biased Average Power Range Monitor 
    high neutron flux scram and a power-flow map exclusion region 
    consistent with one of the NRC approved BWR Owners' Group solutions. In 
    addition, a change to correct an error in the Average Planar Linear 
    Heat Generation Rate during single loop operation is also proposed.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards 
    
    [[Page 45182]]
    consideration, which is presented below:
        a. The proposed amendment will not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The implementation of BWR Owner's Group long term stability 
    solution Option 1-D at Monticello does not modify the assumptions 
    contained in the existing accident analysis. The use of an exclusion 
    region and the operator actions required to avoid and minimize 
    operation inside the region do not increase the possibility of an 
    accident. Conditions of operation outside of the exclusion region 
    are within the analytical envelope of the existing safety analysis. 
    The operator action requirement to exit the exclusion region upon 
    entry minimizes the probability of an oscillation occurring. The 
    actions to drive control rods and/or to increase recirculation flow 
    to exit the region are maneuvers within the envelope of normal plant 
    evolutions. The flow based scram has been analyzed and will provide 
    automatic fuel protection in the event of a core wide instability. 
    Thus, each proposed operating requirement provides defense in depth 
    for protection from an instability event while maintaining the 
    existing assumptions of the accident analysis. The proposed change 
    to the method by which the MAPLHGR [maximum average planar linear 
    heat-generation rate] is obtained for single loop operation is 
    consistent with the analysis performed for the Average Power Range 
    Monitor/Rod Block Monitor Technical Specifications (ARTS) program. 
    The analysis performed in support of the ARTS program demonstrated 
    that the limits established assure compliance with fuel limits. 
    Therefore, this amendment will not cause a significant increase in 
    the probability or consequences of an accident previously evaluated 
    for the Monticello plant.
        b. The proposed amendment will not create the possibility of a 
    new or different kind of accident from any accident previously 
    analyzed.
        As stated above, the proposed operating requirements either 
    mandate operation within the envelope of existing plant operating 
    conditions or force specific operating maneuvers within those 
    carried out in normal operation. Since operation of the plant with 
    all of the proposed requirements is within the existing operating 
    basis, an unanalyzed accident will not be created through 
    implementation of the proposed change. Therefore, the proposed 
    amendment will not create the possibility of a new or different kind 
    of accident.
        c. The proposed amendment will not involve a significant 
    reduction in the margin of safety.
        Each of the proposed requirements for the plant thermal-
    hydraulic stability provides a means for fuel protection. The 
    combination of avoiding possible unstable conditions and the 
    automatic flow biased reactor scram provides an in-depth means for 
    fuel protection. Therefore, the individual or combination of means 
    to avoid and suppress an instability supplements the margin of 
    safety. The operating limits established for the single loop 
    operation MAPLHGR provide an acceptable margin of safety as 
    demonstrated in NEDC-30492, ``Average Power Range Monitor, Rod Block 
    Monitor and Technical Specification Improvement (ARTS) Program for 
    Monticello Nuclear Generating Plant-April 1984.'' The proposed 
    amendment will 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: Minneapolis Public Library, 
    Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
    Minnesota 55401
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    and Trowbridge, 2300 N Street, NW, Washington, DC 20037
        NRC Project Director: John N. Hannon
    
    Northern States Power Company, Docket No. 50-263, Monticello 
    Nuclear Generating Plant, Wright County, Minnesota
    
        Date of amendment request: July 5, 1995
        Description of amendment request: The proposed amendment, part of 
    the Monticello Surveillance Test Interval/Allowed Outage Time (STI/AOT) 
    Program, extends the surveillance test intervals and allowable out-of-
    service times for selected instrumentation. The proposed changes are 
    intended to minimize unnecessary testing and remove excessively 
    restrictive out-of-service times that could potentially degrade overall 
    plant safety and availability.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        a. The proposed amendment will not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The maximum failure frequency change is for the ECCS Actuation 
    Instrumentation as identified by General Electric topical report 
    NEDC-30936P-A, and Monticello specific report RE-006. These reports 
    concluded core damage frequency changed by less than 4% when STIs 
    were increased to once per 3 months, AOTs for surveillance were 
    increased to 6 hours, and AOTs for repair were increased to 24 
    hours. Since this small increase was within the guideline of 
    acceptability stated in NEDC-30936P-A, and Monticello only proposes 
    to increase the repair AOT to 12 hours rather than 24 hours, this 
    amendment will not cause a significant increase in the probability 
    or consequences of an accident previously evaluated for the 
    Monticello plant (see RE-006).
        The drift analysis determined the associated instrumentation 
    would not be adversely effected with the longer calibration 
    intervals. Pertinent process parameters including instrument drift 
    will still be within acceptance criteria with the longer 
    surveillance intervals.
        The recirculation flow meters and flow instrumentation are not 
    used in any safety or accident analysis. Therefore, no analysis 
    would be changed by increasing the calibration interval to once per 
    cycle.
        b. The proposed amendment will not create the possibility of a 
    new or different kind of accident from any accident previously 
    analyzed.
        These changes only affect the instrument STI and AOT times. No 
    changes are being made to the functions of the instrumentation. 
    Therefore, the proposed amendment will not create the possibility of 
    a new or different kind of accident.
        c. The proposed amendment will not involve a significant 
    reduction in the margin of safety.
        These changes will improve the performance of equipment and are 
    intended to reduce the potential for equipment failures due to 
    unnecessary testing. The safety limits and the limiting safety 
    system setpoints will not be affected by these changes. No safety 
    margins are affected, therefore, the drift will remain within the 
    margins of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Minneapolis Public Library, 
    Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
    Minnesota 55401
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    and Trowbridge, 2300 N Street, NW, Washington, DC 20037
        NRC Project Director: John N. Hannon
    
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
    Station,Unit No. 1, Washington County, Nebraska
    
        Date of amendment request: August 4, 1995
        Description of amendment request: This proposed amendment would 
    revise the Technical Specifications (TS) for the requirements for the 
    containment radiation high signal (CRHS) and the safety injection and 
    refueling water (SIRW) tank low signal (STLS) contained in TS 2.15, 
    Tables 2-3 and 2-4. Specification 3.1, Table 3-2 will also be revised 
    to include administrative changes to the CRHS surveillance 
    
    [[Page 45183]]
    methods to be consistent with the applicable surveillance functions. 
    The Basis for Specification 2.15 is being revised to clarify that the 
    number of installed channels for CRHS is two. The term ``SOURCE CHECK'' 
    is being deleted from the Definitions section.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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.
        The Omaha Public Power District (OPPD) proposes to revise 
    Technical Specification (TS) 2.15, Table 2-3 by revising the 
    requirement for placing the Safety Injection Refueling Water (SIRW) 
    tank low level channel(s) in the tripped condition to placing them 
    in the bypassed condition. Due to the derived signal, if a channel 
    was in the tripped condition and a single failure occurred, (that 
    being one channel of STLS on either A or B circuits), a premature 
    SIRW tank low signal (STLS) would be generated. During a design 
    basis accident (DBA) with a valid Containment Pressure High Signal 
    (CPHS) or Pressurizer Pressure Low Signal (PPLS), this single 
    failure would prevent the contents of the SIRW tank from being 
    injected into the reactor coolant system. The resulting logic of 
    placing the SIRW tank low level channels in BYPASS rather than TRIP 
    would not cause a premature switchover of the high pressure safety 
    injection pumps to the containment sump and it would not prevent the 
    switchover when needed.
        OPPD also proposes to revise TS 2.15, Table 2-4, by reducing the 
    number of minimum operable Containment Radiation High Signal (CRHS) 
    channels from two to one. This proposed change revises the 
    requirements of TS 2.15 to coincide with changes to the TS and 
    Offsite Dose Calculation Manual (ODCM) that were implemented by TS 
    Amendment 152. The Engineered Safety Feature (ESF) actuation system 
    supervisory A and B safeguard initiation channels will not be 
    affected by this proposed TS change. The minimum level of engineered 
    safeguards performance acceptable for the DBA, (i.e., minimum 
    safeguards) will continue to be maintained in accordance with IEEE 
    279 - 1971, ``Criteria for Protection Systems for Nuclear Power 
    Generating Stations.''
        Included in this change are administrative revisions to TS 3.1, 
    Table 3-2, for replacing the current surveillance methods for 
    checking and testing the CRHS instrumentation with the defined terms 
    ``CHANNEL CHECK'' and ``CHANNEL FUNCTIONAL TEST,'' respectively. 
    These proposed revisions are administrative in nature and reflect 
    TS-defined terminology for the instrumentation surveillance methods 
    utilized to ensure that the CRHS instrumentation is operable. A 
    channel check requires a qualitative determination of acceptable 
    operability by observation of channel behavior during normal plant 
    operation. A channel functional test requires the injection of a 
    simulated signal into the channel to verify that it is operable, 
    including any alarm and/or trip initiating actions. Other proposed 
    administrative changes include deleting the term ``SOURCE CHECK'' 
    from the TS Definitions section as source check will no longer be 
    used in the FCS TS and adding verbiage to the TS 2.15 Basis for 
    clarifying that the number of installed channels for CRHS is two.
        Therefore, the proposed change, as described above, would not 
    increase 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.
        There will be no physical alterations to the plant 
    configuration, changes to setpoint values, or changes to the 
    implementation of setpoints or limits as a result of the proposed 
    changes to TS 2.15, Tables 2-3 and 2-4. The proposed revisions to TS 
    3.1, Table 3-2 are administrative changes to make the TS more 
    accurately reflect defined terminology and the methods utilized to 
    ensure that the CRHS instrumentation is operable. The proposed TS 
    revisions do not require any changes to the present methods of 
    verifying CRHS instrumentation operability. Therefore, the proposed 
    change does not create the possibility of a new or different kind of 
    accident from any previously evaluated.
        3. Involve a significant reduction in a margin of safety.
        There are no changes to the equipment or plant operations as a 
    result of the changes being made to the number of minimum operable 
    CRHS channels. The proposed changes to the STLS will require that 
    the inoperable channel be placed in BYPASS rather than TRIP. This 
    action would ensure that a single failure would not cause a 
    premature safety injection switchover to the containment sump and 
    would not prevent switchover when needed. Therefore, this proposed 
    change does not reduce a margin of safety.
        The proposed revisions to TS 3.1, Table 3-2 are administrative 
    changes to make the TS more accurately reflect defined terminology 
    and the methods utilized to ensure that the CRHS instrumentation is 
    operable. The proposed TS revisions do not require any changes to 
    the present methods of verifying CRHS instrumentation operability. 
    The proposed changes to the Definitions and TS 2.15 Basis sections 
    are administrative in nature. Therefore, these proposed changes do 
    not involve a significant reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location:  W. Dale Clark Library, 215 
    South 15th Street, Omaha, Nebraska 68102
        Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
    Street, N.W., Washington, DC 20005-3502
        NRC Project Director: William H. Bateman
    
    Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
    311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
    County, New Jersey
    
        Date of amendment request: June 22, 1995
        Description of amendment request: The amendments would revise the 
    Technical Specifications 3.4.1.4 and 3.9.8.2 by deleting footnotes and 
    associated information regarding Service Water header operation and its 
    support function for Residual Heat Removal operation. These footnotes 
    and associated information had been placed in the Technical 
    Specifications because of the concern about Service Water system piping 
    integrity in the mid-1980's.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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. Do not involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        Even though one service water loop will be out for maintenance, 
    both loops of residual heat removal (RHR) will be kept operable, 
    consistent with the requirements of STS (NUREG 1431). A minimum of 
    two RHR, two component cooling (CC), and two service water (SW) 
    pumps, powered from two different vital busses, will be kept 
    operable.
        Only one component cooling heat exchanger will be operable since 
    only one service water loop is operable. The CC heat exchangers for 
    both Units 1 and 2 have a very high reliability. The primary heat 
    transfer surfaces of the heat exchangers are made of titanium; no 
    material problems have been experienced in ten years of service.
        The remaining active components that, through misoperation, 
    could potentially defeat RHR capability are, (1) the motor operated 
    valves in RHR or SW that could develop a ``hot short'' and 
    subsequently close and (2) the air operated temperature/ flow 
    control valves of the CC heat exchangers. Additional actions will be 
    taken to effectively eliminate the possibility of these single point 
    valves from failing and defeating RHR capability. The motor operator 
    breakers will be tagged open during MODES 5 and 6, except for 
    flooding the cavity, when the RHR suction valves must be closed. The 
    CC Heat Exchanger air operated temperature/flow control valves fail 
    open, or as is, on loss of air which is the safe position. Operators 
    will monitor critical temperatures; this equipment is accessible if 
    any corrective action is required. Thus, with one service water 
    header out of service, the intent of the 
    
    [[Page 45184]]
    technical specifications as defined in the bases section (to have a 
    single failure proof RHR system) is met with the proposed system 
    configuration. Therefore, the proposed changes do not involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        2. Do not create the possibility of a new or different kind of 
    accident from any previously evaluated.
        The catastrophic failure of a moderate energy Class 3 piping 
    system is not a credible event, based on the upgraded reliability of 
    the system, the redundancy of active components, the elimination of 
    single failure points, and on the industry and regulatory positions 
    established for this type of system. Since SW is a Class 3 moderate 
    energy system, the only postulated passive failure mode is a leakage 
    crack. In accordance with Generic Letter (GL) 91-18 and GL 90-05, a 
    leak in the SW system, following acceptable evaluation, does not 
    constitute a failure that causes the loss of capability to perform 
    it's intended safety function. A moderate energy Class 3 piping leak 
    does not cause the system to be declared inoperable. Therefore, the 
    proposed changes do not create the possibility of a new or different 
    type of accident from any previously evaluated.
        3. Do not involve a significant reduction in a margin of safety.
        RHR redundancy is maintained; no credible single failure point 
    exists that could cause a nonrecoverable loss of SW. Therefore, the 
    proposed changes do not involve a significant reduction in a margin 
    of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Salem Free Public library, 112 
    West Broadway, Salem, New Jersey 08079
        Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
    Strawn, 1400 L Street, NW, Washington, DC 20005-3502
        NRC Project Director: John F. Stolz
    
    Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
    Ginna Nuclear Power Plant, Wayne County, New York
    
        Date of amendment request: September 15, 1992, as supplemented 
    April 20, 1993, April 26, 1995, and July 27, 1995.
        Description of amendment request: The proposed amendment would 
    revise Technical Specifications (TSs) 3.1.1.4, 3.1.1.6, and 4.3.4, and 
    add a Basis to address Generic Letter (GL) 90-06. GL 90-06 represents 
    the technical resolution of Generic Issue (GI) 70, ``Power Operated 
    Relief Valve and Block Valve Reliability,'' and GI 94, ``Additional Low 
    Temperature Overpressure Protection for Light Water Reactors.'' The 
    resolution of these issues proposes new requirements and TS changes 
    that enhance the reliability of power-operated relief valves (PORVs) 
    and block valves along with TS changes that will provide additional 
    low-temperature overpressure protection (LTOP).
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        There is no significant increase in the probability or 
    consequences of an accident previously evaluated because the 
    accident conditions and assumptions are not significantly affected 
    by the proposed change.
        The proposed change to action statement 3.1.1.4a(i) [proposed to 
    be renumbered to 3.1.1.6c] to include the removal of power from a 
    closed block valve will provide additional assurance to preclude any 
    inadvertent opening of the block valve at a time in which the PORV 
    may not be operable to assure RCS [reactor coolant system] 
    integrity.
        The provision of the generic letter requires, with one or both 
    PORV(s) inoperable to initiate shutdown actions if PORV operability 
    is not restored within 72 hours or 1 hour respectively. RG&E 
    [Rochester Gas and Electric Corporation] does not address these 
    shutdown actions, but rather will concentrate on re-establishing 
    valve operability. If the block valve(s) and power are not removed 
    within 1 hour shutdown provisions must be initiated. [***].
        Proposed action statement 3.1.1.4a(ii) [proposed to be 
    renumbered to 3.1.1.6d] includes a provision to place the block 
    valves associated PORV(s) switch in manual control due to an 
    inoperable block valve(s). This requirement precludes the automatic 
    opening for an overpressure event to avoid the potential for a 
    stuck-open PORV at a time that the block valve is open and 
    inoperable. [***].
        The proposed change of maintaining power to closed block valves 
    could potentially increase the probability of an inadvertent opening 
    of a block valve. The safety impact is, however, not significant 
    since the proposed changes are only applicable if the PORV is 
    inoperable due to excessive seat leakage (proposed action 3.1.1.6b). 
    [***].
        Proposed action statement 3.1.1.6b establishes reactor coolant 
    pressure boundary integrity for a PORV that has excessive seat 
    leakage and is therefore considered operable to perform its intended 
    safety function. [***].
        Proposed Surveillance Requirement 4.3.4.3 addresses operability 
    of the Nitrogen System by demonstration of the PORVs at least once 
    per 18 months by operating the PORVs through a complete cycle of 
    full travel. [***].
        Based on the above efforts, the proposed amendment does not 
    involve a significant increase in the probability or consequences of 
    any accident previously evaluated.
        The possibility of a new or different kind of accident from any 
    previously evaluated is not created. In matters related to nuclear 
    safety, all accidents continue to bound previous analyses. The 
    proposed changes do not add or modify any equipment design nor do 
    the proposed changes involve any significant operational changes to 
    any plant systems.
        The proposed amendment does not involve a significant reduction 
    in the margin of safety as defined in the basis for any technical 
    specification because the results of the accident analyses which are 
    documented in the UFSAR [Updated Final Safety Analysis Report] 
    continue to bound operation under the proposed changes so that there 
    is no safety margin reduction. [***].
        Therefore, the proposed changes do not involve a significant 
    reduction in margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Rochester Public Library, 115 
    South Avenue, Rochester, New York 14610
        Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
    L Street, NW., Washington, DC 20005
        NRC Project Director: Ledyard B. Marsh
    
    Sacramento Municipal Utility District (SMUD), Docket No. 50-312, 
    Rancho Seco Nuclear Station, Sacramento County, California
    
        Date of amendment request: June 20, 1995 and as amended August 14, 
    1995
        Description of amendment request: The proposed amendment (PA-191) 
    would permit SMUD to change the Fuel Storage Building load handling 
    limits to allow placing the shield plugs on the dry shielded cannisters 
    in order to permit transfer of spent fuel assemblies from the spent 
    fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage 
    Installation.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        PA-191 will not create a significant increase in the probability 
    or consequences of an accident previously evaluated in the Safety 
    Analysis Report (SAR), because dropping the dry shielded canister 
    (DSC) top shield plug over a DSC loaded with 24 spent fuel 
    assemblies is not considered a credible event. Also, the gantry 
    crane is designed such 
    
    [[Page 45185]]
    that it can only handle loads over the SFP cask pit area and can not 
    move a load over the SFP fuel storage racks.
        PA-191 will not create the possibility of a new or different 
    type of accident than previously evaluated in the SAR, because the 
    proposed Permanently Defueled Technical Specification heavy load 
    handling exceptions do not create a new credible accident scenario. 
    Dropping the DSC top shield plug and damaging spent fuel assemblies 
    is not considered a credible event.
        PA-191 will not involve a significant reduction in the margin of 
    safety, because the proposed heavy load handling exceptions do not 
    create a credible accident scenario.
        The NRC staff has reviewed the licensee's analyses of June 20, 1995 
    and August 14, 1995. The August 14 submittal enhanced these analyses by 
    providing design details regarding the significant safety factors built 
    into the crane and other lifting hardware. Based on this review, it 
    appears that the three standards of 50.92(c) are satisfied. Therefore, 
    the NRC staff proposes to determine that the amendment request involves 
    no significant hazards consideration.
        Local Public Document Room location:  Central Library, Government 
    Documents 828 I Street, Sacramento, CA 95814
        Attorney for licensee: Dana Appling, Esq. Sacramento Municipal 
    Utility District, P. O. Box 15830, Sacramento, CA 95852-1830
        NRC Project Director: Seymour H. Weiss
    
    Southern California Edison Company, et al., Docket Nos. 50-361 and 
    50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
    San Diego County, California
    
        Date of amendment requests: July 17, 1995
        Description of amendment requests: The licensee proposes to revise 
    surveillance requirements associated with Technical Specifications 3/
    4.3.1, ``Reactor Protective Instrumentation,'' and 3/4.3.2, 
    ``Engineered Safety Feature Actuation System Instrumentation.'' The 
    surveillance interval is to be increased to 120 days for performance of 
    channel functional tests for certain reactor protective system and 
    engineered safety feature actuation system instrumentation. The 
    proposed change also revises Bases 3/4.3.1, ``Reactor Protective and 
    Engineered Safety Features Actuation System Instrumentation,'' to 
    reflect the new interval.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        1. The proposed change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated.
        The proposed change would extend the current sequential Channel 
    Functional Test (CFT) surveillance interval for Plant Protective 
    System (PPS) instrumentation and Nuclear Instrumentation (NI). This 
    change does not involve any changes to plant equipment or operation. 
    The proposed change actually maintains or decreases the PPS system 
    unavailability. PPS uncertainty and setpoint modifications will 
    account for the new surveillance interval. Therefore, the proposed 
    change will not involve a significant increase in the probability or 
    consequences of any 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 amendment request does not involve any change to plant 
    equipment or operation. The PPS system is used for monitoring and 
    mitigation of evaluated accidents. Increasing the availability of 
    the PPS system, as proposed in this amendment request, will not 
    create the possibility of a new or different kind of accident from 
    any previously evaluated.
        3. The proposed change does not involve a significant reduction 
    in a margin of safety.
        This amendment does not change the manner in which safety 
    limits, limiting safety settings, or limiting conditions for 
    operation are determined. This amendment request will increase 
    Reactor Protective System and Engineered Safety Features Actuation 
    System availability. Therefore, this amendment will not involve a 
    significant reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment requests involve no significant hazards consideration.
        Local Public Document Room location: Main Library, University of 
    California, P. O. Box 19557, Irvine, California 92713
        Attorney for licensee: T. E. Oubre, Esquire, Southern California 
    Edison Company, P. O. Box 800, Rosemead, California 91770
        NRC Project Director: William H. Bateman
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of amendment request: August 7, 1995 (TS 95-12)
        Description of amendment request: The proposed change would correct 
    various errors of an editorial nature that have been identified in the 
    technical specifications and remove the provisions that have exceeded 
    their allowed time interval for implementation or the required 
    conditions no longer exist.
        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:
        TVA has evaluated the proposed technical specification (TS) 
    change and has determined that it does not represent a significant 
    hazards consideration based on criteria established in 10 CFR 
    50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
    with the proposed amendment will not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        The proposed revisions do not change the TS requirements, plant 
    setpoints or functions, or plant operating practices. These changes 
    provide clarifications to the existing TSs by correcting editorial 
    errors and removing provisions that no longer apply in the 
    specifications. The probability or consequences of an accident will 
    not be increased by providing the proposed verbiage corrections that 
    are editorial and nonintent.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        No plant functions or compliance activities associated with the 
    TS requirements have been affected by the proposed editorial 
    changes. Therefore, the possibility of a new or different kind of 
    accident is not created.
        3. Involve a significant reduction in a margin of safety.
        The proposed changes will not alter TS setpoint values or 
    functions. The proposed corrections will enhance the application of 
    TS requirements and will support the margin of safety provided by 
    the TSs. Therefore, the margin of safety will not be reduced by the 
    proposed revisions.
        The NRC has reviewed the licensee's analysis and, based on 
    thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1101 Broad Street, Chattanooga, Tennessee 37402
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    [[Page 45186]]
    
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of amendment request: August 7, 1995 (TS 95-17)
        Description of amendment request: The proposed change would 
    relocate the heat flux hot channel factor penalty of two percent from 
    Surveillance Requirement 4.2.2.2.e.1 to the Core Operating Limits 
    Report and add a reference to the factor to Specification 6.9.1.14.5. 
    Also, Specification 6.9.1.14.a.2 would be revised to reference Revision 
    1A of Westinghouse Commercial Atomic Power (WCAP) 10216-P-A, 
    ``Relaxation of Constant Axial Offset Control - FQ Surveillance 
    Technical Specifications,'' dated February 1994.
        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:
        TVA has evaluated the proposed technical specification (TS) 
    change and has determined that it does not represent a significant 
    hazards consideration based on criteria established in 10 CFR 
    50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
    with the proposed amendment will not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        The proposed change involves only the manner in which the 
    penalty factors for FQ(Z) would be specified (i.e, a burnup-
    dependent factor specified in the Core Operating Limits Report 
    [COLR] versus a constant factor specified in the TS). This is simply 
    used to account for the fact that FQ(Z) may increase between 
    surveillance intervals. These penalty factors are not assumed in any 
    of the initiating events for the accident analyses. Therefore, the 
    proposed change will have no effect on the probability of any 
    accidents previously evaluated. The penalty factors specified in the 
    COLR will be calculated using NRC-approved methodology and will 
    therefore continue to provide an equivalent level of protection as 
    the existing TS requirement. Therefore, the proposed change will not 
    affect the consequences of any accident previously evaluated.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        The proposed change does not involve a physical alteration to 
    the plant (no new or different kind of equipment will be installed) 
    or alter the manner in which the plant would be operated. Thus, this 
    change does not create the possibility of a new or different kind of 
    accident from any accident previously evaluated.
        3. Involve a significant reduction in a margin of safety.
        The proposed change will continue to ensure that potential 
    increases in FQ(Z) over a surveillance interval will be 
    properly accounted for. The penalty factors will be calculated using 
    NRC-approved methodology. Therefore, the proposed change will not 
    involve a reduction in margin of safety.
        The NRC has reviewed the licensee's analysis and, based on 
    thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1101 Broad Street, Chattanooga, Tennessee 37402
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of amendment request: August 7, 1995 (TS 95-18)
        Description of amendment request: The proposed change would revise 
    the titles of various administrative positions found in Section 6.0 of 
    the Technical Specifications.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        TVA has evaluated the proposed technical specification (TS) 
    change and has determined that it does not represent a significant 
    hazards consideration based on criteria established in 10 CFR 
    50.92(c).
        Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
    proposed amendment will not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        The proposed changes only involve the administrative titles of 
    management positions in TVA [Tennessee Valley Authority]. Plant 
    equipment and operating practices are not affected by the proposed 
    administrative changes. Therefore, there is no increase in the 
    probability or consequences of an accident.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        Plant features are not impacted by the proposed revision; 
    therefore, this revision can not create the possibility of a new or 
    different accident.
        3. Involve a significant reduction in a margin of safety.
        Plant setpoints and features that establish and maintain the 
    margin of safety for SQN are not involved in the proposed 
    administrative TS change. Therefore, the margin of safety is not 
    reduced by the proposed change.
        The NRC has reviewed the licensee's analysis and, based on 
    thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.Local 
    Public Document Romm location: Chattanooga-Hamilton County Library,1101 
    Broad Street, Chattanooga, Tennessee 37402
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of amendment request: August 7, 1995 (TS 95-03)
        Description of amendment request: The proposed change would modify 
    Technical Specifications (TS) 3/4.1.3, ``Movable Control Assemblies,'' 
    and Bases 3/4.1.3. The proposed change addresses operation with a rod 
    urgent failure condition (the control rods are out-of-service because 
    of failures external to the individual rod drive mechanisms; i.e., 
    programming circuitry, but the control rods remain operable), including 
    limited operation with one control or shutdown bank inserted up to 18 
    steps below its insertion point. In addition, the surveillance interval 
    for rod movement verifications would be increased from 31 days to 92 
    days.
        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:
        TVA has evaluated the proposed technical specification (TS) 
    change and has determined that it does not represent a significant 
    hazards consideration based on criteria established in 10 CFR 
    50.92(c).
        Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
    proposed amendment will not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        Allowing for continued operation during diagnosis and repair as 
    a result of electronic or electrical malfunctions of the rod control 
    system is acceptable, since the design safety function of the 
    control rods (reactor trip will remain unaffected during the 
    diagnosis and repair period. During the extended 
    
    [[Page 45187]]
    troubleshooting and repair period, the requirements for control rod 
    alignment, insertion limits (except for a small allowed deviation 
    for one bank) and shutdown margin will be maintained. The small 
    deviation from the control rod insertion limits allowed for one 
    bank, for up to 72 hours, will not adversely impact the current TS 
    requirements for normal operation core power distributions. The 
    proposed changes do not affect the ability of the control rods to 
    perform their intended safety function (rods remain trippable) when 
    a safety system setting is reached. No new or unique accident 
    precursors be introduced by the proposed changes. Therefore, the 
    probability and consequences of accidents related to or dependent on 
    control rod operation will remain unaffected.
        The proposed change will result in a small increase in the 
    probability, that at any given time, a control or shutdown bank will 
    be inserted slightly below (i.e., up to 18 steps) its insertion 
    limit. However, by design, the control and shutdown banks will 
    continue to meet the safety analysis criterion for steady state and 
    American Nuclear Society (ANS) Condition II (moderate frequency) 
    transients. The allowed insertion is not a malfunction of equipment 
    important to safety in this case; therefore, the probability of such 
    a malfunction is not increased. Limiting the allowed time for 
    operation with the rod control system out-of-service, but with the 
    rods trippable and with a control or shutdown bank below the 
    insertion limit, eliminates the need for consideration of this 
    condition coincident with any of the low frequency (ANS Condition 
    III or IV) design basis accidents.
        2. Create the possibility of a new or different kind of accident 
    from any previously analyzed.
        There are no new failure mechanisms associated with plant 
    operation for an extended period to perform diagnosis and repair on 
    the rod control system. Limited periods of operation with immovable, 
    but trippable control rods, does not involve any modification to the 
    operational limits or physical design of the involved systems. There 
    are no new accident precursors created because of the allowed 
    diagnosis and repair period.
        3. Involve a significant reduction in a margin of safety.
        The results of the current accident analyses are not impacted by 
    the change. In addition, the margin of safety as defined in the 
    basis of the TS has not been reduced because current core design 
    limits continue to be met for the accidents of concern. Therefore, 
    the margin of safety is not impacted.
        The NRC has reviewed the licensee's analysis and, based on 
    thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1101 Broad Street, Chattanooga, Tennessee 37402
        Attorney for licensee: General Counsel, Tennessee Valley Authority, 
    400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
        NRC Project Director: Frederick J. Hebdon
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of amendment request: June 23, 1995
        Description of amendment request: The proposed amendment would 
    revise Technical Specification (TS) Surveillance Requirements 
    4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2, and Table 4.3-3 in 
    accordance with guidance provided in NRC Generic Letter (GL) 93-05, 
    ``Line Item Technical Specification Improvements to Reduce Surveillance 
    Requirements for Testing During Power Operations.'' Additionally, the 
    proposed amendment would revise TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1 and 
    associated Bases to implement portions of the Standard Technical 
    Specifications - Westinghouse Plants, NUREG-1431.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 Technical Specification changes do not involve a 
    significant hazards consideration per 10 CFR 50.92 because operation 
    of Callaway Plant with the changes would not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        All changes are in accordance with the recommendations of NRC 
    Generic Letter 93-05, Line-Item Technical Specifications 
    Improvements to Reduce Surveillance Requirements for Testing During 
    Power Operation or NUREG 1431, Standard Technical Specifications - 
    Westinghouse Plants. None of the changes affects accident initiators 
    and each has been evaluated against Callaway Plant operating 
    experience.
        2. Create the possibility of a new or different kind of accident 
    from any previously evaluated.
        The proposed Technical Specification changes do not modify any 
    equipment nor create any potential accident initiators. The changes 
    per GL 93-05 involve Technical Specification surveillance 
    frequencies and do not alter the methodology nor associated 
    acceptance criteria. The changes per NUREG-1431 do not create any 
    accident initiators and are consistent with Callaway design and 
    operation.
        3. Involve a significant reduction in a margin of safety.
        The surveillance frequency changes were recommended via GL 93-05 
    and are compatible with Callaway Plant experience. The changes per 
    NUREG-1431 do not impact the margin of safety. The Shutdown margin 
    requirements and associated safety margins are unaffected by these 
    changes.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location:  Callaway County Public 
    Library, 710 Court Street, Fulton, Missouri 65251
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    & Trowbridge, 2300 N Street, NW., Washington, DC 20037
        NRC Project Director: Gail H. Marcus
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of amendment request: June 26, 1995
        Description of amendment request: The proposed amendment would 
    revise the allowed outage time for component cooling water motor 
    operated containment isolation valves, remove the list of containment 
    isolation valves, and allow containment penetration check valves to be 
    used as isolation devices.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 revision to TS 3/4.6 to remove the listing of 
    containment isolation valves, revise the ACTION Statement for the 
    CCW MOVs, and credit penetration check valves as isolation devices 
    does not involve a significant hazards consideration because 
    operation of Callaway Plant with this change would not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        The proposed changes simplify the TS, meet the regulatory 
    requirements for control of containment isolation and are consistent 
    with the guidelines of GL 91-08. The information contained in Table 
    3.6-1 has not been changed, but only relocated to a different 
    controlling document. This is an administrative change which should 
    result in improved plant practices and have no impact on plant 
    operations. Addition of the footnote to allow up to 12 hours for 
    valve testing does not affect the severity of any accident 
    previously evaluated. The proposed revision to the TS will not 
    adversely impact plant safety since the second barrier of the two 
    required is still available to provide isolation between the 
    containment atmosphere or the reactor coolant system and the outside 
    atmosphere.
    
    [[Page 45188]]
    
        2. Create the possibility of a new or different kind of accident 
    from any previously evaluated.
        There are no design changes being made that would create a new 
    type of accident or malfunction and the method and manner of plant 
    operation remain unchanged. Addition of the footnote to allow up to 
    12 hours for valve testing does not affect the severity of any 
    accident previously evaluated. The additional time provides 
    assurance that the inoperable valve is in proper working order prior 
    to returning it to OPERABLE condition.
        3. Involve a significant reduction in a margin of safety.
        There are no changes being made to the safety limits or safety 
    system settings that would adversely impact plant safety. 
    Containment isolation will still be maintained as provided by the 
    second isolation valve to ensure that the release of radioactive 
    material to the environment will be consistent with the assumptions 
    used in the analyses for a LOCA. This will assure that containment 
    integrity is maintained.
        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:  Callaway County Public 
    Library, 710 Court Street, Fulton, Missouri 65251
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    & Trowbridge, 2300 N Street, NW., Washington, DC 20037
        NRC Project Director: Gail H. Marcus
    
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
    Callaway County, Missouri
    
        Date of amendment request: July 25, 1995
        Description of amendment request: The proposed amendment would 
    revise Technical Specification (TS) 3/4.8.1 and its associated Bases to 
    improve overall emergency diesel generator reliability and 
    availibility.
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 do not involve a significant hazards 
    consideration because operation of Callaway Plant with these changes 
    would not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        These proposed changes do not involve a change in the 
    operational limits or physical design of the emergency power system. 
    Emergency diesel generator operability and reliability will continue 
    to be assured while minimizing the number of required emergency 
    diesel generator starts. Also, emergency diesel generator 
    reliability will be enhanced by minimizing severe test conditions 
    which can lead to premature failures.
        2. Create the possibility of a new or different kind of accident 
    from any previously evaluated.
        These proposed changes do not involve a change in the 
    operational limits or physical design of the emergency power system. 
    The performance capability of the emergency diesel generator will 
    not be affected. Emergency diesel generator reliability and 
    availability will be improved by the implementation of the proposed 
    changes. There is no actual impact on any accident anaiysis.
        3. Involve a significant reduction in the margin of safety.
        These proposed changes do not involve a change in the 
    operational limits or physical design of the emergency power system. 
    The performance capability of the emergency diesel generator will 
    not be affected. Emergency diesel generator reliability and 
    availability will be improved by the implementation of the proposed 
    changes. No margin of safety is reduced.
        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: Callaway County Public 
    Library, 710 Court Street, Fulton, Missouri 65251
        Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
    & Trowbridge, 2300 N Street, NW., Washington, DC 20037
        NRC Project Director: Gail H. Marcus
    
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
    North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
    Virginia
    
        Date of amendment request: November 29, 1994
        Description of amendment request: The proposed change would revise 
    and update the NA-1&2 Environmental Protection Plan (EPP) to reflect 
    current obligations to the Commonwealth of Virginia, revise portions of 
    the transmission corridor rights-of-way erosion control program for 
    clarification and to be consistent with the state regulations, 
    eliminate inconsistencies, and delete obsolete material. Specifically, 
    references to National Pollutant Discharge Elimination System (NPDES) 
    permits are changed to reflect the correct permit title, Virginia 
    Pollutant Discharge Elimination System (VPDES). Vegetation and aquatic 
    biota studies referred to in the EPP were satisfactorily completed on 
    or before June 24, 1986. The discussion of the detailed subject matter 
    in these studies is removed because it is extraneous information. A 
    reference to 10 CFR 51.5(b)(2) (which does not exist) is corrected to 
    10 CFR 51.60(b)(2). The explicit reporting requirements for unusual or 
    important environmental events are replaced with the reporting 
    requirement which the NRC has required pursuant to 10 CFR 50.72 
    (b)(2)(vi). Therefore, the reporting inconsistency (EPP requires report 
    to NRC within 24 hours, whereas the 10 CFR 50.72 requires a four hour 
    report to the NRC) is resolved. The description of the audit program to 
    be utilized for auditing the EPP is replaced by referring to the Audit 
    Program established in accordance with 10 CFR 50, Appendix B. Another 
    inconsistency is eliminated by revising the two year records retention 
    requirement for erosion control inspection field logs to five years. 
    This makes the requirement consistent with EPP Section 5.2, Records 
    Retention. References to the State Water Control Board are updated to 
    that agency's successor, the Department of Environmental Quality. 
    Additionally, the licensee's obligation to comply with Virginia 
    regulations concerning erosion and sediment control within the 
    transmission corridor rights-of-way are recognized to eliminate 
    redundancy with previous EPP commitments. The Virginia Soil and Water 
    Conservation Board is recognized as the regulatory authority concerning 
    erosion within the transmission corridor rights-of-way. The Virginia 
    Soil and Water Conservation Board reviews and approves erosion and 
    sediment control specifications submitted by utilities on an annual 
    basis.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        Specifically, operation of the North Anna Power Station in 
    accordance with the EPP changes will not:
        (1) Involve a significant increase in the probability or 
    consequences of an accident previously evaluated. The likelihood 
    that an accident will occur is neither increased or decreased by the 
    proposed changes to the EPP. Sufficient controls are established to 
    ensure that environmental controls impacting safety-related 
    structures, systems, and components are maintained current and 
    accurate. The only potentially credible accident which might be 
    affected is the Loss of Offsite Power (if erosion were severe 
    
    [[Page 45189]]
    enough to undermine the bases of a transmission tower). Each of the 
    three 500 KV transmission lines connected to North Anna Power 
    Station can supply sufficient power to the site. This limits the 
    effect that one transmission tower has on safe operation of the 
    nuclear facility. However, the erosion noted to date has not been 
    severe enough to make such an accident credible. Additionally, each 
    of the 500 KV transmission lines are inspected for material 
    condition annually. Although the intent of this inspection is not 
    soil erosion (the annual erosion inspections are currently conducted 
    by another group who specializes in land management), evidence of 
    severe erosion would be noted and addressed as appropriate. 
    Therefore, this EPP change will not impact the function or method of 
    operation of plant equipment. Thus, a significant increase in the 
    probability of a previously analyzed accident does not result due to 
    this change. Nuclear station systems, equipment, or components are 
    not affected by the proposed changes. Thus, the consequences of a 
    malfunction of equipment important to safety previously evaluated in 
    the UFSAR [Updated Final Safety Analysis Report] are not increased 
    by this change.
        (2) Create the possibility of a new or different kind of 
    accident from any accident previously evaluated. The proposed 
    changes do not involve changes to the physical plant or operations. 
    ... the proposed EPP changes do not contribute to accident 
    initiation and therefore do not produce a new accident scenario or 
    produce a new type of equipment malfunction. Also, this EPP change 
    does not alter any existing accident scenarios. The proposed changes 
    do not affect nuclear plant equipment or its operation, and thus do 
    not create the possibility of a new or different kind of accident. 
    Therefore, the proposed changes do not create the possibility of a 
    new or different kind of accident.
        (3) Involve a significant reduction in a margin of safety. The 
    EPP does not have a formal basis description other than the 
    discussion in the FES-OL [Final Environmental Statement-Operating 
    License]. The FES-OL discusses the non-radiological impacts of 
    facility construction and operation on the environment. The 
    discussion indicates that the environment will be managed to a 
    stabilized condition during the operations phase, and a program will 
    be implemented to maintain the environment in a stabilized 
    condition. This intent is not altered by the proposed changes to the 
    EPP. The proposed changes do not affect nuclear plant equipment or 
    its operation, and thus do not involve any reduction in the margin 
    of safety.
        Therefore, use of the proposed EPP would not involve any 
    reduction in the margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: The Alderman Library, Special 
    Collections Department, University of Virginia, Charlottesville, 
    Virginia 22903-2498
        Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
    Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
    Virginia 23219
        NRC Project Director: David B. Matthews
    
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
    North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
    Virginia
    
        Date of amendment request: July 26, 1995
        Description of amendment request: The proposed changes would revise 
    the Technical Specifications (TS) for the North Anna Power Station, 
    Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes 
    would increase the pressurizer safety valve lift setpoint tolerance as 
    well as reduce the pressurizer high pressure reactor trip setpoint and 
    allowable value.
        The licensee has prepared a safety evaluation which justifies 
    increasing the current TS pressurizer safety valve (PSV) at-power 
    (Modes 1-3) lift setpoint tolerance from plus or minus 1% as-found and 
    plus or minus 1% as-left to +2%/-3% average as-found with no single 
    valve outside plus or minus 3% as-found and plus or minus 1% per valve 
    as-left. The as-found value is based on testing, the results of which 
    are expressed as an error (i.e., positive or negative percentage 
    deviation from the nominal lift setpoint). The errors of the tested 
    valves are summed and the result divided by the number of valves 
    tested. This result is compared to the acceptable range of +2% to -3%. 
    No single valve is allowed to be outside of the plus or minus 3% 
    tolerance.
        The safety evaluation also supports an increase to the Hot Shutdown 
    (Mode-4) required PSV lift setpoint tolerance from plus or minus 1% as-
    found and plus or minus 1% as-left to plus or minus 3% per valve as-
    found and plus or minus 1% per valve as-left. These proposed changes 
    will provide greater operational flexibility in meeting periodic test 
    requirements established by the safety analyses.
        A concurrent reduction in the pressurizer high pressure reactor 
    trip setpoint and allowable value of TS Table 2.2-1 are also proposed. 
    These changes ensure that the analysis results for the loss of external 
    load accident continue to meet the acceptance criteria with the higher 
    PSV tolerance.
        The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses 
    demonstrate that increasing the at-power PSV lift setpoint tolerance to 
    +2%/-3% average as-found with no single valve outside plus or minus 3% 
    as-found and plus or minus 1% per valve as-left does not result in a 
    transient pressure in excess of the overpressure safety limit. Further, 
    the increased setpoint tolerance does not adversely impact the DNBR 
    [departure from nucleate boiling ratio] results of any North Anna UFSAR 
    [Updated Final Safety Analysis Report] Chapter 15 transient analysis. 
    Mode 4 overpressure protection is adequate with one PSV with a 
    tolerance of plus or minus 3%.
        Finally, the increased PSV setpoint tolerances and reduction of the 
    high pressurizer pressure reactor trip setpoint do not present any 
    operational considerations which would significantly impact the 
    performance of the plant during normal operation or during postulated 
    accident conditions. In summary, each pertinent safety criterion was 
    evaluated for the proposed TS changes, and all were found to be 
    acceptable.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        Specifically, operation of North Anna Power Station in 
    accordance with the proposed Technical Specifications changes will 
    not:
        1. Involve a significant increase in the probability or 
    consequence of an accident previously evaluated.
        Affected safety related parameters were analyzed for a change to 
    North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and 
    Table 2.2-1 item 10. It was determined that the overpressure safety 
    limits would not be exceeded in the most limiting overpressure 
    transients (Loss of Load, Locked Rotor, and Rod Withdrawal events) 
    with the as-found pressurizer safety valve lift setpoint tolerance 
    increased to an average of +2%/-3%, no single valve outside of [plus 
    or minus] 3%, and the 25 psi reduction in the Pressurizer High 
    Pressure Reactor Trip setpoint. The DNBR results of transients 
    impacted by the proposed setpoint tolerance increase meet the 
    acceptance criterion after accounting for the impact of the proposed 
    changes. The increased setpoint tolerance will not result in an 
    inadvertent opening of the pressurizer safety valves. Mode 4 
    overpressure protection is adequate with one PSV with a tolerance of 
    [plus or minus] 3%.
        2. Create the possibility of a new or different kind of accident 
    from any accident previously identified.
        The proposed change to North Anna 1 and 2 Technical 
    Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not 
    involve any changes which would introduce any new or unique 
    operational modes or accident precursors. Only the allowable 
    tolerance about the existing PSV lift setpoint will be changed, 
    along with a reduction in the 
    
    [[Page 45190]]
    pressurizer high pressure reactor trip setpoint.
        3. Involve a significant reduction in a margin of safety.
        It was determined that the most limiting overpressure transients 
    do not result in maximum pressures in excess of the overpressure 
    safety limits. The DNBR results of transients impacted by the 
    proposed setpoint tolerance increase meet the acceptance criterion 
    after accounting for the impact of the proposed changes. Therefore, 
    the margin of safety is unchanged by the proposed increase in the 
    safety valve setpoint tolerances.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: The Alderman Library, Special 
    Collections Department, University of Virginia, Charlottesville, 
    Virginia 22903-2498
        Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
    Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
    Virginia 23219
        NRC Project Director: David B. Matthews
    
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
    North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
    Virginia
    
        Date of amendment request: July 26, 1995
        Description of amendment request: The proposed change would revise 
    the Technical Specifications (TS) for the North Anna Power Station, 
    Units No. 1 and No. 2 (NA-1&2). Specifically, the change would clarify 
    the TS to allow switching of charging and low-head safety injection 
    pumps during unit shutdown conditions. The proposed changes would also 
    allow additional methods of rendering these same pumps incapable of 
    injecting into the reactor coolant system (RCS) when required for low-
    temperature conditions. NA-1&2 is equipped with three charging pumps. 
    These charging pumps provide inventory control, normal boration to the 
    RCS, and flow to the reactor coolant pump seals. They also act as the 
    high-head safety injection pumps during accident conditions. During 
    certain shutdown conditions, it is necessary to render two of the three 
    charging pumps inoperable to maintain the low-temperature overpressure 
    protection (LTOP) design bases assumptions. This provides assurance 
    that a mass addition pressure transient can be relieved by the 
    operation of a single pressurizer power-operated relief valve (PORV). 
    Low-temperature overpressure protection for each NA-1&2 unit is 
    provided by two pressurizer PORVs.
        During shutdown conditions, periodic surveillance testing of the 
    charging pumps is required by the NA-1&2 TS. Also during shutdown 
    conditions, it may be desirable to switch from one charging pump to 
    another to allow for other activities such as maintenance or testing.
        The current NA-1&2 TS associated with charging pumps during 
    shutdown conditions are very restrictive and do not allow sufficient 
    latitude for surveillance testing or pump switching. The current NA-1&2 
    TS specifically state in the surveillance requirements that the method 
    used to render a charging pump inoperable is to place the pump control 
    switch in the pull-to-lock position. This requirement would not allow 
    for surveillance or post-maintenance testing of the inoperable charging 
    pumps since this switch is used to start those pumps.
        Therefore, the licensee proposes to modify NA-1&2 TS to allow more 
    than one charging pump to be operable and capable of injecting into the 
    RCS for pump switching operations. Additionally, the methods used to 
    render charging pumps inoperable will be expanded to allow for post-
    maintenance and surveillance testing.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:
        Specifically, operation of North Anna Power Station in 
    accordance with the proposed Technical Specifications changes will 
    not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        Allowing more than one charging pump to be operable and capable 
    of injecting into the RCS during RCS low temperature operation for 
    pump switching for post-maintenance and surveillance testing does 
    not increase the probability of occurrence or the consequences of 
    any previously analyzed accident. Pump switching operations will be 
    under the direct administrative control of a licensed operator and 
    will only be for a short duration of time. Any situation that could 
    result in an excessive RCS mass addition would be immediately 
    recognized by the operator and remedial action would be taken to 
    prevent challenges to RCS integrity. Using methods such as opening 
    the charging pump power supply breaker or closing the charging pump 
    discharge valve(s) to render a charging pump inoperable will ensure 
    that these pumps will not be capable of injecting water into the 
    RCS. These alternate methods are as effective as placing the control 
    switches in the pull-to-lock position.
        2. Create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        Allowing more than one charging pump to be operable and capable 
    of injecting into the RCS during low-temperature operation for pump 
    switching for post-maintenance and surveillance testing does not 
    involve any physical modifications of the plant nor result in a 
    change in a method of operation. Licensed operator control of 
    charging pump switching operations will continue to ensure that the 
    RCS will not be challenged by excessive mass addition events. Using 
    methods other than placing charging pump control switches in the 
    pull-to-lock position to render the pump inoperable will still 
    ensure that only one pump will be capable of injecting into the RCS 
    during low temperature operations. Therefore, a new or different 
    type of accident is not made possible.
        3. Involve a significant reduction in a margin of safety.
        Allowing more than one charging pump to be operable and capable 
    of injecting into the RCS during RCS low temperature operation for 
    pump switching for post-maintenance and surveillance testing does 
    not affect any safety limits or limiting safety system settings. The 
    alternate methods of rendering pumps inoperable provide the same 
    level of assurance that the pump is incapable of flowing into the 
    RCS as placing the pump control switch in the pull-to-lock position. 
    System operating parameters remain unaffected. The availability of 
    equipment required to mitigate or assess the consequence of an 
    accident is not reduced. Safety margins are, therefore, not 
    decreased.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied.Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: The Alderman Library, Special 
    Collections Department, University of Virginia, Charlottesville, 
    Virginia 22903-2498
        Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
    Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
    Virginia 23219
        NRC Project Director: David B. Matthews
    
    Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
    Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
    
        Date of amendment request: July 20, 1995
        Description of amendment request: The proposed amendments would: 1) 
    revise three Reactor Protection System/Engineered Safety Features 
    Actuation Systems channel trip setpoint limits, 2) 
    
    [[Page 45191]]
    add a new setpoint limit for high high steam generator water level, and 
    3) incorporate editorial changes to revise the measurement units of one 
    setpoint limit and to delete certain references to two-loop operation.
        Basis for proposed no significant hazards consideration 
    determination: As requied by 10 CFR 50.91(a), the licensee has provided 
    its analysis of the issue of no significant hazards consideration, 
    which is presented below:Specifically, operation of Surry Power Station 
    with the proposed change will not:
        (1) Involve a significant increase in either the probability of 
    occurrence or consequences of any accident or equipment malfunction 
    scenario which is important to safety and which has been previously 
    evaluated in the Updated Safety Analysis Report (UFSAR). The effect 
    of the proposed change is to ensure that actual plant setpoints 
    remain conservative consistent with respect to accident analysis 
    assumptions. The proposed change requires safety system actuation 
    limits that are more conservative than those currently in Technical 
    Specifications. The change does not invalidate currently implemented 
    station setpoints or currently applicable accident analysis 
    assumptions regarding these setpoints. Consequently, the results and 
    conclusions of the current UFSAR accident analyses are not affected 
    by these changes. The proposed Technical Specifications change 
    revises setpoints used to mitigate accidents and therefore has no 
    bearing on the probability of an accident. Further, the change 
    ensures that the setpoints used to mitigate an accident bound the 
    setpoints used in the accident analyses. Therefore, the probability 
    of an accident or consequences of an accident is not adversely 
    affected as a result of this change.
        (2) Create the possibility of a new or different type of 
    accident than those previously evaluated in the UFSAR. Implementing 
    the proposed Technical Specifications setpoint limits cannot create 
    the possibility of an accident of a different type than was 
    previously evaluated in the UFSAR. Since actual plant setpoints are 
    not being affected, new accident precursors will not be introduced. 
    Furthermore, spurious challenges to safety systems are also not 
    expected to increase in frequency as a result of these changes since 
    actual setpoints installed in the plant are not being changed. 
    Consequently, no new accident precursors are created as a result of 
    the new Technical Specifications setpoint limits.
        (3) Involve a significant reduction in a margin of safety. Since 
    the results of the existing UFSAR accident analyses remain bounding, 
    safety margins are not impacted. The proposed Technical 
    Specifications setpoint limits ensure plant setpoints remain 
    conservative and consistent with design base accident analysis 
    assumptions including appropriate instrument channel uncertainties 
    due to harsh environmental conditions. Therefore, the margin of 
    safety as defined in the Technical Specifications bases is 
    unaffected.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Swem Library, College of 
    William and Mary, Williamsburg, Virginia 23185
        Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
    Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
    Virginia 23219
        NRC Project Director: David B. Matthews
    
    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
    Creek Generating Station, Coffey County, Kansas
    
        Date of amendment request: July 25, 1995
        Description of amendment request: This license amendment request 
    proposes to revise Technical Specification 4.0.5a and Bases Section 3/
    4.4.10 to delete the clause ``(g), except where specific written relief 
    has been granted by the Commission pursuant to 10 CFR Part 50, Section 
    50.55a(g)(6)(i).'' This proposed change is consistent with NUREG-1482, 
    ``Guidelines for Inservice Testing and Nuclear Power Plants.''
        Basis for proposed no significant hazards consideration 
    determination: As requied 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 would remove the wording ''...(g), except 
    where specific written relief has been granted by the Commission 
    pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
    Inspection and Testing Programs are described in the technical 
    specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
    change, in accordance with NUREG-1431 and NUREG-1482, would provide 
    relief to the ASME Code requirement in the interim between the time 
    of submittal of a relief request until the NRC has issued a safety 
    evaluation and granted the relief. The change being proposed is 
    administrative in nature and does not affect assumptions contained 
    in plant safety analyses, the physical design and/or operation of 
    the plant, nor does it affect any technical specification that 
    preserves safety analysis assumptions. Any relief from the approved 
    ASME Section XI Code requirements will require a 10 CFR 50.59 
    evaluation to ensure no technical specification changes or 
    unreviewed safety questions exist. Therefore, operation of the 
    facility in accordance with the proposed change would not affect the 
    probability or consequences of an accident previously analyzed.
        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 would remove the wording ''...(g), except 
    where specific written relief has been granted by the Commission 
    pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
    Inspection and Testing Programs are described in the technical 
    specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
    change, in accordance with NUREG-1431 and NUREG-1482, would provide 
    relief to the ASME Code requirement in the interim between the time 
    of submittal of a relief request until the NRC had issued a safety 
    evaluation and granted the relief. The change being proposed is 
    administrative in nature and will not change the physical plant or 
    the modes of operation defined in the facility license. The change 
    does not involve the addition or modification of equipment nor does 
    it alter the design or operation of plant systems. Any relief from 
    the approved ASME Section XI Code requirements will require a 10 CFR 
    50.59 evaluation to ensure no technical specification changes or 
    unreviewed safety questions exist. Therefore, operation of the 
    facility in accordance with the proposed change would 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 would remove the wording ''...(g), except 
    where specific written relief has been granted by the Commission 
    pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
    Inspection and Testing Programs are described in the technical 
    specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
    change, in accordance with NUREG-1431 and NUREG-1482, would provide 
    relief to the ASME Code requirement in the interim between the time 
    of submittal of a relief request until the NRC has issued a safety 
    evaluation and granted the relief. The change being proposed is 
    administrative in nature and will not alter the bases for assurance 
    that safety-related activities are performed correctly or the basis 
    for any technical specification that is related to the establishment 
    or maintenance of a safety margin. Any relief from the approved ASME 
    Section XI Code requirements will require a 10 CFR 50.59 evaluation 
    to ensure no technical specification changes or unreviewed safety 
    questions exist. Therefore, operation of the facility in accordance 
    with the proposed change would not involve a significant reduction 
    in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 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 locations: Emporia State University, 
    
    [[Page 45192]]
        William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
    66801 and Washburn University School of Law Library, Topeka, Kansas 
    66621
        Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
    Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
        NRC Project Director: William H. Bateman
    
    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.
    
    GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
    Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
    
        Date of amendment request: August 11, 1995
        Description of amendment request: The proposed amendment would 
    remove Technical Specification Section 3.2, ``Makeup and Purification 
    and Chemical Addition Systems,'' and its bases. The pertinent 
    requirements and bases applicable to these systems are being 
    incorporated in the TMI-1 Updated Final Safety Analysis Report (UFSAR).
        Date of publication of individual notice in Federal Register: 
    August 18, 1995 (60 FR 43172)
        Expiration date of individual notice: September 18, 1995
        Local Public Document Room location: Law/Government Publications 
    Section, State Library of Pennsylvania (REGIONAL DEPOSITORY) Walnut 
    Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
    
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
    Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    
        Date of application for amendments: April 6, 1995, and superseded 
    on August 7, 1995
        Description of amendments request: Amend the Sequoyah Nuclear 
    Plant, Units 1 and 2 Technical Specification (TS) to revise the 
    numerical values for the overtemperature and overpower delta-
    temperature equation constants in TS Table 2.2-1, Reactor Trip System 
    Instrumentation Trip Setpoints.
        Date of publication of individual notice in the Federal Register: 
    August 15, 1995 (60 FR 42187)
        Expiration date of individual notice: September 14, 1995
        Local Public Document Room location: Chattanooga-Hamilton County 
    Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    
    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.
    
    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
    50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
    Units 1, 2, and 3, Maricopa County, Arizona
    
        Date of application for amendments: May 2, 1995
        Brief description of amendments: The amendments remove from the 
    technical specifications (TS) plant elevations for the minimum water 
    volume required in the spent fuel pool and relocate them to site 
    procedures. The TS amendment also includes two changes to correct 
    administrative errors in the TS.
        Date of issuance: August 7, 1995
        Effective date: August 7, 1995
        Amendment Nos.: Unit 1 - Amendment No. 97 ; Unit 2 - Amendment No. 
    85; Unit 3 - Amendment No. 68
        Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
    amendments revised the Technical Specifications.
        Date of initial notice in Federal Register: July 5, 1995 (60 FR 
    35060) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated August 7, 1995.No significant 
    hazards consideration comments received: No.
        Local Public Document Room location: Phoenix Public Library, 12 
    East McDowell Road, Phoenix, Arizona 85004
    
    Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
    324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
    County, North Carolina
    
        Date of application for amendments: January 25, 1993, as 
    supplemented on December 28, 1993, September 13, 1994, January 13, 
    1995, and May 25, 1995. The supplemental submittals did not expand the 
    scope of the original Federal Register notice or change the no 
    significant hazards determination.
        Brief description of amendments: The amendments allow unit entry 
    into Operational Condition 1 (Power Operation) from Operational 
    Condition 2 (Startup) with up to eight inoperable control rods, 
    provided those control rods are not inoperable due to being immovable 
    or untrippable.
        Date of issuance: August 11, 1992
        Effectove date: August 11, 1992 
    
    [[Page 45193]]
    
        Amendment Nos.: 178 and 209
        Facility Operating License Nos. DPR-71 and DPR-62.
        Date of initial notice in Federal Register: July 7, 1993 (58 FR 
    36428) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated August 11, 1995.Significant 
    hazards consideration comments received: No.
        Local Public Document Room location: University of North Carolina 
    at Wilmington, William Madison Randall Library, 601 S. College Road, 
    Wilmington, North Carolina 28403-3297
    
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
    Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
    North Carolina
    
        Date of application for amendment: April 5, 1995, as supplemented 
    July 31, 1995
        Brief description of amendment: The amendment revises various 
    portions of TS 3/4.9, Refueling Operations, to be consistent with 
    NUREG-1431, ``Standard Technical Specifications, Westinghouse Plants,'' 
    and allows the relocation of applicable sections from the TS that do 
    not meet the Commission screening criteria for retention.
        Date of issuance: August 9, 1995
        Effectove date: August 9, 1995
        Amendment No.: 61
        Facility Operating License No. NPF-63. Amendment revises the 
    Technical Specifications
        Date of initial notice in Federal Register: May 10, 1995 (60 FR 
    24906) The July 31, 1995 letter 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 August 9, 1995.No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Cameron Village Regional 
    Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    
    Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
    County Station, Units 1 and 2, LaSalle County, Illinois
    
        Date of application for amendments: January 13, 1995
        Brief description of amendments: The amendments revise the pressure 
    alarm setpoint allowable values for the emergency core cooling system 
    (ECCS) and reactor core isolation cooling (RCIC) system ``keep filled'' 
    pressure instrumentation channels. The purpose of the change is to 
    lower the setpoint allowable values for these parameters to more 
    realistic values based upon calculations performed by the licensee 
    reflecting design changes and system performance. Also, the term 
    ``setpoint'' is being changed to ``setpoint allowable value'' to 
    clarify the use of the values. Additionally, two administrative/
    editorial changes are included to delete technical specification 
    footnotes which are no longer applicable.
        Date of issuance: August 15, 1995
        Effectove date: Immediately, to be implemented within 90 days.
        Amendment Nos.: 105 and 91
        Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: March 1, 1995 (60 FR 
    11128) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated August 15, 1995.No significant 
    hazards consideration comments received: No
        Local Public Document Room location:  Jacobs Memorial Library, 
    Illinois Valley Community College, Oglesby, Illinois 61348
    
    Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
    Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
    
        Date of application of amendments: February 24, 1994, as 
    supplemented by letters dated April 19, May 25, August 25, 1994, 
    January 4, January 27, February 22, March 15, April 19, and May 31, 
    1995
        Brief description of amendments: The amendments provide 
    surveillance requirements for a planned modification to the Keowee 
    emergency power generators' underground power path breaker closing 
    logic.
    
        Date of issuance:  August 15, 1995
        Effectove date: As of the date of issuance to be implemented within 
    30 days
        Amendment Nos.: 210, 210, and 207
        Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
    amendments revised the Technical Specifications.
        Date of initial notice in Federal Register: March 30, 1994 (59 FR 
    14887) The April 19, May 25, August 25, 1994, January 4, January 27, 
    February 22, March 15, April 19, and May 31, 1995, letters provided 
    clarifying information that did not change the scope of the February 
    24, 1994, application and initial no proposed significant hazards 
    consideration determination. The Commission's related evaluation of the 
    amendments is contained in a Safety Evaluation dated August 15, 1995.No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Oconee County Library, 501 
    West South Broad Street, Walhalla, South Carolina 29691
    
    Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
    Turkey Point Plant Units 3 and 4, Dade County, Florida
    
        Date of application for amendments: March 30, 1995, as supplemented 
    May 5, 1995 and June 19, 1995
        Brief description of amendments: These amendments relate to 
    separation of the 24-hour emergency diesel generator test and hot 
    restart test from the loss of offsite power test.
        Date of issuance: August 8, 1995
        Effectove date: August 8, 1995
        Amendment Nos.: 175 and 169Facility Operating Licenses Nos. DPR-31 
    and DPR-41: Amendments revised the Technical Specifications.
        Date of initial notice in Federal Register: May 23, 1995 (60 FR 
    27339), and July 5, 1995 (60 FR 35072) The Commission's related 
    evaluation of the amendments is contained in a Safety Evaluation dated 
    August 8, 1995.No significant hazards consideration comments received: 
    No
        Local Public Document Room location: Florida International 
    University, University Park, Miami, Florida 33199
    
    GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
    Nuclear Generating Station, Ocean County, New Jersey
    
        Date of application for amendment: April 15, 1995, as supplemented 
    by letters on May 20, 1994, and March 8, 1995
        Brief description of amendment: The amendment revises Technical 
    Specification Section 6.5.3, ``AUDITS,'' by removing the specified 
    frequency for internal audits. These frequency specifications will now 
    be located in Appendix E of the GPU Nuclear Operational Quality 
    Assurance Plan (1000-PLN-7200.01). A minor editorial change has been 
    incorporated into TS 6.5.1.14 correcting a reference in response to a 
    finding in the Operational Safety Team Inspection (OSTI) report of 
    December 23, 1993.
        Date of issuance: August 7, 1995
        Effectove date: As of the date of issuance to be implemented within 
    30 days
        Amendment No.: 181
        Facility Operating License No. DPR-16. Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 25, 1994 (59 FR 
    27056) 
    
    [[Page 45194]]
    The letters of May 20, 1994, and March 8, 1995, provided clarifying 
    information that did not change the initial proposed no significant 
    hazards consideration determination. The Commission's related 
    evaluation of this amendment is contained in a Safety Evaluation dated 
    August 7, 1995.No significant hazards consideration comments received: 
    No.
        Local Public Document Room location: Ocean County Library, 
    Reference Department, 101 Washington Street, Toms River, NJ 08753
    
    GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
    Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
    
        Date of application for amendment: April 19, 1994, supplemented 
    March 8, 1995
        Brief description of amendment: The amendment revises the TMI-1 
    Technical Specification (TS) Section 6.5.3 to remove the specified 
    frequency of various licensee-conducted audits, including those related 
    to quality assurance, fire protection, security, emergency 
    preparedness, and offsite dose calculations. The frequencies for 
    conduct of these audits will now be specified in the licensee's 
    Operational Quality Assurance Plan, which requires NRC approval for 
    significant changes. The Commission has determined that these audit 
    frequencies need not be in the TS to assure public health and safety.
        Date of issuance: August 14, 1995
        Effectove date: August 14, 1995
        Amendment No.: 195
        Facility Operating License No. DPR-50. Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 8, 1994 (59 FR 
    29627) The March 8, 1995, submittal provided clarifying information 
    that did not change the initial proposed no significant hazards 
    consideration determination. The Commission's related evaluation of 
    this amendment is contained in a Safety Evaluation dated August 14, 
    1995.No significant hazards consideration comments received: No.
        Local Public Document Room location: Law/Government Publications 
    Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
    Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
    
    Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
    Point Nuclear Station, Unit 2, Oswego County, New York
    
        Date of application for amendment: December 13, 1994, as 
    supplemented April 3, 1995
        Brief description of amendment: The amendment revises Table 
    3.6.1.2-1 to allow a maximum leakage of 24.0 scfh for each of the 8 
    main steam isolation valves instead of the current 6.0 scfh.
        Date of issuance: August 10, 1995
        Effectove date: As of the date of issuance to be implemented within 
    60 days
        Amendment No.: 67
        Facility Operating License No. NPF-69: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: January 18, 1995 (60 FR 
    3675) The April 3, 1995, letter provided clarifying information that 
    did not change the initial no proposed significant hazards 
    consideration determination. The Commission's related evaluation of the 
    amendment is contained in a Safety Evaluation dated August 10, 1995.No 
    significant hazards consideration comments received: No
        Local Public Document Room location: Reference and Documents 
    Department, Penfield Library, State University of New York, Oswego, New 
    York 13126
    
    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: March 15, 1995 (published in 
    Federal Register as March 15, 1994) as supplemented by letter dated 
    August 5, 1995
        Brief description of amendments: These amendments modify the 
    Susquehanna Steam Electric Station Technical Specification Table 3.6.3-
    1, Primary Containment Isolation Valves, concerning the scope of Type C 
    testing on specified emergency core cooling system and reactor core 
    isolation cooling containment isolation valves. Specifically, the 
    subject valves on systems which terminate below the minimum water level 
    of the suppression pool will no longer require Type C testing but will 
    instead be tested using requirements of the American Society of 
    Mechanical Engineers' Section XI Code.
        Date of issuance: August 15, 1995
        Effectove date: August 15, 1995
        Amendment Nos.: 149 and 119
        Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
    revised the Technical Specifications.The supplemental letter did not 
    change the proposed no significant hazards consideration determination 
    nor the Federal Register notice.
        Date of initial notice in Federal Register: April 26, 1995 (60 FR 
    20521) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated August 15, 1995.No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
    
    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: March 31, 1995, as supplemented 
    by letter dated June 22, 1995
        Brief description of amendments: These amendments delete from the 
    Technical Specifications of each unit, the operational condition 
    restriction in Surveillance Requirement 4.8.1.1.2.d.7, which requires 
    that 24-hour emergency diesel generator testing be performed with at 
    least one unit in operational condition 4 or 5 (cold shutdown or 
    refueling).
        Date of issuance:  August 15, 1995
        Effectove date: Units 1 and 2, effective as of the date of issuance 
    and shall be implemented within 60 days
        Amendment Nos.:  150 and 120
        Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: April 26, 1995 (60 FR 
    20523) The Commission's related evaluation of the amendments is 
    contained in a Safety Evaluation dated August 15, 1995.No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
    
    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: November 21, 1994, as 
    supplemented by letters dated February 21, 1995, March 28, 1995, April 
    10, 1995, May 24, 1995, and June 23, 1995
        Brief description of amendments: These amendments change the 
    Technical Specifications for the two units by deleting reference to the 
    main steamline isolation valve (MSIV) leakage control system and its 
    associated primary containment isolation valves, and increase the 
    allowable leakage rate for any MSIV and the total maximum 
    
    [[Page 45195]]
    pathway leakage for all four main steam lines.
        Date of issuance: August 15, 1995
        Effectove date: Units 1 and 2 as of date of issuance and shall be 
    implemented within 30 days
        Amendment Nos.: 151 and 121
        Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: January 4, 1995 (60 FR 
    503) The Commission's related evaluation of the amendments is contained 
    in a Safety Evaluation dated August 15, 1995.No significant hazards 
    consideration comments received: No
        Local Public Document Room location: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
    
    Power Authority of the State of New York, Docket No. 50-333, James 
    A. FitzPatrick Nuclear Power Plant, Oswego County, New York
    
        Date of application for amendment: March 2, 1995
        Brief description of amendment: The amendment extends the 
    surveillance test intervals for the snubber systems to support 24-month 
    operating cycles. Surveillance test interval extensions are denoted as 
    being performed ``every 24 months'' or ``at least once per 24 months'' 
    consistent with the guidance provided in Generic Letter (GL) 91-04, 
    ``Changes in Technical Specification Surveillance Intervals to 
    Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff 
    has determined that the proposed Technical Specification changes are in 
    accordance with GL 91-04, and are, therefore, acceptable.
        Date of issuance: August 8, 1995
        Effectove date: As of the date of issuance to be implemented within 
    30 days
        Amendment No.: 226
        Facility Operating License No. DPR-59: Amendment revised the 
    Technical Specifications.
        Date of initial notice in Federal Register: May 10, 1995 (60 FR 
    24916) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated August 8, 1995.No significant 
    hazards consideration comments received: No
        Local Public Document Room location: Reference and Documents 
    Department, Penfield Library, State University of New York, Oswego, New 
    York 13126
    
    Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
    311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
    County, New Jersey
    
        Date of application for amendments: February 5, 1993, supplemented 
    April 13, June 11 and November 17, 1993
        Brief description of amendments: The amendment eliminates the 
    Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level 
    Reactor Trip due to the installation of the digital feedwater control 
    system incorporating a median signal selector.
        Date of issuance: August 7, 1995
        Effectove date: Unit 1, as of the date of issuance, to be 
    implemented by the startup following the twelfth refueling outage, Unit 
    2, as of the date of issuance, to be implemented by the startup 
    following the current outage
        Amendment Nos.: 173 and 154
        Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: April 28, 1993 (58 FR 
    25864) The April 13, June 11, and November 17, 1993 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 August 7, 1995.No significant hazards consideration comments 
    received: No
        Local Public Document Room location: Salem Free Public Library, 112 
    West Broadway, Salem, New Jersey 08079
    
    South Carolina Electric & Gas Company, South Carolina Public 
    Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
    Station, Unit No. 1, Fairfield County, South Carolina
    
        Date of amendment request: March 11, 1994
        Description of amendment request: The amendment decreases the 
    allowable time for operation with one inoperable residual heat removal 
    (RHR) relief valve from 7 days to 72 hours. This amendment request has 
    been submitted in response to Generic Issue 94 as discussed in Generic 
    Letter 90-06.
        Date of issuance: August 11, 1995
        Effectove date: August 11, 1995
        Amendment No.: 125
        Facility Operating License No. NPF-12: Amendment revises the 
    Technical Specifications.
        Date of initial notice in Federal Register: June 22, 1994 (59 FR 
    32236) The Commission's related evaluation of the amendment is 
    contained in a Safety Evaluation dated August 11, 1995.No significant 
    hazards consideration comments received: No
        Local Public Document Room location:  Fairfield County Library, 300 
    Washington Street, Winnsboro, South Carolina 29180
    
    TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
    Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
    
        Date of amendment request: February 14, 1994 (TXX-94045), as 
    supplemented by letter dated May 23, 1995 (TXX-95147)
        Brief description of amendments: The amendments incorporated 
    appropriate references to and provisions of the new 10 CFR Part 20 
    regulations. These changes revised a definition and aspects of 
    radiological effluent technical specifications, clarified the 
    administrative specification for reporting individual annual exposures 
    greater than 100 mrem by work/job function, and revised the 
    administrative specifications for providing alternative measures for 
    control of access to high radiation areas and designating record 
    retention for radioactive shipments.
        Date of issuance: August 11, 1995
        Effectove date: August 11, 1995
        Amendment Nos.: Unit 1 - Amendment No. 42; Unit 2 - Amendment No. 
    28
        Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
    revised the Technical Specifications.
        Date of initial notice in Federal Register: April 28, 1994 (59 FR 
    22016) The additional information contained in the supplemental letter 
    dated May 23, 1995, was clarifying in nature and thus, within the scope 
    of the initial notice and did not affect the staff's proposed no 
    significant hazards consideration determinations. The Commission's 
    related evaluation of the amendments is contained in a Safety 
    Evaluation dated August 11, 1995.No significant hazards consideration 
    comments received: No.
        Local Public Document Room location: University of Texas at 
    Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
    19497, Arlington, TX 76019
    
    Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
    Vermont Yankee Nuclear Power Station, Vernon, Vermont
    
        Date of application for amendment: March 31, 1994, as supplemented 
    by letters dated September 9, 1994, and June 22, 1995
        Brief description of amendment: The amendment modifies the 
    requirements for avoidance and protection from thermal hydraulic 
    instabilities to be 
    
    [[Page 45196]]
    consistent with the Boiling Water Reactor (BWR) Owners Group long-term 
    solution Option I-D described in the Licensing Topical Report, ``BWR 
    Owners Group Long-Term Stability Solutions Licensing Methodology, NEDO-
    31960 June 1991'' and NEDO-31960, Supplement 1, Dated March 1992. NEDO-
    31960 and NEDO-31960, Supplement 1, were accepted by the NRC staff in a 
    letter to L.A. England (BWR Owners Group) dated July 12, 1993.
    
        Date of issuance: August 9, 1995
    
        Effectove date: As of the date of issuance to be implemented within 
    30 days
    
        Amendment No.: 146
    
        Facility Operating License No. DPR-28. Amendment revised the 
    Technical Specifications.
    
        Date of initial notice in Federal Register: January 4, 1995 (60 FR 
    507) The September 9, 1994, and June 22, 1995, 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 August 9, 1995. No significant hazards consideration comments 
    received: No
    
        Local Public Document Room location: Brooks Memorial Library, 224 
    Main Street, Brattleboro, VT 05301
    
        Dated at Rockville, Maryland, this 23rd day of August.
    
        For The Nuclear Regulatory Commission
    
    Elinor G. Adensam,
    Acting Director, Division of Reactor Projects III/IV, Office of Nuclear 
    Reactor Regulation
    
    [Doc. 95-21389 Filed 8-29-95; 8:45 am]
    
    BILLING CODE 7590-01-F
    
    

Document Information

Effective Date:
8/7/1995
Published:
08/30/1995
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
X95-10830
Dates:
August 7, 1995
Pages:
45172-45196 (25 pages)
PDF File:
x95-10830.pdf