2018-06668. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations  

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    AGENCY:

    Nuclear Regulatory Commission.

    ACTION:

    Biweekly notice.

    SUMMARY:

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, 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 13, 2018, to March 26, 2018. The last biweekly notice was published on March 27, 2018.

    DATES:

    Comments must be filed by May 10, 2018. A request for a hearing must be filed by June 11, 2018.

    ADDRESSES:

    You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

    • Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0064. Address questions about NRC dockets to Jennifer Borges; telephone: 301-287-9127; email: Jennifer.Borges@nrc.gov. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document.
    • Mail Comments to: May Ma, Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

    For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

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    FOR FURTHER INFORMATION CONTACT:

    Kay Goldstein, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: kay.goldstein@nrc.gov.

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    SUPPLEMENTARY INFORMATION:

    I. Obtaining Information and Submitting Comments

    A. Obtaining Information

    Please refer to Docket ID NRC-2018-0064, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

    • Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0064.
    • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
    • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

    B. Submitting Comments

    Please include Docket ID NRC-2018-0064, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

    The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov,, as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.Start Printed Page 15413

    If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.

    II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination

    The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

    The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

    Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

    A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's website at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

    As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest.

    In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for 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 the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) 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. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures.

    Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document.

    If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish 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 would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

    A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from Start Printed Page 15414the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

    If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

    B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

    To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

    Information about applying for a digital ID certificate is available on the NRC's public website at http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public website at http://www.nrc.gov/​site-help/​electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.

    A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the “Contact Us” link located on the NRC's public website at http://www.nrc.gov/​site-help/​e-submittals.html,, by email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.

    Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

    Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/​ehd,, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly-available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate Start Printed Page 15415proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

    For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

    Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP), Darlington County, South Carolina

    Date of amendment request: February 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18038B289.

    Description of amendment request: The amendment would revise Technical Specification (TS) Section 3.4.3, “RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,” to reduce the applicability terms from 50 effective full power years (EFPY) to 46.3 EFPY in Figures 3.4.3-1 and 3.4.3-2, as a result of the removal of part length fuel assemblies (PLSAs) and the migration to 24-month fuel cycles.

    Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

    Response: No.

    The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. The proposed change does not involve physical changes to the plant or alter the reactor coolant system (RCS) pressure boundary (i.e., there are no changes in operating pressure, materials or seismic loading). The P/T limit curves and Adjusted Reference Temperature (ART) values will remain as-is. Only the term to which the limit curves applies is effected by the proposed change. The P/T limit curves in TS 3.4.3 with an applicability term of 46.3 EFPY provide continued assurance that the fracture toughness of the reactor pressure vessel (RPV) is consistent with analysis assumptions and NRC regulations. The methodology used to develop the existing P/T limit curves provides assurance that the probability of a rapidly propagating failure will be minimized. The P/T limit curves, with the applicability term reduced to a proposed 46.3 EFPY, will continue to prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained.

    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?

    Response: No.

    The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. The proposed change does not affect the design or assumed accident performance of any structure, system or component, or introduce any new modes of system operation or failure modes. Compliance with the proposed P/T curves (same as the existing P/T curves with the applicability term reduced to 46.3 EFPY) will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated.

    Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

    3. Does the proposed change involve a significant reduction in a margin of safety?

    Response: No.

    The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. HBRSEP adheres to applicable NRC regulations (i.e., 10 CFR 50, Appendices G and H) and NRC-approved methodologies (i.e., Regulatory Guides 1.99 and 1.190) with respect to the P/T limit curves in TS 3.4.3 in order to provide an adequate margin of safety to the conditions at which brittle fracture may occur. The P/T limit curves, with the applicability term reduced to 46.3 EFPY, continue to provide assurance that the established P/T limits are not exceeded.

    Therefore, 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.

    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street, DEC45A, Charlotte NC 28202.

    NRC Acting Branch Chief: Brian W. Tindell.

    Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: February 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18039A123.

    Description of amendment request: LSCS Technical Specifications (TS) 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” currently requires performance of Surveillance Requirement (SR) 3.6.1.3.8 on each excess flow check valve (EFCV) during each refueling outage. The proposed amendments would revise the number of EFCVs tested by TS SR 3.6.1.3.8 from “each” to a “representative sample.” The representative sample is based on approximately 20 percent of the reactor instrumentation line EFCVs such that each EFCV will be tested at least once every 10 years (nominal). Therefore, approximately 20 percent of the EFCVs will be tested every operating cycle.

    The reduced testing associated with the proposed change will result in an increase in the availability of the associated instrumentation during outages and will result in dose savings.

    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 analyzed?

    Response: No.

    The EFCVs at LSCS, Unit 1 and Unit 2, are designed so that they will not close accidently during normal operations, will close if a rupture of the instrument line is indicated downstream of the valve, can be reopened when appropriate, and have their status indicated in the control room. This proposed change relaxes the number of EFCVs tested for TS SR 3.6.1.3.8 from “each” to a “representative sample” in accordance with the SFCP [Surveillance Frequency Control Program]. There are no physical plant modifications associated with this change. Industry and LSCS operating experience demonstrate a high reliability of these valves. Neither EFCVs nor their failures are capable of initiating previously evaluated accidents; therefore, there can be no increase in the probability of occurrence of an accident regarding this proposed change.

    The LSCS Updated Final Safety Analysis Report (UFSAR) demonstrates, consistent with BWROG [Boiling Water Reactor Owners Group] topical report NEDO-32977-A, that the failure of an EFCV has very low Start Printed Page 15416consequences. The LSCS UFSAR evaluates a circumferential rupture of an instrument line that is connected to the primary coolant system. The evaluation credits the 0.25-inch diameter flow-restricting orifice installed in the line with limiting flow following the instrumentation line break and does not credit the EFCV with actuating to limit leakage. The dose consequences of the instrument line break are determined using the calculated mass of coolant released over approximately a five-hour period. The reactor was assumed to be operating at design power conditions prior to the break. The Standby Gas Treatment System (SGTS) and secondary containment are not impaired by the event. The evaluation concludes that the consequences of the event are well within 10 CFR 100 limits. Thus, the failure of an EFCV, though not expected as a result of the proposed change, does not affect the dose consequences of an instrument line break.

    Based on the above, it is concluded that the proposed change to the EFCV surveillance requirement does not involve a significant increase in the probability or consequences of an accident previously analyzed.

    2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    This proposed change allows a reduced number of EFCVs to be tested in accordance with the SFCP [Surveillance Frequency Control Program]. The proposed change would revise SR 3.6.1.3.8 to verify that a “representative sample” (i.e., approximately 20 percent) of reactor instrumentation line EFCVs are tested, in accordance with the SFCP, such that each EFCV will be tested at least once every 10 years (nominal). No other changes in the requirements are being proposed. Industry and LSCS-specific operating experience demonstrates the high degree of reliability of the EFCVs and the low consequences of an EFCV failure. The potential failure of an EFCV to isolate by the proposed reduction in test frequency is bounded by the previous evaluation of an instrument line rupture. This change will not alter the operation or process variables, structures, systems, or components as described in the safety analysis. Thus, a new or different kind of accident will not be created from implementation of the proposed change.

    Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

    3. Does the proposed change involve a significant reduction in a margin of safety?

    Response: No.

    The proposed changes do not involve a significant reduction in the margin of safety. The LSCS UFSAR evaluates a circumferential rupture of an instrument line that is connected to the primary coolant system. The evaluation credits the 0.25-inch diameter flow-restricting orifice installed in the line with limiting flow following the instrumentation line break and does not credit the EFCV with actuating to limit leakage. The dose consequences of the instrument line break are determined using the calculated mass of coolant released over approximately a five-hour period. The reactor was assumed to be operating at design power conditions prior to the break. The SGTS [Standby Gas Treatment System] and secondary containment are not impaired by the event. The evaluation concludes that the consequences of the event are well within 10 CFR 100 limits. Thus, the failure of an EFCV, though not expected as a result of the proposed change, does not affect the dose consequences of an instrument line break.

    Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

    Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

    NRC Branch Chief: David J. Wrona.

    Exelon Generation Company, LLC, (EGC) Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: February 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18058A257.

    Description of amendment request: The proposed amendments would revise LSCS Technical Specifications (TS) 3.4, Reactor Coolant System (RCS), Section 3.4.4, “Safety/Relief Valves (S/RVs).”

    Specifically, EGC proposes a new safety function lift setpoint lower tolerance for the S/RVs as delineated in Surveillance Requirement 3.4.4.1. The proposed change will revise the lower setpoint tolerances from −3 percent (%) to −5%.

    This change is limited to the lower tolerances and does not affect the upper tolerances; therefore, the upper tolerance will remain at +3% of the safety function lift setpoint. In addition, this change only applies to the as-found tolerance and not to the as-left tolerance, which will remain unchanged at ±1% of the safety lift setpoint. The as-found tolerances are used for determining operability and to increase sample sizes for S/RV testing should the tolerance be exceeded. There will be no revision to the actual setpoints of the valves installed in the plant due to this change.

    This proposed change will preclude the submittal of previously-reportable licensee event reports (LERs) to the NRC due to setpoint drift in the low (conservative) direction.

    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. Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated?

    Response: No.

    This change has no influence on the probability or consequences of any accident previously evaluated. The lower setpoint tolerance change does not affect the operation of the valves and it does not change the as-left setpoint tolerance. The change only affects the lower tolerance for valve opening and does not change the upper tolerance, which is the limit that protects from overpressurization.

    The proposed amendments do not involve physical changes to the valves, nor do they change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components.

    The proposed amendments do not change any other behavior or operation of any safety/relief valves (S/RVs), and, therefore, has no significant impact on reactor operation. They also have no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the Updated Final Safety Analysis Report (UFSAR).

    Based on the above, it is concluded that the proposed change to the S/RV surveillance requirement does not involve a significant increase in the probability or consequences of an accident previously evaluated.

    2. Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    The proposed change to the S/RV safety lower setpoint tolerance from −3% to −5% only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance.

    The proposed lower setpoint tolerance change does not adversely affect the operation of any safety-related components or equipment. The proposed amendments do not involve physical changes to the S/RVs, nor do they change the safety function of the S/RVs. The proposed amendments do not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed, and installed equipment is not being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure Start Printed Page 15417the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis.

    Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

    3. Do the proposed amendments involve a significant reduction in a margin of safety?

    Response: No.

    The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the S/RV setpoint upper setpoint tolerance. The TS setpoints for the S/RVs are not changed. The as-left setpoint tolerances are not changed by this proposed change and remain at ±1% of the safety lift setpoint.

    The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not significantly impact the condition or performance of structures, systems, and components relied upon for accident mitigation.

    Therefore, this proposed change does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

    Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

    NRC Branch Chief: David J. Wrona.

    Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

    Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

    Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249, Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, Illinois

    Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

    Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: January 31, 2018. A publicly-available version is in ADAMS under Package Accession No. ML18053A159.

    Description of amendment request: The amendments would revise the emergency response organization (ERO) positions identified in the emergency plan for each site.

    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 for each site, which is presented below:

    1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

    Response: No.

    The proposed changes to the [site] Emergency Plan do not increase the probability or consequences of an accident. The proposed changes do not impact the function of plant Structures, Systems, or Components (SSCs). The proposed changes do not affect accident initiators or accident precursors, nor do the changes alter design assumptions. The proposed changes do not alter or prevent the ability of the onsite ERO to perform their intended functions to mitigate the consequences of an accident or event. The proposed changes remove ERO positions no longer credited or considered necessary in support of Emergency Plan implementation.

    Therefore, the proposed changes to the [site] Emergency Plan do not involve a significant increase in the probability or consequences of an accident previously evaluated.

    2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    The proposed changes have no impact on the design, function, or operation of any plant SSCs. The proposed changes do not affect plant equipment or accident analyses. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed changes do not introduce failure modes that could result in a new accident, and the proposed changes do not alter assumptions made in the safety analysis. The proposed changes remove ERO positions no longer credited or considered necessary in support of Emergency Plan implementation.

    Therefore, the proposed changes to the [site] Emergency Plan do not create the possibility of a new or different kind of accident from any accident previously evaluated.

    3. Does the proposed amendment involve a significant reduction in a margin of safety?

    Response: No.

    Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public.

    The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. Margins of safety are unaffected by the proposed changes to the ERO staffing.

    The proposed changes are associated with the [site] Emergency Plan staffing and do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by these proposed changes. The proposed changes to the Emergency Plan will continue to provide the necessary onsite ERO response staff.

    Therefore, the proposed changes to the [site] Emergency Plan do not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis for each site 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.

    Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

    NRC Branch Chief: David J. Wrona.

    Florida Power & Light Company, Docket Nos. 50-250 and 251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

    Date of application for amendment: June 28, 2017, as supplemented by letter dated February 28, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML17180A447 and ML18075A023, respectively.

    Description of amendment request: The amendments would modify the Technical Specifications (TSs) by Start Printed Page 15418relocating to licensee-controlled documents select acceptance criteria specified in TS surveillance requirements (SRs) credited for satisfying Inservice Testing (IST) Program and Inservice Inspection Program requirements; deleting the SRs for the American Society of Mechanical Engineers Code Class 1, 2, and 3 components; replacing references to the Surveillance Frequency Control Program (SFCP) with reference to the Turkey Point IST Program where appropriate; establishing a Reactor Coolant Pump (RCP) Flywheel Inspection Program; and related editorial changes. Additionally, the amendments would delete a redundant SR for Accumulator check valve testing and add a footnote to the SR for Pressure Isolation Valve (PIV) testing.

    The license amendment request was originally noticed in the Federal Register on August 29, 2017 (82 FR 41069). The notice is being reissued in its entirety to include the revised scope, description of the amendment request, and proposed no significant hazards consideration determination.

    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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

    Response: No.

    The proposed changes provide assurance that inservice testing will be performed in the manner and within the timeframes established by 10 CFR 50.55(a). The deletion of SR 4.0.5 and the deletion of IST acceptance criteria from SR 4.5.2.c and SR 4.6.2.1.b neither affect the conduct nor the periodicity of the inservice testing. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative in nature and can neither initiate nor affect the outcome of any accident previously evaluated. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot affect the likelihood or the outcome of accident previously evaluated. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and related editorial changes are administrative changes and cannot affect the likelihood or the outcome of any accident previously evaluated. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not affect the likelihood or the outcome of any accident previously evaluated.

    Therefore, facility operation in accordance with the proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated.

    2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    The deletion of IST acceptance criteria from the TS does not affect the manner in which any SSC [structure, system, or component] is maintained or operated and does not introduce new SSCs or new methods for maintaining existing plant SSCs. Inservice testing will continue in the manner and periodicity specified in the IST program such that no new or different kind of accident can result. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative changes and cannot introduce new or different kinds of accidents. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot be an initiator of a new or different kind of accident. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and the other editorial changes are administrative changes and cannot introduce new or different kinds of accidents. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not introduce new or different kinds of accidents.

    Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

    3. Does the proposed amendment involve a significant reduction in a margin of safety?

    Response: No.

    The proposed changes do not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings and do not adversely impact plant operating margins or the reliability of equipment credited in safety analyses. The proposed changes provides assurance that inservice inspection and inservice testing will be performed in the manner and within the timeframes established by 10 CFR 50.55(a). The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes with no impact on the margin of safety currently described the Updated Final Safety Analysis Report. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and the other editorial changes are administrative changes with no impact on nuclear safety. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not affect any safety analyses assumptions, safety limits, or limiting safety system settings.

    Therefore, operation of the facility in accordance with the proposed changes will 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.

    Attorney for licensee: Debbie Hendel, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB, Juno Beach, FL 33408-0420.

    NRC Acting Branch Chief: Brian W. Tindell.

    Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant (BFN), Unit 1, Limestone County, Alabama

    Date of amendment request: March 16, 2018. A publicly-available version is in ADAMS under Accession No. ML18080A171.

    Description of amendment request: The amendment would revise License Condition 2.C(18)(a)3 for Unit 1 that requires the submittal of a revised BFN Unit 1 replacement steam dryer (RSD) analysis utilizing the BFN Unit 3 on-dryer strain gauge based end-to-end bias and uncertainties at extended power conditions “at least 90 days prior to the start of the BFN Unit 1 EPU [extended power uprate] outage.” Specifically, the amendment reduces the time from 90 days to 15 days before the BFN Unit 1 EPU outage for the submittal of the revised analysis of the RSD.

    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 amendment involve a significant increase in the probability or consequence of an accident previously evaluated?

    Response: No.

    The proposed license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the Browns Ferry Nuclear Plant (BFN) Unit 1 replacement steam dryer (RSD), performed using an NRC-approved methodology benchmarked on the BFN Unit 3 RSD, must be submitted to the NRC for Start Printed Page 15419information. There is no required review or approval of the revised analysis needed to satisfy the license condition. The proposed change is an administrative change to the period before the outage and does not impact any system, structure or component in such a way as to affect the probability or consequences of an accident previously evaluated. The proposed amendment is purely administrative and has no technical or safety aspects. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

    2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    The proposed license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the BFN Unit 1 RSD must be submitted to the NRC for information. The proposed amendment is purely administrative and has no technical or safety aspects. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

    3. Does the proposed amendment involve a significant reduction in a margin of safety?

    Response: No.

    The proposed license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the BFN Unit 1 RSD must be submitted to the NRC for information. The proposed amendment is purely administrative and has no technical or safety aspects. Therefore, 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.

    Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.

    NRC Acting Branch Chief: Brian W. Tindell.

    III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses

    During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment.

    A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.

    Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

    For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment, as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

    Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: July 31, 2017.

    Brief description of amendment: The amendment revised the license to authorize the description of the emergency response organization requalification training frequency defined in the Emergency Plan to be changed from “annually” to “once per calendar year not to exceed 18 months between training sessions.”

    Date of issuance: March 26, 2018.

    Effective date: As of the date of its issuance and shall be implemented within 90 days.

    Amendment No.: 318. A publicly-available version is in ADAMS under Accession No. ML17289A175; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

    Renewed Facility Operating License No. DPR-59: The amendment revised the Renewed Facility Operating License and Emergency Plan.

    Date of initial notice in Federal Register: September 26, 2017 (82 FR 44854).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2018.

    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

    Date of amendment request: May 1, 2017, as supplemented by letters dated November 15 and December 20, 2017.

    Brief description of amendment: The amendment replaced existing technical specification requirements related to “operations with a potential for draining the reactor vessel” with new requirements on reactor pressure vessel water inventory control to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor pressure vessel water level to be greater than the top of active irradiated fuel. The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control.”

    Date of issuance: March 22, 2018.

    Effective date: As of the date of issuance and shall be implemented prior to entering Mode 4 during the next refueling outage, C1R18, currently planned for April 2018.

    Amendment No.: 216. A publicly-available version is in ADAMS under Accession No. ML18043A505; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

    Facility Operating License No. NPF-62: The amendment revised the Facility Operating License and Technical Specifications.

    Date of initial notice in Federal Register: July 5, 2017 (82 FR 31096). The supplement letters dated November 15, 2017, and December 20, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 22, 2018.

    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: November 3, 2017.

    Brief description of amendment: The amendment changed the safety limit Start Printed Page 15420minimum critical power ratio numeric values for Operating Cycle 17. Specifically, the amendment increased the numeric values of the safety limit minimum critical power ratio for Nine Mile Point Nuclear Station, Unit 2, from ≥1.15 to ≥1.17 for two recirculation loop operation, and from ≥1.15 to ≥1.17 for single recirculation loop operation.

    Date of issuance: March 16, 2018.

    Effective date: As of the date of issuance and shall be implemented prior to startup from the next refueling outage.

    Amendment No.: 167. A publicly-available version is in ADAMS under Accession No. ML18060A016; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

    Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and Technical Specifications.

    Date of initial notice in Federal Register: February 6, 2018 (83 FR 5280).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2018.

    No significant hazards consideration comments received: No.

    Florida Power & 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: January 31, 2018.

    Brief description of amendments: The amendments revised the Emergency Plan for St. Lucie to adopt the fire-related notification of unusual event requirement of the Nuclear Energy Institute 99-01, Revision 6, Emergency Action Level scheme.

    Date of issuance: March 26, 2018.

    Effective date: As of the date of issuance and shall be implemented within 90 days.

    Amendment Nos.: 244 and 195. A publicly-available version is in ADAMS under Accession No. ML18046A712; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License Nos. DPR-67 and NPF-16: The amendments revised the Renewed Facility Operating Licenses.

    Date of initial notice in Federal Register: February 14, 2018 (83 FR 6621). This notice provided an opportunity to request a hearing by April 15, 2018, but indicated that if the Commission makes a final no significant hazards consideration determination, any such hearing would take place after issuance of the amendments.

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 2018.

    No significant hazards consideration comments received: No.

    Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

    Date of amendment request: April 9, 2017, as supplemented by letter dated October 4, 2017.

    Brief description of amendments: The amendments revised the Technical Specifications (TSs) to remove various reporting requirements. Specifically, the amendments removed the requirements to prepare the Startup Report, the Annual Report, and various special reports. In addition, the amendments revised the TSs to remove the completion time for restoring spent fuel pool water level, to address inoperability of one of the two parallel flow paths in the residual heal removal or safely injection headers for the Emergency Core Cooling Systems, and to make other administrative changes, including updating plant staff and responsibilities.

    Date of issuance: March 19, 2018.

    Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.

    Amendment Nos.: 279 (Unit No. 3) and 274 (Unit No. 4). A publicly-available version is in ADAMS under Accession No. ML18019A078; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments.

    Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs.

    Date of initial notice in Federal Register: June 19, 2017 (82 FR 27889). The supplemental letter dated October 4, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

    The Commission's related evaluation of the amendments is contained in an SE dated March 19, 2018.

    No significant hazards consideration comments received: No.

    Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia

    Date of amendment request: August 31, 2017.

    Description of amendment: The amendments authorized changes to the VEGP Units 3 and 4 Combined Operating License (COL) page 7 and COL Appendix A, Technical Specifications, to make necessary changes so that there will be adequate detection of reactor coolant system and main steam line leakage at all times and that the associated limits account for instrumentation sensitivities not accounted for in the current VEGP Technical Specification 3.4.9.

    Date of issuance: March 12, 2018.

    Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

    Amendment No.: 115 (Unit 3) and 114 (Unit 4). Publicly-available versions are in ADAMS Package Accession No. ML18036A782, which includes the Safety Evaluation that references documents related to these amendments.

    Facility Combined License Nos. NPF-91 and NPF-92: Amendments revised the Facility Combined Licenses.

    Date of initial notice in Federal Register: October 10, 2017 (82 FR 47032).

    The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated March 12, 2018.

    No significant hazards consideration comments received: No.

    Start Signature

    Dated at Rockville, Maryland, this 28th day of March 2018.

    For the Nuclear Regulatory Commission.

    Tara Inverso,

    Acting Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

    End Signature End Supplemental Information

    [FR Doc. 2018-06668 Filed 4-9-18; 8:45 am]

    BILLING CODE 7590-01-P

Document Information

Published:
04/10/2018
Department:
Nuclear Regulatory Commission
Entry Type:
Notice
Action:
Biweekly notice.
Document Number:
2018-06668
Dates:
Comments must be filed by May 10, 2018. A request for a hearing must be filed by June 11, 2018.
Pages:
15412-15420 (9 pages)
Docket Numbers:
NRC-2018-0064
PDF File:
2018-06668.Pdf