X95-20329. Biweekly Notice  

  • [Federal Register Volume 60, Number 60 (Wednesday, March 29, 1995)]
    [Notices]
    [Pages 16181-16196]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: X95-20329]
    
    
    
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    NUCLEAR REGULATORY COMMISSION
    
    Biweekly Notice
    
    Applications and Amendments to Facility Operating LicensesInvolving 
    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 March 3, 1995, through March 17, 1995. The 
    last biweekly notice was published on March 15, 1995.
    
    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 
    [[Page 16182]] 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 April 28, 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., 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 
    [[Page 16183]] 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.
    
    Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
    Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
    
        Date of amendment request: February 24, 1995
        Description of amendment request: The proposed change would remove 
    Section 4.3 from the Technical Specifications (TS) because the primary 
    system testing following opening is already performed in accordance 
    with the American Society of Mechanical Engineers Boiler and Pressure 
    Vessel Code, as implemented in the licensee's inservice inspection 
    program as required by TS 4.0.1.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        This change does not involve a significant hazards consideration 
    for the following reasons.
        1. The requested change does not involve a significant increase 
    in the probability or consequences of an accident previously 
    evaluated. This requested change will provide consistency between 
    our Technical Specifications (TS) and 10 CFR 50.55a which requires 
    testing in accordance with Section XI of the ASME Boiler and 
    Pressure Vessel Code. The requirements contained in TS Section 4.3 
    were placed into TS prior to incorporation of Section XI into the 
    ASME Boiler and Pressure Vessel Code. The NRC and industry have 
    since recognized the ASME Boiler and Pressure Vessel Code, Section 
    XI as the appropriate testing program. Adequate assurance of primary 
    system integrity will be provided since primary system testing will 
    continue to be controlled and performed in accordance with the rules 
    for inservice inspections provided by ASME Boiler and Pressure 
    Vessel Code, Section XI as implemented by our approved In-Service 
    Inspection (ISI) Program, as required by TS Section 4.0.1. 
    Therefore, there would be no increase in the probability or 
    consequences of an accident previously evaluated.
        2. The requested change does not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated. The requested change deletes the current TS requirements 
    for primary system testing by recognizing that we will continue to 
    perform required testing consistent with 10 CFR 50.55a and ASME 
    Boiler and Pressure Vessel Code, Section XI, as implemented by our 
    approved ISI Program, as required by TS Section 4.0.1. This 
    requested change does not involve the addition or modification of 
    plant equipment, nor does it alter the design or operation of plant 
    systems. Therefore, the requested change does not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        3. The requested change does not involve a significant reduction 
    in a margin of safety. The requested change deletes the current TS 
    Section 4.3 requirements for primary system testing and maintains 
    the margin of safety by continuing to perform required testing in 
    accordance with 10 CFR 50.55a and ASME Boiler and Pressure Vessel 
    Code, Section XI, as implemented by our approved ISI Program, as 
    required by TS Section 4.0.1. Therefore, the requested change does 
    not involve a significant reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Hartsville Memorial Library, 
    147 West College Avenue, Hartsville, South Carolina 29550
        Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
    & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
        NRC Project Director: William H. Bateman
    
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
    Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
    North Carolina
    
        Date of amendment request: March 3, 1995
        Description of amendment request: The proposed amendment would 
    eliminate the requirement to perform periodic measurement testing of 
    the response times for selected pressure and differential pressure 
    sensors. The requirement that reactor trip and engineered safety 
    feature response time functions be within their specified limit at 
    least once per 18 months will be verified instead of demonstrated. The 
    associated bases section for response time requirements will be changed 
    to allow the sensor response time portion of the channel response time 
    to use historical records, testing results, or vendor supplied 
    engineering specifications. No other changes to response time methods 
    are included in this change.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        1. The proposed amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        The proposed amendment does not result in a condition where the 
    design, material, or construction standards that were
        applicable prior to the change are altered nor does it modify 
    any system interface. The same Reactor Trip System and Engineered 
    Safety Features Actuation System instrumentation is being used; the 
    time response allocations/modeling assumptions in the Final Safety 
    Analysis Report (FSAR) Chapter 15 analyses are still the same; only 
    the method of verifying time response is changed. The proposed 
    activity will not change, degrade, or prevent actions or alter any 
    assumptions previously made in evaluating the radiological 
    consequences of an accident described in the FSAR. Therefore, there 
    would be no increase in the probability or consequences of an 
    accident previously evaluated.
        2. The proposed amendment does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        The proposed amendment does not alter the performance of the 
    pressure and the differential pressure transmitters used in the 
    plant protection systems. The sensors will still have response time 
    verified by test before placing the sensor in operational service 
    and after any maintenance that could affect response time. Changing 
    the method of periodically verifying instrument response for certain 
    sensors (assuring equipemt operable) from time response testing to 
    calibration and channel checks will not create any new accident 
    initiators or scenarios. Periodic surveillance of these instruments 
    will detect significant degradation in the sensor response 
    characteristic. Therefore, the proposed changes do not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated. [[Page 16184]] 
        3. The proposed amendment does not involve a significant 
    reduction in the margin of safety.
        The proposed amendment to [sic] does not affect the total system 
    response time assumed in the safety analysis. The periodic system 
    response time verification method for selected pressure and 
    differential pressure sensors is modified to allow use of actual 
    test data or engineering data. The method of verification still 
    provides assurance that the total system response is within that 
    defined in the safety analysis, since calibration tests will detect 
    any degradation which might significantly affect sensor response 
    time. 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 request involves no significant hazards consideration.
        Local Public Document Room location: Cameron Village Regional 
    Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
        Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
    & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
        NRC Project Director: William H. Bateman
    
    Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
    Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
    STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
    Will County, Illinois
    
        Date of amendment request: May 20, 1994, as supplemented February 
    2, 1995
        Description of amendment request: The proposed amendment would 
    permit the licensee to use an alternate repair criteria (ARC), 
    designated as the F* criteria. Use of the F* criteria would 
    allow tubes with otherwise pluggable indications, to remain in service 
    as long as the indications are below the designated minimum distance of 
    the F* criteria. The F* criteria for Byron and Braidwood 
    defines a length of 1.7 inches of undegraded expanded tube within the 
    tubesheet as the minimum distance acceptable for implementing the ARC. 
    Below the F* length, a circumferential tube defect can exist and 
    the tube can remain in service. The proposed amendment will change the 
    plugging limit definition and would exclude plugging steam generator 
    tubes with indications that satisfy the F* criteria. The F* 
    criteria maintains the structural integrity of the degraded tube as the 
    primary pressure boundary and allows the tube to remain in service for 
    heat transfer and core cooling.
        This alternate repair criteria qualification is documented in 
    Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P 
    Revision 1, ``W-D4 F* Qualification Report,'' which is included as 
    part of the licensee's submittal. The staff's proposed no significant 
    hazards consideration determination for the requested change was 
    published on July 6, 1994 (59 FR 34659). In response to the staff's 
    request for additional information by letter dated February 2, 1995, 
    the licensee revised their previous submittal.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        A. The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident previously evaluated.
        The supporting qualification report for subject criteria 
    demonstrates that the presence of the tubesheet will enhance the 
    tube integrity in the region of the tube-to-tubesheet roll 
    expansions by precluding tube deformation beyond its initial 
    expanded outside diameter. The resistance to a tube rupture is 
    strengthened by the presence of the tubesheet in that region. The 
    results of hardrolling of the tube into tubesheet provides a 
    mechanical leak limiting seal between the tube and the tubesheet. A 
    tube rupture cannot occur because the contact between the tube and 
    the tubesheet does not permit sufficient movement of tube material.
        The type of degradation for which the F* criteria has been 
    developed (cracking with a circumferential orientation) can 
    theoretically lead to a postulated tube rupture event provided that 
    the postulated through-wall circumferential crack exists near the 
    top of the tubesheet. An evaluation including analysis and testing 
    has been done to determine the resistive strength of the expanded 
    tubes within the tubesheet. This evaluation provides the basis for 
    the acceptance criteria for tube degradation subject to the F* 
    criteria. The F* length of roll expansion is sufficient to 
    preclude tube pullout from tube degradation located below the 
    F* distance, regardless of the extent of the tube degradation. 
    The Technical Specification leakage rate requirements and accident 
    analysis assumptions remain unchanged in the unlikely event that 
    significant leakage from this region does occur. For consistency 
    with current offsite dose limits, the site allowable leakage limit 
    during a MSLB has been conservatively calculated to be 12.8 gpm for 
    Byron and 9.1 gpm for Braidwood, which includes the accident leakage 
    from IPC in addition to the accident leakage from F* on the 
    faulted steam generator and the operational leakage limit. The 
    operational leakage limit of Specification 3.4.6.2.c in each of the 
    three remaining intact steam generators shall include the 
    operational leakage from F*. As a requirement for operation 
    following application IPC, the projected distribution of crack 
    indications over the operating period must be verified to result in 
    primary to secondary accident leakage less than the site allowable 
    leakage limit. Thus, the consequences of a MSLB remain unchanged.
        The tube rupture and pullout is fully bounded by the existing 
    steam generator tube rupture analysis included in the UFSAR. The 
    leakage testing of the roll expanded tubes indicates that for tube 
    expansion lengths approximately equal to the * distance, any 
    postulated primary to secondary leakage from * tubes would be 
    insignificant. The proposed alternate repair criteria does not 
    adversely impact any other previously evaluated design basis 
    accident.
        The leakage from an F* tube would be limited by the tube-
    to-tubesheet interface since this leak would occur below the 
    secondary face of the tubesheet. Qualification testing and previous 
    experience indicate that normal and faulted leakage is well below 
    Technical Specification and administrative limits creating no 
    increase in the consequences associated with tube rupture type 
    leakages. The UFSAR analyzed accident scenarios are still bounding 
    since the normal and faulted leak rates are well within the normal 
    operating limit of 150 gallons per day. This conclusion is 
    consistent with previous F* programs approved and used at other 
    operating plants.
        All of the design and operating characteristics of the steam 
    generator and connected systems are preserved since the F* 
    criteria utilizes the ``as rolled'' tube configuration that exists 
    as part of the original steam generator design. The F* joint 
    has been analyzed and tested for design, operating, and faulted 
    condition loadings in accordance with Regulatory Guide 1.121 safety 
    factors. The potential for a tube rupture is not increased from the 
    original submittal as demonstrated in the qualification analyses and 
    testing completed in the BWNT report.
        Therefore, this change does not 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 type of accident from any accident previously 
    evaluated.
        Implementation of the proposed F* criteria does not 
    introduce any changes to the plant design basis. Use of the criteria 
    does not provide a mechanism to initiate an accident outside of the 
    region of the expanded portion of the tube. In the unlikely event 
    the failed tube severed completely at a point below the F* 
    region, the remaining F* joint would retain engagement in the 
    tubesheet due to its length of expanded contact within the tubesheet 
    bore. This engagement length would prevent any interaction of the 
    severed tube with neighboring tubes. Any hypothetical accident as a 
    result of any tube degradation in the expanded region of the tube 
    would be bounded by the existing tube rupture accident analysis. 
    Tube bundle structural integrity will be maintained. Tube bundle 
    [[Page 16185]] leak tightness will be maintained such that any 
    postulated accident leakage from F* tubes will be negligible 
    with regard to offsite doses.
        Therefore, there is not a potential for creating the possibility 
    of a new or different type of accident from any accident previously 
    evaluated.
        C. The proposed changes do not involve a significant reduction 
    in a margin of safety.
        The use of the F* criteria has been demonstrated to 
    maintain the integrity of the tube bundle commensurate with the 
    requirements of Regulatory Guide 1.121 and the primary to secondary 
    pressure boundary under normal and postulated accident conditions. 
    Acceptable tube degradation for the * criteria is any 
    degradation indication in the tubesheet region, more than the 
    F* distance from the secondary face of the tubesheet or the top 
    of the last hardroll contact point whichever is further into the 
    tubesheet. The safety factors used in the verification of the 
    strength of the degraded tube are consistent with the safety factors 
    in the ASME Boiler and Pressure Vessel Code and Regulatory Guide 
    1.121 used in steam generator design. The * distance has been 
    verified by various testing to be greater than the length of the 
    roll expanded tube-to-tubesheet interface required to preclude both 
    tube pullout and significant leakage during normal and postulated 
    accident conditions. The protective boundaries of the steam 
    generator continue to be maintained with the use of the F* 
    criteria. A tube with the indication of degradation previously 
    requiring removal from service can be kept in service through the 
    F* criteria. Since the joint is contained within the tubesheet 
    bore, there is no additional risk associated with the previously 
    analyzed tube rupture event. The leak testing acceptance criteria 
    are based on the primary to secondary leakage limit in the Technical 
    Specifications and the leakage assumptions used in the UFSAR 
    accident analyses.
        Implementation of the alternate repair criteria will decrease 
    the number of tubes which must be taken out of service with tube 
    plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS 
    flow margin; thus, implementation of the F* criteria will 
    maintain the margin of flow that would otherwise be reduced in the 
    event of increased plugging or sleeving.
        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, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
    Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
    Street, Wilmington, Illinois 60481.
        Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
    Austin, One First National Plaza, Chicago, Illinois 60690
        NRC Project Director: Robert A. Capra
    
    Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee 
    Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
    
        Date of amendment request: November 22, 1994, as supplemented 
    January 30, March 2, and March 13, 1995.
        Description of amendment request: This request was previously 
    published in the Federal Register on February 15, 1995 (60 FR 8746). It 
    is being renoticed to provide clarification to the scope of the 
    original request. The amendments would revise Technical Specification 
    (TS) 3.8 to establish restricted loading patterns and associated burnup 
    criteria for placing fuel in the Oconee spent fuel pools. In addition, 
    the Design Features sections associated with the reactor and fuel 
    storage would be revised. These changes are necessary to address two 
    new fuel designs which have increased initial fuel enrichment and 
    therefore cannot be stored in the spent fuel pools under existing TS or 
    loaded into the reactor. An administrative change would be made to TS 
    6.9.1 to include spent fuel pool boron concentration in the Core 
    Operating Limits Report. Other administrative changes would be made in 
    the Design Features section to make the specification consistent with 
    wording in the standard TS. Finally, the two additional supplements to 
    the original request are referenced herein.
        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:
        Standard 1. The proposed amendments will not involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        Each accident analysis addressed in the Oconee Final Safety 
    Analysis Report (FSAR) has been examined with respect to changes in 
    Cycle 15 parameters to determine the effect of the Cycle 16 reload 
    and to ensure that the acceptance criteria of the FSAR safety 
    analyses remain satisfied. The transient evaluation of Cycle 16 is 
    considered to be bounded by previously accepted analyses. Section 7 
    of the Reload Report addresses ``Accident and Transient Analysis'' 
    for this core reload.
        There is no increase in the probability or consequences of an 
    accident due to the spent fuel storage restrictions proposed in this 
    amendment request. It has been shown that the calculated, worst case 
    keff for this area is [less than or equal to] 0.95 under all 
    conditions. There is no increase in the probability of a fuel drop 
    accident in the SFP [spent fuel pool] since the mass of the new 
    assemblies is not significantly different from the mass of the old 
    assemblies. The likelihood of other accidents, previously evaluated 
    and described in the FSAR, is also not affected by the proposed 
    changes. In fact, it could be postulated that since the increase in 
    fuel enrichment will allow for extended fuel cycle lengths, there 
    will be a decrease in fuel movement and the probability of an 
    accident may actually be reduced. There is also no increase in the 
    consequences of a fuel rod drop accident in the SFP since the 
    fission product inventory of individual fuel assemblies will not 
    change significantly as a result of increasing the initial 
    enrichment. In addition, no change to safety related systems is 
    being made. Therefore, the consequences of a fuel rupture accident 
    remain unchanged. In addition, it has been shown that Keff all 
    conditions. Therefore, the consequences of a criticality accident in 
    the SFP remain unchanged as well. The above analysis ensures that 
    the proposed reload amendment request will not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        2. The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        The analyses performed in support of this reload are in 
    accordance with the NRC approved methods delineated in Specification 
    6.9.2. The predicted operating characteristics of Oconee 3 Cycle 16 
    are similar to previously licensed designs. The Mark B10T and Mark 
    B11 fuel assembly designs remain mechanically compatible with all 
    fuel handling equipment. Therefore, no new or different kind of fuel 
    handling accident is created by the proposed amendment request.
        Section 15.11 of the Oconee FSAR states that the refueling boron 
    concentration is maintained such that a criticality accident during 
    refueling is not considered credible. The proposed amendment request 
    continues to assure that a criticality accident in the SFP or during 
    refueling is not credible. The double contingency principle 
    discussed in ANSI N-16.1-1975 and the April 1978 NRC letter allows 
    credit for soluble boron under other abnormal or accident 
    conditions, since only a single accident need be considered at one 
    time. Thus, by requiring a minimum boron concentration in the SFP, a 
    criticality accident caused by violating the SFP storage 
    restrictions is not considered credible. Therefore, the proposed 
    amendment request does not create the possibility of a new or 
    different kind of accident from any accident previously evaluated.
        3. The proposed changes do not involve a significant reduction 
    in the margin of safety.
        The Oconee 3 Cycle 16 design was performed using the NRC 
    approved methods given in Specification 6.9.2. The safety limits for 
    Oconee 3 Cycle 16 are unchanged from previous cycles. The limits and 
    margins summarized in the Oconee 3 Cycle 16 Reload Report are well 
    within the allowable limits and requirements, and reflect no 
    reductions to any margins of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    [[Page 16186]] satisfied. Therefore, the NRC staff proposes to 
    determine that the amendment request involves no significant hazards 
    consideration.
        Local Public Document Room location: Oconee County Library, 501 
    West South Broad Street, Walhalla, South Carolina 29691
        Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
    1200 17th Street, NW., Washington, DC 20036
        NRC Project Director: Herbert N. Berkow
    
    Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
    ElectricStation, Unit 3, St. Charles Parish, Louisiana
    
        Date of amendment request: December 14, 1993, as supplemented by 
    letter dated March 3, 1995.
        Description of amendment request: The proposed amendment would 
    revise the Technical Specifications (TSs) by removing the reactor 
    vessel material specimen withdrawal schedule and by updating the 
    reactor coolant system pressure-temperature (P-T) curves. The specimen 
    withdrawal schedule will be relocated to the Updated Final Safety 
    Analysis Report (UFSAR). The original Notice was published on January 
    19, 1994 (59 FR 2867).
        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:
        Although the Reactor Vessel material specimens withdrawal 
    schedule will be removed from the Technical Specifications, the 
    Technical Specifications bases will continue to provide background 
    information on the use of the data obtained from material specimens. 
    Also, updates to the schedule will continue to be submitted to the 
    NRC for approval prior to implementation.
        Operating the plant in accordance with the new, updated P-T 
    Curves will assure preserving the structural integrity of the 
    reactor vessel over the life of the plant. The pressure and 
    temperature limits were developed in accordance with 10 CFR [Part] 
    50 Appendix G requirements.
        Removing the requirements associated with the previous exemption 
    to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative 
    change.
        Therefore, the proposed changes will not significantly increase 
    the probability or consequences of any accident previously 
    evaluated.
        Removal of the Reactor Vessel material specimen schedule from 
    the Technical Specifications has no impact on accidents at the 
    plant. Updates to the schedule will still be required to be 
    submitted to the NRC prior to implementation per Section II.B.3 of 
    Appendix H to 10 CFR Part 50.
        Also, updates to the P-T Curves will not create a new or 
    different type [of] accident. The reactor vessel beltline P-T limits 
    were revised applying the general guidance of the ASME Code, 
    Appendix G procedures with the necessary margins of safety for 
    heatup, cooldown and inservice hydro test conditions.
        The change to TS 4.4.8.1.2 items a & b is purely administrative.
        Therefore, the proposed changes will not create the possibility 
    of a new or different kind of accident from any accident previously 
    evaluated.
        Removal of the schedule for Reactor Vessel material specimen 
    withdrawal from the Technical Specifications does not impact the 
    margin of safety. The schedule will continue to receive NRC review 
    and approval prior to implementation of updates to the schedule.
        Updates to the P-T Curves are provided to preserve the margin to 
    [sic] safety to assure that when stressed under operating, 
    maintenance and testing the boundary behaves in a non-brittle manner 
    and the probability of rapidly propagating fracture is minimized.
        The change to TS 4.4.8.1.2 items a & b is purely administrative.
        Therefore, the proposed changes will not result in 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: University of New Orleans 
    Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
        Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
    Street N.W., Washington, D.C. 20005-3502
        NRC Project Director: William D. Beckner
    
    Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
    389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    
        Date of amendment request: February 22, 1995
        Description of amendment request: The proposed changes are 
    administrative in nature in that reference to an ``automatic'' 
    containment air lock tester will be deleted from TS 4.6.1.3. The 
    automatic airlock tester is no longer being used.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        (1) Operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        The proposed amendment is administrative in nature in that the 
    revision will eliminate the wording associated with optional use of 
    the personnel airlock automatic leakage tester. The requirement for 
    testing the personnel airlock at a pressure greater than or equal to 
    Pa for at least 15 minutes remains unchanged. The acceptance 
    criteria of personnel airlock seal leakage less than 0.01 La is 
    also unchanged. The automatic leakage tester is not an accident 
    initiator nor a part of the success path(s) which function to 
    mitigate accidents evaluated in the plant safety analyses. The 
    proposal does not involve any changes to the configuration or method 
    of operation of any plant equipment that is used to mitigate the 
    consequences of an accident, nor does it alter any assumptions or 
    conditions in the plant safety analyses. Therefore, operation of the 
    facility in accordance with the proposed amendment would not involve 
    a significant increase in the probability or consequences of an 
    accident previously evaluated.
        (2)Operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        The proposed amendment to remove the reference to the personnel 
    airlock automatic tester from the technical specifications will not 
    introduce any new failure modes or system interactions, nor will it 
    require the installation of any new or modified equipment. The 
    requirement to leak test the personnel air locks will not be 
    changed. Thus, operation of the facility in accordance with the 
    proposed amendment would not create the possibility of a new or 
    different kind of accident from any accident previously evaluated.
        (3)Operation of the facility in accordance with the proposed 
    amendment would not involve a significant reduction in a margin of 
    safety.
        The proposed amendment is administrative in nature in that it 
    eliminates the reference to the personnel airlock automatic leakage 
    tester but does not alter the surveillance and acceptance criteria 
    for such testing. Seal leakage testing is performed in accordance 
    with an approved plant procedure which allows use of either an 
    automatic tester or a portable testing cart. The automatic leakage 
    tester is not used to actuate safety related equipment, provide 
    interlocks, or perform plant control functions. The conditions 
    evaluated in the plant accident and transient analyses do not 
    involve this tester. The proposed change does not alter the basis 
    for any technical specification that is related to the establishment 
    of, or the maintenance of, a nuclear safety margin. Therefore, 
    operation of the facility in accordance with the proposed amendment 
    does not involve a significant reduction in a margin of safety.
        Based on the above discussion and the supporting Evaluation of 
    Technical Specification changes, FPL has determined that the 
    proposed license amendment involves no significant hazards 
    consideration.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this [[Page 16187]] 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: Indian River Junior College 
    Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
        Attorney for licensee: Harold F. Reis, Esquire, Newman and 
    Holtzinger, 1615 L Street, NW., Washington, DC 20036
        NRC Project Director: David B. Matthews
    
    Florida Power and Light Company, et al., Docket No. 50-389, St. 
    Lucie Plant, Unit No. 2, St. Lucie County, Florida
    
        Date of amendment request: February 27, 1995
        Description of amendment request: The proposed amendment will 
    change Table 3.3-3 and 3.3-4 to accommodate an improved coincidence 
    logic and relay replacement for the 4.16 kV Loss of Voltage Relays. 
    Actions required for certain trip units with the number of operable 
    channels one less than the total number of channels will also be 
    changed. In addition, the format used to state the time delay for the 
    4.16 kV Degraded Voltage trip unit will be revised.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        (1)Operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        The proposed change will result in a better overall posture of 
    the plant under degraded/loss of voltage conditions. The design 
    upgrade for the 4.16 kV Loss of Voltage system is more reliable, has 
    inherently higher accuracy, and is easier to maintain and calibrate 
    in the field. The coincidence logic will eliminate the spurious 
    plant trip potential from the existing design. Restating the maximum 
    time delay for the 4.16 kV Degraded Voltage (coincident with SIAS 
    [safety injection actuation signal]) protective relays in a ``less 
    than'' format will assure that the transfer of power to the on-site 
    sources occurs before the level of voltage becomes injurious to the 
    equipment under accident conditions, and will ensure that stripping 
    of the emergency power busses and loading of the EDG (s) [emergency 
    diesel generators] will occur within the time allowed by original 
    design criteria. The maximum allowed time delay for this function is 
    not being increased, and the time delay assumed in the accident 
    analyses for connecting the emergency bus to the diesel generator 
    will not be exceeded. Therefore, operation of the facility in 
    accordance with the proposed amendment will not involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        (2)Operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        The proposed amendment does not change the operation, function 
    or modes of plant operation. The ability of the loss of power and 
    degraded grid voltage protection scheme to properly transfer from 
    the off-site to the on-site power sources is being maintained. The 
    relays in the improved design of the 4.16 kV Loss of Voltage 
    function are of the type presently being used in identical 
    applications at both St. Lucie plant units. No new hazards are 
    created or postulated which may cause an accident different from any 
    accident previously analyzed. The modifications will result in a 
    more sensitive protection scheme allowing continuous operation 
    without unnecessary challenges to the safety systems, and will 
    continue to provide adequate protection to all the safety equipment. 
    Therefore, operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        (3)Operation of the facility in accordance with the proposed 
    amendment would not involve a significant reduction in a margin of 
    safety.
        The capability of the loss of power and degraded grid voltage 
    protection scheme is enhanced by the changes being proposed and is 
    confirmed by the existing surveillance requirements. The planned 
    modifications to the 4.16 kV Loss of Voltage function will result in 
    a more sensitive undervoltage detection system and reduce the 
    possibility of spurious actuation. The maximum time assumed in the 
    safety analyses for connecting each Emergency Bus to its dedicated 
    Emergency Diesel Generator is not being changed, and assurance that 
    separation from a degraded off-site power source will occur before 
    this time interval is exceeded during accident conditions will be 
    maintained by the proposed amendment. Accordingly, the margin of 
    safety is not affected. Therefore, operation of the facility in 
    accordance with the proposed amendment would not involve a 
    significant reduction in a margin of safety.
        Based on the discussion presented above and on the supporting 
    Evaluation of Proposed TS [Technical Specifications] Changes, FPL 
    has concluded that this proposed license amendment involves no 
    significant hazards consideration.
        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: Indian River Junior College 
    Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
        Attorney for licensee: Harold F. Reis, Esquire, Newman and 
    Holtzinger, 1615 L Street, NW., Washington, DC 20036
        NRC Project Director: David B. Matthews
    
    Florida Power and Light Company, et al., Docket No. 50-389, St. 
    Lucie Plant, Unit No. 2, St. Lucie County, Florida
    
        Date of amendment request: February 27, 1995
        Description of amendment request: The proposed amendment will 
    modify surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a 
    reduction in the required minimum shutdown cooling flow rate under 
    certain conditions during operational MODE 6. In addition, the format 
    of the SR will be changed to clarify the intent of the stated 
    surveillances.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        (1) Operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        Operation of the SDCS [shutdown cooling system] is not an 
    accident initiator and, therefore, does not significantly increase 
    the probability of an accident previously evaluated.
        The proposed change will allow a plant configuration needed to 
    perform maintenance activities on LPSI [low-pressure safety 
    injection]/SDCS headers by isolating one injection flow line for an 
    operable SDCS train during certain MODE 6 conditions. In the event 
    of a failure or unavailability of the alternate SDCS train, this 
    configuration could result in the proposed minimum flow rate.
        The proposed change only modifies the minimum required flow 
    rate, and does not affect the probability of this event. FPL has 
    evaluated the proposed value of reactor coolant flow and has shown 
    that the bases for the existing LCO [limiting condition for 
    operation] will continue to be satisfied. Therefore, there are no 
    significant increases in the consequences of any event from the 
    proposed change. No other system interactions are involved related 
    to previously evaluated accidents, and the proposed change has no 
    adverse effect on any other system performance.
        Therefore, operation of the facility in accordance with the 
    proposed amendment will not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        (2) Operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated.
        The proposed change does not affect the normal operation of the 
    plant. No new [[Page 16188]] systems are introduced and there is no 
    adverse effect on any other system configuration or performance. The 
    change will, however, allow isolation of one SDCS injection flow 
    path for maintenance activities in MODE 6 under controlled 
    conditions. The failure of the alternate SDCS train does not create 
    a new accident and has been further evaluated in the reduced flow 
    configuration, and shown to meet all the TS bases requirements. 
    Therefore, operation of the facility in accordance with the proposed 
    amendment will not create the possibility of a new or different kind 
    of accident from any accident previously evaluated.
        (3)Operation of the facility in accordance with the proposed 
    amendment would not involve a significant reduction in a margin of 
    safety.
        The safety considerations related to the proposed change are 
    described in the bases to TS [Technical Specification] 3/4.9.8. FPL 
    has evaluated the proposed reduction in SDCS flow requirement, under 
    stated conditions, and has shown that the proposed flow rate meets 
    all the TS bases requirements involving decay heat removal, boron 
    dilution, and stratification. Established acceptance criteria 
    providing margins of safety are not being changed by the proposed 
    amendment. Therefore, operation of the facility in accordance with 
    the proposed amendment would not involve a significant reduction in 
    a margin of safety.
        Based on the discussion presented above and on the supporting 
    Evaluation of Proposed TS Changes, FPL has concluded that this 
    proposed license amendment involves no significant hazards 
    consideration.
        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:Indian River Junior College 
    Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
        Attorney for licensee: Harold F. Reis, Esquire, Newman and 
    Holtzinger, 1615 L Street, NW., Washington, DC 20036
        NRC Project Director: David B. Matthews
    
    Georgia Power Company, Oglethorpe Power Corporation, Municipal 
    Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 
    50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, 
    Georgia
    
        Date of amendment request: March 14, 1995
        Description of amendment request: Georgia Power Company (GPC or the 
    licensee) has proposed a temporary change to Hatch Unit 2 Technical 
    Specification (TS) Required Action 3.3.6.1.F.1, and associated Bases. 
    The proposed change would add a note to the Primary Containment 
    Isolation Instrumentation actions to permit the drywell and wetwell 
    purge valves which are isolated by the drywell radiation monitor signal 
    to be opened with one inoperable drywell radiation monitor. The note 
    will expire prior to startup from the Hatch Unit 2 refueling/
    maintenance outage scheduled in the fall of 1995, at which time the 
    radiation monitor can be repaired or replaced. Should the unit be 
    forced into a cold shutdown of sufficient duration (i.e., drywell de-
    inerted), the inoperable radiation monitor will be repaired at that 
    time. The TS containment sections allow these valves to be opened for 
    inerting, de-inerting, and pressure control. However, with radiation 
    monitor 2D11-K621B inoperable, the primary containment isolation 
    instrumentation TS require the valves be closed until the unit achieves 
    a cold shutdown condition. Without the ability to open these valves 
    until cold shutdown, pressure control and de-inerting are difficult.
        The purpose of the high drywell radiation primary containment 
    isolation signal is to limit fission product release following a 
    postulated loss-of-coolant accident (LOCA) with significant fuel 
    damage. It is one of several signals which isolate the primary 
    containment vent and purge valves. A high drywell pressure signal will 
    not only shut down the reactor and generate a LOCA signal, it will also 
    isolate these valves.
        High drywell radiation indicates possible gross failure of the fuel 
    cladding. The generation of this isolation signal is not credited in 
    any accident or transient analysis. Chapter 15 of the Hatch Unit 2 
    Final Safety Analysis Report (FSAR) discusses the radiological 
    consequences of a postulated large break LOCA with fuel failure to show 
    conformance to 10 CFR Part 100 and 10 CFR Part 50, Appendix A. This 
    analysis is not affected.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration which is presented below:
        The change does not involve a significant hazards consideration 
    for the following reasons:
        1. The proposed amendment does not involve a significant 
    increase in the probability or consequences of an accident 
    previously evaluated.
        Opening the containment purge and vent valves with an inoperable 
    drywell radiation monitor will not increase the probability of any 
    previously evaluated accident. The fact that the monitor cannot send 
    an automatic isolation signal will not significantly affect the 
    consequences of an accident. The function of the primary containment 
    isolation signal is to detect and limit release of fission products 
    following significant fuel damage. The generation of this isolation 
    signal is not credited in any accident or transient analysis. 
    Chapter 15 of the Unit 2 FSAR evaluates the radiological 
    consequences of a postulated design basis LOCA with non-mechanistic 
    fuel damage. This licensing evaluation shows conformance to the 
    radiological limits presented in 10 CFR 100 and 10 CFR 50, Appendix 
    A. The results of this analysis are not affected since the valves 
    are otherwise operable and receive isolation signals from other 
    instrumentation.
        2. The proposed amendment does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated.
        The proposed change does not involve the installation of any new 
    equipment, or the modification of any equipment designed to prevent 
    or mitigate the consequences of accidents or transients. Therefore, 
    the change has no effect on any accident initiator, and no new or 
    different type of accidents are postulated to occur.
        3. The proposed amendment does not result in a significant 
    reduction in the margin of safety.
        As discussed in Item 1 above, the assumptions and results of the 
    licensing evaluations remain unchanged. Therefore, the margin of 
    safety is not significantly affected.
        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:Appling County Public Library, 
    301 City Hall Drive, Baxley, Georgia 31513
        Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
    Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
        NRC Project Director: Herbert N. Berkow
    
    GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
    Nuclear Generating Station, Ocean County, New Jersey
    
        Date of amendment request: February 28, 1995
        Description of amendment request: Technical Specification (TS) 
    Section 6.5.1.12 would be revised to delete the requirement to render 
    determinations in writing with regard to whether or not activities 
    listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed 
    safety question.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the [[Page 16189]] issue of no significant 
    hazards consideration, which is presented below:
        . Operation of the facility in accordance with the proposed 
    amendment would not involve a significant increase in the 
    probability of occurrence or the consequences of an accident 
    previously evaluated. The proposed change removes the requirement to 
    render determinations in writing with regard to whether or not 
    opposed changes to the Technical Specifications and investigations 
    of violations of Technical Specifications constitute an unreviewed 
    safety question. This change is considered an administrative change 
    to remove a requirement which is not relevant to these activities 
    and which is also consistent with the BWR Revised Standard Technical 
    Specifications (NUREG 1433). Existing requirements to perform 
    Technical and Independent Safety Reviews of these activities are not 
    affected. Therefore, the proposed amendment does not significantly 
    increase the probability of occurrence or the consequences of an 
    accident previously evaluated.
        2. Operation of the facility in accordance with the proposed 
    amendment would not create the possibility of a new or different 
    kind of accident from any accident previously evaluated. The 
    proposed change is considered administrative since it removes a 
    requirement which is not relevant to the affected activities, and 
    which is also consistent with the BWR Revised Standard Technical 
    Specifications Administrative Controls for Review and Audit. 
    Existing requirements to perform Technical and Independent Safety 
    Reviews for the affected activities are not changed. Therefore, this 
    change has no effect on the possibility of creating a new or 
    different king of accident from any accident previously evaluated.
        3. Operation of the facility in accordance with the proposed 
    amendment would not involve a significant reduction in a margin of 
    safety. The proposed change removes a requirement which is not 
    relevant to the affected activities. Existing Technical 
    Specification requirements to perform Technical and Independent 
    Safety Reviews for the affected activities are not changed and 
    therefore, will continue to ensure that such activities properly 
    address nuclear safety and safe plant operation. Therefore, it is 
    concluded that operation of the facility in accordance with the 
    proposed amendment does not involve a reduction in a margin of 
    safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location:Ocean County Library, Reference 
    Department, 101 Washington Street, Toms River, NJ 08753
        Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
    Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
        NRC Project Director: Phillip F. McKee
    
    Houston Lighting & Power Company, City Public Service Board of San 
    Antonio, Central Power and Light Company, City of Austin, Texas, 
    Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
    Matagorda County, Texas
    
        Date of amendment request: February 15, 1995
        Description of amendment request: The proposed amendment would 
    modify (by relocation to the Technical Requirements Manual) Technical 
    Specification (TS) 3/4.3.3.7, Chemical Detection Systems, and TS 3/
    4.8.4.1, Electrical Equipment Protective Devices - Containment 
    Penetration Conductor Overcurrent Protective Devices, and the 
    associated Bases.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        1. Does the proposed change involve a significant increase in 
    the probability or consequences of an accident previously evaluated?
        The proposed change to Technical Specification 3.3.3.7, Chemical 
    Detection Systems and 3.8.4.1, Electrical Equipment Protective 
    Devices-Containment Penetration Conductor Overcurrent Protective 
    Devices, is of an administrative nature in that the listed Technical 
    Specifications and Bases will be relocated in entirety to the 
    Technical Requirements Manual (TRM). Any future changes to the 
    relocated requirements will be in accordance with 10CFR 50.59 and 
    approved station procedures. Whether the listed Technical 
    Specifications and Bases are located in Technical Specifications or 
    the Technical Requirements Manual has no effect on the probability 
    or consequences of any accident previously evaluated.
        The proposed change does not alter the assumptions previously 
    made in the listed Technical Specifications. The proposed change 
    allows the Commission and South Texas more effective use of 
    personnel resources to control requirements that meet the four 
    Criteria in the Final Policy Statement. The proposed change will not 
    change the dose to workers.
        Since the probability of a [sic] accident is unaffected by the 
    administrative relocation of the listed Technical Specifications, 
    and the doses are not affected and do not exceed acceptance limits, 
    the proposed change does not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        2. Does the proposed change create the possibility of a new or 
    different kind of accident from any accident previously evaluated?
        The proposed change to Technical Specification 3.3.3.7, Chemical 
    Detection Systems and 3.8.4.1, Electrical Equipment Protective 
    Devices-Containment Penetration Conductor Overcurrent Protective 
    Devices, is of an administrative nature in that the listed Technical 
    Specifications and Bases will be relocated in entirety to the 
    Technical Requirements Manual (TRM). Any future changes to the 
    relocated requirements will be in accordance with 10CFR 50.59 and 
    approved station procedures. Whether the listed Technical 
    Specifications and Bases are located in Technical Specifications or 
    the Technical Requirements Manual has no effect on any previously 
    evaluated accident. It does not represent a change in the 
    configuration or operation of the plant and, therefore, does not 
    create the possibility of a new or different type of accident from 
    any accident previously evaluated.
        3. Does the proposed change involve a significant reduction in 
    the margin of safety?
        The proposed change to Technical Specification 3.3.3.7, Chemical 
    Detection Systems and 3.8.4.1, Electrical Equipment Protective 
    Devices-Containment Penetration Conductor Overcurrent Protective 
    Devices, is of an administrative nature in that the listed Technical 
    Specifications and Bases will be relocated in entirety to the 
    Technical Requirements Manual (TRM). Any future changes to the 
    relocated requirements will be in accordance with 10CFR 50.59 and 
    approved station procedures. The margin of safety is not reduced 
    when the requirements are relocated to a Licensee-controlled 
    document because the requirements to change a License Basis Document 
    via the 10CFR 50.59 process ensures the same questions concerning 
    the margin to safety required for a License Amendment are asked. The 
    major difference is the time and expense required for the License 
    Amendments. Therefore, this proposed change does not significantly 
    reduce the margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    request for amendments involves no significant hazards consideration.
        Local Public Document Room location: Wharton County Junior College, 
    J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
        Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
    P.C., 1615 L Street, N.W., Washington, D.C. 20036
        NRC Project Director: William D. Beckner
    
    Houston Lighting & Power Company, City Public Service Board of San 
    Antonio, Central Power and Light Company, City of Austin, Texas, 
    Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
    Matagorda County, Texas
    
        Date of amendment request: February 15, 1995 [[Page 16190]] 
        Description of amendment request: The proposed amendment would 
    modify Technical Specification 4.6.2.3.a.2 (and associated Bases) to 
    reflect the reactor containment fan cooler flow rate assumed in the 
    accident analyses and to specify that this flow is provided by the 
    component cooling water system.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        1. Does the proposed change involve a significant increase in 
    the probability or consequences of an accident previously evaluated?
        The proposed change to Technical Specification 4.6.2.3.a.2 is to 
    reflect the cooling water temperature assumed in the accident 
    analyses. The revised Technical Specification surveillance 
    requirement will change the cooling water flow rate requirement to 
    each Reactor Containment Fan Cooler from greater than or equal to 
    550 gallons per minute to greater than or equal to 1800 gallons per 
    minute.
        The proposed change, which will result in an increased 
    acceptance criteria for the flow to the Reactor Containment Fan 
    Coolers, is not indicative of accident initiators. The change will 
    ensure that the surveillance requirement reflects the flow rate 
    value assumed in the South Texas Project accident analyses and that 
    the design and operability requirements of equipment important to 
    safety are ensured.
        The accident mitigation features of the plant are not affected 
    by the proposed change since the change reflects the original 
    assumptions made in the design of the accident mitigation features 
    of the South Texas Project. Therefore, the proposed change does not 
    involve a significant increase in the probability or consequences of 
    an accident previously evaluated.
    
        2. Does the proposed change create the possibility of a new or 
    different kind of accident from any accident previously evaluated?
        The proposed change does not create the possibility of a new or 
    different kind of accident previously evaluated in the Safety 
    Analysis Report because all the accidents were analyzed with a flow 
    rate of 1800 gallons per minute to the Reactor Containment Fan 
    Cooler.
        3. Does the proposed change involve a significant reduction in a 
    margin of safety?
        There will be no adverse affects on margins of safety since a 
    more stringent surveillance requirement will be applied to the 
    Reactor Containment Fan Cooler. The Technical Specification 
    operability and surveillance requirements are not reduced but rather 
    made more restrictive by this proposed change. The change ensures 
    that the margin of safety originally intended for the Reactor 
    Containment Fan Coolers is maintained.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    request for amendments involves no significant hazards consideration.
        Local Public Document Room location:Wharton County Junior College, 
    J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
        Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
    P.C., 1615 L Street, N.W., Washington, D.C. 20036
        NRC Project Director: William D. Beckner
    
    IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, 
    Linn County, Iowa
    
        Date of amendment request: February 13, 1995
        Description of amendment request: The proposed amendment would 
    delete the audit frequency requirements from the Duane Arnold Energy 
    Center Technical Specifications (TS) and add them to the Quality 
    Assurance Program Description located in the Updated Final Safety 
    Analysis Report.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        (1) The proposed amendment does not involve a change in the 
    probability or consequences of an accident previously evaluated. No 
    physical changes will occur as a result of this amendment. The 
    change is administrative in nature and does not impact the operation 
    of the plant or the plant's response to any accident. Because it 
    will allow management the flexibility to adjust the audit 
    frequencies based upon the performance of the program or 
    organization being audited, the overall performance of the 
    organization will be improved.
        (2) The proposed amendment does not create the possibility of a 
    new or different kind of accident from any accident previously 
    evaluated. No physical changes will occur as a result of this 
    amendment. The change is administrative in nature and does not 
    affect the operation or design of the plant; therefore, it does not 
    create the possibility of a new or different kind of accident from 
    any accident previously evaluated. The audits will continue to be 
    performed to provide assurance of conformance to the applicable 
    requirements.
        (3) The proposed amendment will not reduce the margin of safety. 
    No physical changes will occur as a result of this amendment. The 
    change is administrative in nature and does not affect the operation 
    or design of the plant. Safety limits and limiting safety system 
    settings are not affected by this proposed amendment. The amendment 
    removes requirements for frequency of audits from the TS, thus 
    permitting more effective scheduling of audits based on performance 
    and the status of the activities audited. This should result in a 
    more effective audit program that will contribute to an improvement 
    in the overall performance of the organization.
        The NRC staff 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: Cedar Rapids Public Library, 
    500 First Street, S.E., Cedar Rapids, Iowa 52401
        Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
    & Bockius, 1800 M Street, N. W., Washington, D. C. 20036-5869NRC Acting 
    Project Director: John N. Hannon
    
    Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
    No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
    Illinois
    
        Date of amendment request: February 10, 1995
        Description of amendment request: The proposed amendment would 
    modify Technical Specification 3.3.2.1, ``Control Rod Block 
    Instrumentation,'' to revise two surveillance requirements and their 
    associated notes for the Rod Withdrawal Limiter (RWL) mode of the Rod 
    Pattern Control System. These changes will conform these requirements 
    to their original bases and eliminate the potential for unnecessary 
    power reductions.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration which is presented below:
        (1) The proposed changes are consistent with the Rod Withdrawal 
    Error (RWE) analysis presented in Clinton Power Station (CPS) 
    Updated Safety Analysis Report (USAR) Section 15.4.2. The proposed 
    changes do not result in any change to plant equipment or operation; 
    only the plant conditions for which the Rod Withdrawal Limiter (RWL) 
    function(s) are required to be tested are being revised. The 
    proposed changes continue to ensure that the RWL is OPERABLE and 
    tested to ensure that continuous control rod withdrawals remain 
    within the assumptions of the RWE analyses. The proposed changes 
    have no impact on the probability of occurrence of a RWE event. 
    Therefore, the proposed changes do not result in a significant 
    increase in the probability or consequences of any accident 
    previously evaluated.
        (2) The proposed changes do not result in any changes to plant 
    equipment or operation; only the plant conditions for which the RWL 
    [[Page 16191]] function(s) are required to be OPERABLE and tested 
    are being revised. The proposed changes continue to ensure that the 
    RWL is OPERABLE and tested to ensure that continuous control rod 
    withdrawals remain within the assumptions of the RWE analyses. As a 
    result, no new failure modes are introduced. The proposed changes 
    are clearly within the limits of plant operation as described in the 
    USAR and the RWE analyses. Therefore, the proposed changes cannot 
    create the possibility of a new or different kind of accident from 
    any accident previously evaluated.
        (3) The proposed changes revise the testing requirements to be 
    consistent with the testing required prior to Amendment No. 95. The 
    proposed changes ensure that the RWL is OPERABLE and tested to 
    ensure that continuous control rod withdrawals remain within the 
    assumptions of the RWE analyses. The proposed changes are clearly 
    within the limits of plant operation as described in the USAR and 
    the RWE analyses. Therefore, the proposed changes do not involve a 
    reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location:Vespasian Warner Public 
    Library, 120 West Johnson Street, Clinton, Illinois 61727
        Attorney for licensee: Leah Manning Stetzner, Vice President, 
    General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
    Illinois 62525.
        NRC Acting Project Director: John N. Hannon
    
    Maine Yankee Atomic Power Company, Docket No. 50-309, Maine 
    YankeeAtomic Power Station, Lincoln County, Maine
    
        Date of amendment request: February 14, 1995
        Description of amendment request: The proposed amendment would 
    change responsibility for audits of the emergency and security plans 
    and their implementing procedures. Audit responsibility would change 
    from the licensee's Nuclear Safety Audit and Review (NSAR) Committee 
    and the Plant Operation Review Committee (PORC), to the respective 
    emergency and security plans. The proposed amendment is consistent with 
    the guidance of NRC Generic Letter 93-07, Modification of the Technical 
    Specification Administrative Control Requirements for Emergency and 
    Security Plans, dated December 28, 1993.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration. The NRC staff has reviewed the licensee's analysis 
    against the Standards of 10 CFR 50.92(c). A summary of the licensee's 
    analysis is presented below:
        1. The proposed amendment would not involve a significant 
    increase in the probability or consequences of an accident 
    previously analyzed.
        The proposed changes do not have a direct effect on the physical 
    plant or the maintenance of the physical plant, but would improve the 
    safe operation of the plant by reducing the administrative burden of 
    PORC and NSAR. This change would allow a better focus of management 
    resources to the operational safety oversight of plant activities. The 
    requirement to review, audit, document, control, and submit for 
    regulatory review, the Emergency Plan and the Security Plan and their 
    implementing procedures, is defined by regulation and remains 
    unchanged. The proposed changes will not, of themselves, result in any 
    reduction in the effectiveness of either the Emergency Plan or the 
    Security Plan to protect the health and safety of the public. The 
    proposed changes, therefore, will not increase the probability or 
    consequences of an accident previously evaluated.
        2. The proposed amendment would not create the possibility of a 
    new or different kind of accident from any previously evaluated.
        This change is administrative in nature and does not change or 
    modify the physical plant or maintenance of the physical plant. 
    Applicable regulations continue to enforce the requirement for review 
    and audit by individuals not responsible for implementation of the 
    existing programs. Consequently, independent oversight of the programs 
    and procedures is not compromised by these proposed changes. The 
    possibility of a new or different accident from any previously 
    evaluated as a result of future changes in the implementation of the 
    Security or Emergency Plans is not created.
        3. The proposed amendment would not involve a significant 
    reduction in a margin of safety.
        The proposed changes will revise the administrative 
    responsibilities of the PORC and NSAR committees allowing a better 
    focus of resources on operational safety reviews. The requirements 
    of the applicable Federal and State regulations ensure the continued 
    effective oversight of the implementation of Security and Emergency 
    Plans. Consequently, the adoption of the proposed changes would 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) 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, Maine 04578
        Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
    Power Company, 329 Bath Road, Brunswick, Maine 04011
        NRC Project Director: Walter R. Butler
    
    Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
    Millstone Nuclear Power Station, Unit No. 2, New London County, 
    Connecticut
    
        Date of amendment request: October 18, 1994, as supplemented 
    February 21, 1995.
        Description of amendment request: The following changes requested 
    in the October 18, 1994, submittal were published in Federal Register 
    on November 9, 1994 (59 FR 35876). The proposed amendment would require 
    three Type A overall Integrated Containment Leakage Tests be conducted 
    at approximately equal intervals during shutdowns during each 10 year 
    service period. For the third Type A test for the second 10-year 
    period, it would be conducted during the thirteenth refueling outage 
    extending the second 10-year service period to the end of the 
    thirteenth refueling outage. The amendment would also change the 
    Containment Leakage Bases by reflecting the conditions of a proposed 
    exemption to 10CFR50, Appendix J, that would remove the requirement 
    that the third Type A test for each 10-year period be conducted when 
    the plant is shutdown for the 10-year plant inservice inspection.
        By letter dated February 21, 1995, the licensee withdrew the action 
    related to conducting the third Type A test for the second 10-year 
    period during the thirteenth refueling outage and the reference to a 
    proposed exemption to 10 CFR 50, Appendix J, that would remove the 
    requrement that the third Type A test for each 10-year period be 
    conducted when the plant is shutdown for the 10-year plant inservice 
    inspection. The following basis for the proposed no significant hazards 
    consideration determination relates to the February 21, 1995, request.
        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 
    [[Page 16192]] consideration, which is presented below:
        The proposed change does not involve a SHC because the change 
    would not:
        1. Involve a significant increase in the probability or 
    consequences of an accident previously analyzed.
        Type A tests are performed to ensure that the total leakage from 
    containment does not exceed the maximum allowable primary 
    containment leakage rate at the design pressure. This assures 
    compliance with the dose limits of 10CFR100.
        The proposed change to Surveillance Requirement 4.6.1.2.a of the 
    Millstone Unit No. 2 Technical Specifications will increase the 
    flexibility for scheduling the Type A tests. It does not modify the 
    maximum allowable leakage rate at the design containment pressure, 
    does not impact the design basis of the containment, and does not 
    make any physical or operational changes to existing plant 
    structures, systems, or components.
        Historically, Type A tests have a relatively low failure rate 
    where Type B and C testing (local leakage rate tests) could not 
    detect the leakage path. Most Type A test failures are attributed to 
    failures of Type B or C components (containment penetrations and 
    isolation valves). Type B and C components are tested per 
    Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 2 
    Technical Specifications. These tests are required to be conducted 
    at intervals no greater than 24 months. These local leakage rate 
    tests provide assurance that containment integrity is maintained. 
    The Type B and C tests will continue to be performed in accordance 
    with the requirements of Surveillance Requirement 4.6.1.2.d.
        The previous Type A, B, and C tests demonstrate that Millstone 
    Unit No. 2 has maintained control of containment integrity by 
    maintaining a conservative margin between the acceptance criterion 
    and the ``As-Found'' and ``As-Left'' leakage results. Based on this, 
    the Millstone Unit No. 2 containment is considered to be in sound 
    condition.
        Based on the above, the proposed change to Surveillance 
    Requirement 4.6.1.2.a of the Millstone Unit No. 2 Technical 
    Specifications does 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 change to Surveillance Requirement 4.6.1.2.a of the 
    Millstone Unit No. 2 Technical Specifications will increase the 
    flexibility in scheduling the Type A tests. It does not make any 
    physical or operational changes to existing plant structures, 
    systems, or components. In addition, the proposed change does not 
    modify the acceptance criteria for the Type A tests. Maintaining the 
    leakage through the containment boundary to the atmosphere within a 
    specific value ensures that the plant complies with the requirements 
    of 10CFR100. The containment boundary serves as an accident 
    mitigator; it is not an accident initiator. Therefore, the proposed 
    change to Surveillance Requirement 4.6.1.2.a does 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 change to Surveillance Requirement 4.6.1.2.a of the 
    Millstone Unit No. 2 Technical Specifications will increase the 
    flexibility for scheduling the Type A tests. It does not modify the 
    maximum allowable leakage rate at the design containment pressure, 
    does not impact the design basis of the containment, and does not 
    make any physical or operational changes to existing plant 
    structures, systems, or components.
        Based on the above, the proposed change does not involve a 
    significant reduction in the margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location: Learning Resource Center, 
    Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
    London Turnpike, Norwich, CT 06360.
        Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
    Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
    06141-0270.
        NRC Project Director: Phillip F. McKee
    
    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 amendment request: February 1, 1995
        Description of amendment request: This Technical Specification (TS) 
    change would modify the applicable operational conditions for the 
    secondary containment isolation radiation monitors located on the 
    refueling floor and for the radiation monitor located in the railroad 
    access shaft. Specifically, for the refueling floor exhaust duct and 
    wall exhaust duct radiation monitors, the proposed change would modify 
    the applicable operational condition during specific control rod 
    testing evolutions which are core alterations and would indicate that 
    the operability requirement does not apply during shutdown margin 
    demonstrations. For the railroad access shaft exhaust duct radiation 
    monitor, the change to the TS would modify the applicable operational 
    condition to address plant evolutions involving irradiated fuel 
    transfer within the railroad accessshaft and above the access shaft 
    with the equipment hatch open.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        I. This proposal does not involve a significant increase in the 
    probability or consequences of an accident previously evaluated.
        a. The proposed change to the applicable operational condition 
    for the refueling floor process radiation monitors does not affect 
    the probability of the design basis accidents. The monitors function 
    in response to an airborne radioactivity concentration in the 
    unfiltered air from the Zone III exhaust system and provide 
    isolation signals which limit offsite doses to within regulatory 
    limits. As such, there is no correlation between monitor operability 
    and accident probability. The monitors act to mitigate the offsite 
    effects of airborne contamination producing accidents, they are not 
    potential accident initiators.
        The proposed change does not result in a significant increase in 
    the consequence of the design basis accidents. The postulated event 
    associated with control rod related CORE ALTERATIONS which could 
    result in increased Zone III airborne radioactivity concentrations 
    is criticality resulting from a single control rod withdrawal, 
    resulting in release of fission products. The probability of an 
    unintended criticality from a single control rod withdrawal is low, 
    and the potential for this criticality to result in fuel failure 
    under shutdown conditions is even more remote. Withdrawal of a 
    single control rod is an analyzed evolution during which time 
    adequate design and operating controls exist to preclude 
    criticality. However, in the unlikely event criticality should 
    occur, the potential offsite effects would not be significant. 
    Localized criticality involving a leaking rod, or criticality 
    induced fuel failure, are the postulated mechanisms by which an 
    increase in Zone III airborne radioactivity could be attained. 
    Neither of these postulated, but very unlikely events, will result 
    in radioactive release in excess of 10CFR100 limits. Any release 
    would be monitored by instrumentation in the Reactor Building vent 
    stack required to be OPERABLE at all times. In addition, Area 
    Radiation Monitors are installed on the refueling floor to 
    supplement the refueling floor process radiation monitors by 
    providing radiological information to plant operators. Operators can 
    use the vent stack and/or ARM information to manually initiate 
    secondary containment isolation if radiological conditions warrant 
    this action. Emergency Operating Procedures direct operator action 
    in the event of higher than normal radiation readings.
        b. The proposed change to the applicable operational condition 
    for the railroad access shaft process radiation monitor does not 
    affect the probability of the design basis accidents. The monitor 
    functions in response to an airborne radioactivity concentration in 
    the unfiltered air from the Zone III exhaust [[Page 16193]] system 
    and provides isolation signals which limit offsite doses to within 
    regulatory limits. As such, there is no correlation between monitor 
    operability and accident probability. The monitor acts to mitigate 
    the offsite effects of airborne contamination producing accidents, 
    it is not a potential accident initiator.
        The proposed change does not result in a significant increase in 
    the consequence of the design basis accidents. The design intent of 
    the railroad access shaft process radiation monitor is to monitor 
    radiation in the unfiltered air from the Zone III railroad access 
    shaft exhaust system, and provide signals which automatically 
    isolate the Zone III portion of the secondary containment, start the 
    Standby Gas Treatment System, and start the Recirculation System 
    (Zone III) on a high radiation condition within the access shaft. 
    This function is intended to limit the consequences of a fuel 
    handling accident in the railroad access shaft. The monitor has no 
    significant capability to react to a CORE ALTERATION related 
    transient, or one resulting from operations with the potential to 
    drain the reactor vessel. The design intent of the monitor is 
    maintained under the proposed change, as the proposed change focuses 
    monitor operability on conditions when irradiated fuel is in the 
    railroad access shaft or above it with the railroad access shaft 
    cover open.
        For the above stated reasons, the applicable operational 
    condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
    Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
    Exhaust Duct Radiation Monitor can be modified without significantly 
    increasing the probability or consequences of an accident previously 
    evaluated.
        II. This proposal does not create the possibility of a new or 
    different kind of accident from any accident previously evaluated.
        The Refueling Floor Exhaust Duct High Radiation Monitors, Wall 
    Exhaust Duct Radiation Monitors, and the Railroad Access 
    ShaftExhaust Duct Radiation Monitor function in response to an 
    airborne radioactivity concentration in the unfiltered air from the 
    Zone III exhaust system and provide isolation signals which limit 
    offsite doses to within regulatory limits. As such, there is no 
    correlation between monitor operability and the potential for 
    creating new or different accident scenarios. The monitors act to 
    mitigate the offsite effects of airborne contamination producing 
    accidents, they are not potential accident initiators.
        For the above stated reasons, the applicable operational 
    condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
    Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
    Exhaust Duct Radiation Monitor can be modified without creating the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        III. This change does not involve a significant reduction in a 
    margin of safety.
        a. The proposed change to the applicable operational condition 
    for the refueling floor process radiation monitors does not involve 
    a significant reduction in the margin of safety. The postulated 
    event associated with control rod related CORE ALTERATIONS which 
    could result in increased Zone III airborne radioactivity 
    concentrations is criticality resulting from a single control rod 
    withdrawal under shutdown conditions. There are multiple barriers to 
    protect against the postulated event of criticality from a single 
    rod withdrawal. Technical Specifications, plant operating 
    procedures, and plant design control the withdrawal of control rods 
    to minimize the potential for an inadvertent criticality event 
    during shutdown. In addition, a fuel loading verification is 
    performed, per procedure, on the as loaded core configuration to 
    ensure that the fuel is loaded correctly. Each reload core is 
    designed such that there is at least a 99.9% probability with a 95% 
    confidence that the core will not be critical as a result of a 
    single control rod withdrawal. The safety margin associated with a 
    potential criticality event from a single control rod withdrawal, 
    under shutdown conditions, is not impacted by the proposed change.
        In the unlikely event that control rod manipulations resulted in 
    reactor criticality, adequate protective measures are provided by 
    core monitoring instrumentation required to be operable in OPCON 5. 
    Under this scenario, assuming the inadvertent control rod withdrawal 
    resulted in a significant reactivity addition, the Reactor 
    Protection System (RPS) would respond by inserting all control rods 
    via the Scram function. The RPS monitors for recriticality during 
    OPCON 5 with SRMs (per Technical Specification Section 3.9.2), and 
    IRMs. The safety margin associated with RPS response to a 
    criticality event, under shutdown conditions, is not impacted by the 
    proposed change.
        Assuming that a criticality did occur as a result of a single 
    control rod withdrawal, any increase in Zone III airborne 
    radioactivity from a previously failed assembly located in the 
    vicinity of the withdrawn control rod or a fuel rod failure 
    associated with the control rod withdrawal would not result in an 
    offsite dose exceeding regulatory limits. Assuming that criticality 
    occurs following core loading and verification (i.e. 20 
    days after shutdown), the offsite dose as a result of the release of 
    fission products from a single failed fuel rod would be much less 
    than 1% of the applicable site boundary limits. In addition, the 
    failure of four complete fuel assemblies (i.e. nearly equal to 300 
    fuel rods in the bundles surrounding the withdrawn control rod) 
    would not result in offsite dose exceeding the applicable regulatory 
    limits. Failure of more than four complete fuel assemblies due to 
    the withdrawal of a single control rod in OPCON 5 is not considered 
    credible. In fact, given the initial conditions of this event (i.e. 
    cold, zero power, subcritical) and the reactivity characteristics of 
    the fuel (i.e. negative fuel temperature reactivity coefficient) it 
    is very unlikely that a criticality of this nature would result in 
    failure of any fuel rods. Although the refueling floor process 
    radiation monitors would not be OPERABLE, Zone III airborne 
    radioactivity concentrations can be independently detected with Area 
    Radiation Monitors (ARMs) which are located on the refueling floor. 
    These monitors provide control room indication, and would alert 
    operators to changing radiological conditions on the refueling 
    floor. In addition to providing personnel notification, the ARMs act 
    as a supplement to the process radiation monitors in detecting 
    abnormal migrations of radioactive material in or from the process 
    streams. Operators can manually initiate secondary containment 
    isolation based on ARM input. The Emergency Operating Procedures 
    require the operators to take appropriate actions on higher than 
    normal radiation readings. Moreover, any airborne radioactivity 
    leakage from Zone III would be monitored via instrumentation in the 
    Reactor Building vent stack required to be OPERABLE at all times; 
    local alarms, remote recording, and main control room and Technical 
    Support Center alarms are provided. Operators can manually initiate 
    secondary containment isolation based on exhaust sample readings. 
    Due to the bounding regulatory limits and the redundant monitoring 
    and operator response capabilities, the safety margin associated 
    with the potential for offsite airborne radioactive release, under 
    shutdown conditions, is not significantly impacted by the proposed 
    change.
        b. The elimination of operability requirements associated with 
    CORE ALTERATIONS, operations with the potential to drain the reactor 
    vessel, and other irradiated fuel moves not associated with the 
    railroad access shaft, do not affect the ability of the railroad 
    access shaft process radiation monitor to implement its design 
    function. As such, the current operability requirements for the 
    monitor which involve evolutions in areas other than the railroad 
    access shaft do not contribute to the margin of plant safety; thus 
    eliminating these operability requirements will not reduce the 
    margin of plant safety.
        For the above stated reasons, the applicable operational 
    condition for Refueling Floor Exhaust Duct High Radiation Monitors, 
    Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft 
    Exhaust Duct Radiation Monitor can be modified without 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: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
        Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
    and Trowbridge, 2300 N Street NW., Washington, DC 20037
        NRC Project Director: John F. Stolz [[Page 16194]] 
    
    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 amendment request: February 2, 1995
        Description of amendment request: This amendment would change the 
    Technical Specifications for the units to increase the licensed 
    discharge fuel assembly exposure for SPC 9X9-2 fuel from 40 to 45 GWD/
    MTU.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        The proposed changes do not:
        I. Involve a significant increase in the probability or 
    consequences of an accident previously evaluated.
        PP&Ls technical basis for increasing the licensed discharge 
    exposure limit as proposed is documented in PL-NF-94-005-P-A. The 
    technical basis includes onsite fuel inspections, fuel design 
    analyses and evaluations, and an in-reactor fuel assembly extended 
    exposure demonstration. In response to NRC concerns on fuel failures 
    at higher exposures, very conservative analyses were performed for 
    the CRDA [control rod drive assembly] assuming very low failure 
    thresholds, and offsite dose calculation results were shown to be 
    well within regulatory limits, even at a failure threshold of 30 
    cal/gm. The NRC has previously reviewed and approved all of the 
    above information, and inspection results have met all approved 
    criteria.
        An evaluation of FSAR [Final Safety Analysis Report] design 
    basis events was performed to determine the impact of the proposed 
    increase in fuel exposure. The LOCA [loss-of-coolant accident] 
    analysis performed in support of PP&Ls Power Uprate efforts 
    incorporated the effects of higher exposure and LHGR [linear heat 
    generation rate]. From a radiological release perspective, the Power 
    Uprate evaluations of LOCA, MSLB [main steam line break], CRDA, and 
    refueling accidents each bound the potential impacts of extended 
    exposure fuel.
        Those reload analyses deemed necessary to confirm that the above 
    conclusions remain valid will be performed on a cycle-specific 
    basis.
        Based on the above, the proposed action will not involve a 
    significant increase in the probability or consequences of an 
    accident previously evaluated.
        II. Create the possibility of a new or different kind of 
    accident from any accident previously evaluated.
        The proposed action will increase the residence time of fuel 
    within the Susquehanna reactors. The potential consequences of this 
    action remain solely with the fuels ability to perform within 
    specified limits during the increased duty, and were reviewed in I 
    above. All required evaluations involving fuel impacts have been 
    previously evaluated.
        Based on the above, the proposed action cannot create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        III. Involve a significant reduction in a margin of safety.
        The proposed action will allow increasing the licensed discharge 
    fuel assembly exposure limit, resulting in increases in the fuel rod 
    LHGR and LHGR for APRM [average power range monitor] Setpoints, 
    which are controlled via the Technical Specifications and the Core 
    Operating Limits Report.
        The discussion in I. above delineates the evaluations performed 
    to support this action. It concludes that neither the probability 
    nor the consequences of events previously evaluated will be 
    affected. Operator performance will not be affected, because the 
    operators only monitor the ratio of the fuel LHGR to the fuel design 
    limit. No other potentially impacted safety margins have been 
    identified.
        Based on the above, the proposed change does not involve a 
    significant reduction in a margin of safety.
        The NRC staff has reviewed the licensee's analysis and, based on 
    this review, it appears that the three standards of 10 CFR 50.92(c) are 
    satisfied. Therefore, the NRC staff proposes to determine that the 
    amendment request involves no significant hazards consideration.
        Local Public Document Room location:Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
        Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
    and Trowbridge, 2300 N Street NW., Washington, DC 20037
        NRC Project Director: John F. Stolz
    
    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 amendment request: February 10, 1995
        Description of amendment request: The proposed amendment would 
    modify the Susquehanna Steam Electric Station, Unit 1 and 2 Technical 
    Specifications (TS) to (1) extend the allowable out-of-service times 
    (AOTs) for maintenance and repair and the surveillance test intervals 
    (STIs) between channel functional tests for the following groups of 
    instruments: reactor protection systems instrumentation (TS 3.3.1), 
    isolation actuation instrumentation (TS 3.3.2), emergency core cooling 
    system actuation instrumentation (TS 3.3.3), ATWS (anticipated 
    transient without scram) recirculation pump trip system instrumentation 
    (TS 3.3.4.1), end-of-cycle recirculation pump trip system 
    instrumentation (TS 3.3.4.2), reactor core isolation cooling system 
    (RCIC) actuation instrumentation (TS 3.3.5), control rod block 
    instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS 
    3.3.7.1), and feedwater/main turbine trip system actuation 
    instrumentation (TS 3.3.90); (2) change the required actions and AOTs 
    for the instruments listed above to make requirements consistent with 
    supporting analysis in General Electric topical reports and change 
    additional actions required to prevent extended AOTs from resulting in 
    extended loss of instrument function; (3) change the required actions 
    and AOTs for the instruments listed above for instrumentation 
    associated with the ADS (automatic depressurization system), 
    recirculation pump trip, and pump suction lineup for HPCI (high 
    pressure core injection) and RCIC; (4) change applicability 
    requirements and required actions for the reactor vessel water level-
    low, level 3 function that isolates the RHR (residual heat removal) 
    system shutdown cooling system so that the function is required to be 
    operable in operational conditions 3, 4, and 5 to prevent inadvertent 
    loss of reactor coolant via the RHR shutdown cooling system; (5) remove 
    notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on 
    leak detection temperature detectors and remove the note to TS 3.3.6 
    for Unit 1 related to a previous relief from TS 3.0.4; and reformat, 
    renumber, and/or reword existing requirements to incorporate the 
    changes listed above.
        Basis for proposed no significant hazards consideration 
    determination: As required by 10 CFR 50.91(a), the licensee has 
    provided its analysis of the issue of no significant hazards 
    consideration, which is presented below:
        I. The proposed changes do not involve a significant increase in 
    the probability or consequences of an accident previously evaluated.
        The proposed TS changes increase the AOTs and STIs for actuation 
    instrumentation intended to detect or mitigate accidents; establish 
    required actions consistent with NUREG-1433 for some instruments 
    that are more specific but equivalent to existing required actions; 
    establish new requirements to prevent inadvertent loss of reactor 
    coolant via the RHR Shutdown Cooling System during OPERATIONAL 
    CONDITIONS 3, 4 and 5; and, eliminate notes that were intended to 
    provide one time only exemptions from certain requirements. The 
    proposed changes affect only those Technical Specification 
    requirements that govern operability, required actions and routine 
    testing of plant instruments that detect or mitigate accidents. The 
    proposed changes do [[Page 16195]] not affect any equipment or 
    requirements that are assumed to be initiators of any analyzed 
    events. Therefore, the proposed changes will not involve an increase 
    in the probability of occurrence of an accident previously 
    evaluated.
        The proposed changes will not increase the consequences of an 
    accident previously evaluated because the changes will not involve 
    any physical changes to plant systems, structures, or components 
    (SSC), or the manner in which these SSC are operated, maintained, 
    modified, tested or inspected. The proposed changes will not alter 
    the operation of equipment assumed to be available for the 
    mitigation of accidents or transients by the plant safety analysis 
    or licensing basis. The proposed changes extend the intervals 
    between required performances of routine instrument testing. The 
    proposed changes also modify time limits allowed for operation with 
    inoperable instrument channels in situations when an inoperable 
    instrument channel would not prevent actuation of the associated 
    equipment. These changes are based on the demonstrated reliability 
    of these instruments and are justified by the analysis in References 
    1 through 8 [See February 10, 1995 application]. The small increases 
    in the probability that the proposed changes will result in an 
    equipment actuation failure has been determined in References 1 
    through 8 [See February 10, 1995 application] to be offset by safety 
    benefits such as a reduction in the number of inadvertent 
    actuations, a reduction in wear due to excessive testing, and better 
    utilization of plant personnel and resources. These changes will not 
    allow continuous plant operation with plant conditions such that a 
    single failure will result in a loss of any safety function.
        Proposed changes to required actions and completion times for 
    instrumentation associated with the ADS initiation, Recirculation 
    Pump Trip, and pump suction lineup for HPCI and RCIC make the 
    required actions and completion times consistent with NUREG-1433, 
    Standard Technical Specifications for General Electric Plants, BWR/
    4, Revision 0 (Reference 12). These changes are also consistent with 
    the assumptions used in References 1 through 8 [See February 10, 
    1995 application]. Therefore, these changes establish or maintain 
    adequate assurance that components are operable when necessary for 
    the prevention or mitigation of accidents or transients and that 
    plant variables are maintained within limits necessary to satisfy 
    the assumptions for initial conditions in the safety analysis. In 
    addition, the proposed change provides the benefit of avoiding an 
    unnecessary shutdown transient when appropriate measures are 
    available to compensate for the inoperable instrumentation. 
    Therefore, the proposed changes will not increase the consequences 
    of an accident previously evaluated.
        There is no significant increase in the probability or 
    consequences of an accident previously evaluated resulting from 
    changes that reformat, renumber, and/or reword existing requirements 
    to incorporate the changes above or from the removal of notes that 
    were intended for one time only use and are no longer applicable.
        II. The proposed changes do not create the possibility of a new 
    or different kind of accident from any accident previously 
    evaluated.
        This proposed change will not involve any physical changes to 
    plant systems, structures, or components (SSC), or the manner in 
    which these SSC are operated, maintained, modified, tested, or 
    inspected. The changes in normal plant operation are consistent with 
    the current safety analysis assumptions. Therefore, this change will 
    not create the possibility of a new or different kind of accident 
    from any accident previously evaluated.
        III. The proposed changes do not involve a significant reduction 
    in a margin of safety.
        The proposed TS changes: increase the AOTs and STIs for 
    actuation instrumentation intended to detect or mitigate accidents; 
    establish required actions consistent with NUREG-1433 for some 
    instruments that are more specific but equivalent to existing 
    required actions; establish new requirements to prevent inadvertent 
    loss of reactor coolant via the RHR Shutdown Cooling System during 
    Operational Conditions 3, 4 and 5; and, eliminate notes that were 
    intended to provide one time only exemptions from certain 
    requirements.
        There is no significant reduction in the margin of safety 
    resulting from changes to the minimum surveillance test intervals 
    (STIs) and allowable out-of-service times (AOTs) for the testing 
    and/or repair of instrumentation. This conclusion is based on the 
    demonstrated reliability of these instruments and is justified by 
    the analysis in References 1 through 8 [See February 10 1995 
    application]. The small increases in the probability that the 
    proposed changes will result in an equipment actuation failure has 
    been determined in References 1 through 8 [See February 10, 1995 
    application] to be offset by safety benefits such as a reduction in 
    the number of inadvertent actuations, a reduction in wear due to 
    excessive testing.
        These changes will not allow continuous plant operation with 
    plant conditions such that a single failure will result in a loss of 
    any safety function.
        There is no significant reduction in the margin of safety 
    resulting from changes to required actions and completion times for 
    instrumentation associated with the ADS initiation, Recirculation 
    Pump Trip, and pump suction lineup for HPCI and RCIC. These changes 
    make the required actions and completion times consistent with 
    NUREG-1433, Standard Technical Specifications for General Electric 
    Plants, BWR/4. These changes are also consistent with the 
    assumptions used in References 1 through 8 [See February 10, 1995 
    application]. Therefore, these changes establish or maintain 
    adequate assurance that components are operable when necessary for 
    the prevention or mitigation of accidents or transients and that 
    plant variables are maintained within limits necessary to satisfy 
    the assumptions for initial conditions in the safety analysis. In 
    addition, the proposed change provides the benefit of avoiding an 
    unnecessary shutdown transient when appropriate measures are 
    available to compensate for the inoperable instrumentation. 
    Additionally, the proposed required actions ensure that actions to 
    mitigate loss of single failure tolerance are initiated within 24 
    hours (12 hours for RPS) in accordance with the results of the 
    analyses in References 1 through 8 [See February 10, 1995 
    application] and action to mitigate a loss of instrument function is 
    initiated within 1 hour. Therefore, these changes will not allow 
    continuous plant operation with plant conditions such that a single 
    failure will result in a loss of any safety function. The 
    Pennsylvania Power & Light Company performed reviews that confirmed 
    the analyses in References 1 through 8 [See February 10, 1995 
    application] are applicable to SSES and that there would be no 
    effect on the identification of excessive instrument setpoint drift 
    as a result of increasing the minimum interval between instrument 
    functional tests from monthly to quarterly.
        There is no significant reduction in the margin of safety 
    resulting from changes that reformat, renumber, and/or reword 
    existing requirements to incorporate the changes above or from the 
    removal of notes that were intended for one time only use and are no 
    longer applicable.
        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: Osterhout Free Library, 
    Reference Department, 71 South Franklin Street, Wilkes-Barre, 
    Pennsylvania 18701
        Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
    and Trowbridge, 2300 N Street NW., Washington, DC 20037
        NRC Project Director: John F. Stolz
    
    Philadelphia Electric Company, Public Service Electric and Gas 
    Company,Delmarva Power and Light Company, and Atlantic City 
    Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom 
    Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania
    
        Date of application for amendments: January 13, 1995
        Description of amendment request: The proposed changes concern a 
    revision to the frequency of calibration for the Local Power Range 
    Monitor (LPRM) signals from every 6 weeks to every 2000 Megawatt Days 
    per Standard Ton (MWD/ST).
        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: [[Page 16196]] 
        1. The proposed change does not involve a significant increase 
    in the probability of consequences of an accident previously 
    evaluated.
        This change does not affect the operation of any equipment. The 
    change does not affect the fundamental method by which the LPRMs are 
    calibrated. The increased time between required LPRM calibrations 
    does not affect either the initiator of any accident previously 
    evaluated or any equipment required to mitigate the consequences of 
    an accident, or the isotopic inventory in the fuel. Thus, the change 
    does not increase either the probability or the radiological 
    consequences of an accident previously evaluated.
        2. The proposed changes do not create the possibility of a new 
    or different kind of accident from any previously evaluated.
        The proposed change does not introduce a new mode of plant 
    operation and does not involve the installation of any new equipment 
    or modifications to the plant. Therefore, it does not create the 
    possibility of a new or different kind of accident from any accident 
    previously evaluated.
        3. The proposed changes do not involve a significant reduction 
    in a margin of safety.
        The GETAB determination of the Maximum Critical Power Ratio 
    (MCPR) Safety Limit allows a maximum total nodal uncertainty of the 
    TIP readings (of which the LPRM Update uncertainty is a part) of 
    8.7%. The change in LPRM calibration frequency results in an LPRM 
    Update uncertainty of 4.2% nodal power. This, combined with the 
    other uncertainties which comprise the total TIP readings 
    uncertainty, yields a total TIP readings uncertainty of less than 
    the allowed 8.7%. Thus the change in LPRM calibration frequency will 
    not affect the MCPR Safety Limit.
        The LPRMs are utilized as input to the APRM and RBM systems. The 
    primary safety function of the APRM system is to initiate a scram 
    during core-wide neutron flux transients before the actual core-wide 
    neutron flux level exceeds the safety analysis design basis. This 
    prevents fuel damage from single operator errors or equipment 
    malfunctions. The APRMs are calibrated at least twice per week to 
    the plant heat balance, utilize a radially and axially diverse group 
    of LPRMs as input and are utilized to detect changes in average, not 
    local, power changes. Therefore, the effects of decreasing the LPRM 
    calibration frequency on the APRM system responses will be minimal 
    due to any individual LPRM drift being practically canceled out (due 
    to diversity of input) and/or due to the frequent recalibration of 
    the APRMs to an independent power calculation (the heat balance). 
    Thus, decreasing the LPRM calibration frequency will not 
    significantly impact the performance of the APRM system's scram 
    function, and there is no impact on transient delta-CPRs.
        The RBM system is utilized in the mitigation of a Rod Withdrawal 
    Error (RWE). The RBM system is designed to prevent the operator from 
    increasing the local power significantly when withdrawing a control 
    rod. On each selection of a control rod, the average of the 
    assigned, unbypassed LPRMs is adjusted to equal a 100% reference 
    signal for each of the two RBM channels. Each RBM channel 
    automatically limits the local thermal margin changes by limiting 
    the allowable change in local average neutron flux to the RBM 
    setpoint. If the local average neutron flux change is greater than 
    that allowed by the RBM setpoint, within either RBM channel, the rod 
    withdrawal permissive is removed preventing further movement. Since 
    the change in local neutron flux is calculated from the change in 
    the average of the LPRM readings, and calibrated on every rod 
    selection to the reference signal, offsets in individual LPRM 
    readings due to calibration differences are effectively eliminated 
    for a given RBM setpoint. Therefore, the constraints on the 
    withdrawal of any given rod are unchanged and there will not be any 
    increase in RWE delta-CPR.
        Since the MCPR Safety Limit is unaffected and the delta-CPR 
    values are unchanged, the cycle CPR limits are unchanged. Therefore, 
    the change in the frequency of LPRM calibration does not result in a 
    reduction in a margin of safet
    
    

Document Information

Published:
03/29/1995
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Document Number:
X95-20329
Pages:
16181-16196 (16 pages)
PDF File:
x95-20329.pdf