2018-03727. 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 January 30, 2018, to February 12, 2018. The last biweekly notice was published on February 13, 2018.

    DATES:

    Comments must be filed by March 29, 2018. A request for a hearing must be filed by April 30, 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-2018-0031. Address questions about NRC dockets to Jennifer Borges; telephone: 301-287-9127; email: Jennifer.Borges@nrc.gov. For technical questions, contact the individual 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:

    Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: Paula.Blechman@nrc.gov.

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    SUPPLEMENTARY INFORMATION:

    I. Obtaining Information and Submitting Comments

    A. Obtaining Information

    Please refer to Docket ID NRC-2018-0031, 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-2018-0031.
    • 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 DAMS) 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-2018-0031, 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.

    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 Start Printed Page 8510issuance. 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 the 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 Start Printed Page 8511accordance 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 proximity 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.

    DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: October 9, 2017. A publicly-available version is in ADAMS under Accession No. ML17283A248.

    Description of amendment request: The amendment would revise Limiting Condition for Operation (LCO) 3.10.1, to expand its scope to include provisions for temperature excursions greater than 200 degrees Fahrenheit (°F) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4. This change is consistent with NRC approved Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-484, “Use of TS 3.10.1 for Scram Time Testing Activities,” Revision 0.

    The NRC staff issued a Notice of Availability for TSTF-484 in the Federal Register on October 27, 2006 (71 FR 63050). The staff also issued a Federal Register notice on August 21, 2006 (71 FR 48561), that provided a model safety evaluation and a model no significant hazards consideration (NSHC) determination that licensees could reference in their plant-specific application. In its application dated October 9, 2017, the licensee affirmed the applicability of the model NSHC determination for Fermi 2.

    Basis for proposed no NSHC determination: As required by 10 CFR 50.91(a), the licensee affirmed the applicability of the model NSHC, which is presented below:

    Criterion 1: The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.Start Printed Page 8512

    Technical Specifications currently allow for operation at greater than 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

    Criterion 2: The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

    Technical Specifications currently allow for operation at greater than 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

    Criterion 3: The proposed change does not involve a significant reduction in a margin of safety.

    Technical Specifications currently allow for operation at greater than 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the above 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: Jon P. Christinidis, DTE Energy, Expert Attorney—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

    NRC Branch Chief: David J. Wrona.

    Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina

    Date of amendment request: May 2, 2017, as supplemented by letters dated July 20 and November 21, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588, respectively.

    Description of amendment request: The amendments would modify CNS Technical Specifications (TSs) to extend the Completion Time (CT) of TS 3.8.1, “AC [Alternating Current] Sources—Operating,” Required Action B.6 (existing Required Action B.4, numbered as B.6) for an inoperable emergency diesel generator (DG) from 72 hours to 14 days. A conforming change is also proposed to extend the maximum CT of TS 3.8.1 Required Actions A.3 and B.4. To support this request, the licensee will add a supplemental power source (i.e., two supplemental diesel generators (SDGs) per station) with the capability to power any emergency bus. The SDGs will have the capacity to bring the affected unit to cold shutdown. Additionally, the amendments would modify TS 3.8.1 to add new two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO 3.8.1.d, to ensure that at least one train of shared components has an operable emergency power supply.

    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 change involves extending the TS CT for an inoperable DG at CNS [. . .]. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at CNS [. . .] has an operable emergency power supply whenever one DG is inoperable. The DGs at both stations are safety related components which provide a backup electrical power supply to the onsite emergency power distribution system. The proposed change does not affect the design of the DGs, the operational characteristics or function of the DGs, the interfaces between the DGs and other plant systems or the reliability of the DGs. The DGs are not accident initiators; the DGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power. Extending the CT for a single DG would not affect the previously evaluated accidents since the remaining DGs supporting the redundant engineered safety feature systems would continue to be available to perform the accident mitigation functions. Thus, allowing a DG to be inoperable for an additional 11 days for performance of maintenance or testing does not increase the probability of a previously evaluated accident.

    Deterministic and probabilistic risk assessment techniques evaluated the effect of the proposed TS change to extend the CT for an inoperable DG on the availability of an electrical power supply to the plant emergency safeguards feature systems. These assessments concluded that the proposed CNS [. . .] TS change does not involve a significant increase in the risk of power supply unavailability.

    There is a small incremental risk associated with continued operation for an additional 11 days with one DG inoperable; however, the calculated impact provides risk metrics consistent with the acceptance guidelines contained in Regulatory Guides 1.177 and 1.174. The remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. A DG is required to operate only if both offsite power sources fail and there is an event which requires operation of the plant engineered safety features such as a design basis accident. The probability of a design basis accident occurring during this period is low.

    The consequences of previously evaluated accidents will remain the same during the proposed 14 day CT as during the current CNS [. . .] 72 hour CT. The ability of the remaining TS required DGs to mitigate the consequences of an accident will not be affected since no additional failures are postulated while equipment is inoperable within the TS CT.

    Regarding the proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply, there is no change to how or under what conditions offsite circuits or DGs are operated nor are there any changes to acceptable operating parameters. Power source operability requirements for shared components are being moved from the TS Bases to TS with the proposed change. The proposed change will ensure that at least one train of shared components has an operable emergency power supply whenever a DG is inoperable.

    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 change involves extending the TS CT for an inoperable DG at CNS [. . .]. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at CNS [. . .] has an operable emergency power supply whenever one DG is inoperable.

    The proposed change does not involve a change in the CNS [. . .] plant design, plant Start Printed Page 8513configuration, system operation or procedures involved with the DGs. The proposed change allows a DG to be inoperable for additional time. Equipment will be operated in the same configuration and manner that is currently allowed and designed for. The functional demands on credited equipment is unchanged. There are no new failure modes or mechanisms created due to plant operation for an extended period to perform DG maintenance or testing. Extended operation with an inoperable DG does not involve any modification to the operational limits or physical design of plant systems. There are no new accident precursors generated due to the extended CT.

    Regarding the proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply, there is no change to how or under what conditions offsite circuits or DGs are operated nor are there any changes to acceptable operating parameters. Power source operability requirements for shared components are being moved from the TS Bases to TS with the proposed change. The proposed change will ensure that at least one train of shared components has an operable emergency power supply whenever a DG is inoperable. This change does not alter the nature of events postulated in the Updated Final Safety Analysis Report nor does it introduce any unique precursor mechanisms.

    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 the margin of safety?

    Response: No.

    The proposed change involves extending the TS CT for an inoperable DG at CNS [. . .]. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at CNS [. . .] has an operable emergency power supply whenever one DG is inoperable.

    Currently, if an inoperable DG is not restored to operable status within 72 hours at CNS [. . .], TS 3.8.1, requires the units to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The proposed TS changes will allow steady state plant operation at 100 percent power for an additional 11 days for performance of DG planned reliability improvements and preventive and corrective maintenance.

    Deterministic and probabilistic risk assessment techniques evaluated the effect of the proposed TS change to extend the CT for an inoperable DG on the availability of an electrical power supply to the plant emergency safeguards feature systems. These assessments concluded that the proposed CNS [. . .] TS change does not involve a significant increase in the risk of power supply unavailability.

    The DGs continue to meet their design requirements; there is no reduction in capability or change in design configuration. The DG response to loss of offsite power, loss of coolant accident, station blackout or fire scenarios is not changed by this proposed amendment; there is no change to the DG operating parameters. In the extended CT, as in the existing CT, the remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. The proposed change to extend the CT for an inoperable DG does not alter a design basis safety limit; therefore, it does not significantly reduce the margin of safety. The DGs will continue to operate per the existing design and regulatory requirements.

    The proposed TS changes (i.e., the inoperable DG CT extension request and proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply) do not alter the plant design nor do they change the assumptions contained in the safety analyses. The standby AC power system is designed with sufficient redundancy such that a DG may be removed from service for maintenance or testing. The remaining DGs are capable of carrying sufficient electrical loads to satisfy the Updated Final Safety Analysis Report requirements for accident mitigation or unit safe shutdown. The proposed change does not impact the redundancy or availability requirements of offsite power circuits or change the ability of the plant to cope with a station blackout. Therefore, based on the considerations given above, the proposed changes do not involve a significant reduction in the margin of safety.

    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: Kate B. Nolan, Deputy General Counsel, Duke Energy Carolinas, LLC, 550 South Tryon Street—DEC45A, Charlotte, NC 28202-1802.

    NRC Branch Chief: Michael T. Markley.

    Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North Carolina

    Date of amendment request: May 2, 2017, as supplemented by letters dated July 20 and November 21, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588, respectively.

    Description of amendment request: The amendments would modify MNS Technical Specifications (TSs) to extend the Completion Time (CT) of TS 3.8.1, “AC [Alternating Current] Sources—Operating,” Required Action B.6 (existing Required Action B.4, numbered as B.6) for an inoperable emergency diesel generator (DG) from 72 hours to 14 days. A conforming change is also proposed to extend the maximum CT of TS 3.8.1 Required Actions A.3 and B.4. To support this request, the licensee will add a supplemental power source (i.e., two supplemental diesel generators (SDGs) per station) with the capability to power any emergency bus. The SDGs will have the capacity to bring the affected unit to cold shutdown. Additionally, the amendments would modify TS 3.8.1 to add new two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO 3.8.1.d, to ensure that at least one train of shared components has an operable emergency power supply.

    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 change involves extending the TS CT for an inoperable DG at [. . .] MNS. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at [. . .] MNS has an operable emergency power supply whenever one DG is inoperable. The DGs at both stations are safety related components which provide a backup electrical power supply to the onsite emergency power distribution system. The proposed change does not affect the design of the DGs, the operational characteristics or function of the DGs, the interfaces between the DGs and other plant systems or the reliability of the DGs. The DGs are not accident initiators; the DGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power. Extending the CT for a single DG would not affect the previously evaluated accidents since the remaining DGs supporting the redundant engineered safety feature systems would continue to be available to perform the accident mitigation functions. Thus, allowing a DG to be inoperable for an additional 11 days for performance of maintenance or testing does not increase the probability of a previously evaluated accident.

    Deterministic and probabilistic risk assessment techniques evaluated the effect of the proposed TS change to extend the [completion time] CT for an inoperable DG on the availability of an electrical power supply to the plant emergency safeguards feature systems. These assessments concluded that the proposed [. . .] MNS TS change does not involve a significant increase in the risk of power supply unavailability.Start Printed Page 8514

    There is a small incremental risk associated with continued operation for an additional 11 days with one DG inoperable; however, the calculated impact provides risk metrics consistent with the acceptance guidelines contained in Regulatory Guides 1.177 and 1.174.

    The remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. A DG is required to operate only if both offsite power sources fail and there is an event which requires operation of the plant engineered safety features such as a design basis accident. The probability of a design basis accident occurring during this period is low.

    The consequences of previously evaluated accidents will remain the same during the proposed 14 day CT as during the current [. . .] MNS 72 hour CT. The ability of the remaining TS required DGs to mitigate the consequences of an accident will not be affected since no additional failures are postulated while equipment is inoperable within the TS CT.

    Regarding the proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply, there is no change to how or under what conditions offsite circuits or DGs are operated nor are there any changes to acceptable operating parameters. Power source operability requirements for shared components are being moved from the TS Bases to TS with the proposed change. The proposed change will ensure that at least one train of shared components has an operable emergency power supply whenever a DG is inoperable.

    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 change involves extending the TS CT for an inoperable DG at [. . .] MNS. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at [. . .] MNS has an operable emergency power supply whenever one DG is inoperable.

    The proposed change does not involve a change in the [. . .] MNS plant design, plant configuration, system operation or procedures involved with the DGs. The proposed change allows a DG to be inoperable for additional time. Equipment will be operated in the same configuration and manner that is currently allowed and designed for. The functional demands on credited equipment is unchanged. There are no new failure modes or mechanisms created due to plant operation for an extended period to perform DG maintenance or testing. Extended operation with an inoperable DG does not involve any modification to the operational limits or physical design of plant systems. There are no new accident precursors generated due to the extended CT.

    Regarding the proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply, there is no change to how or under what conditions offsite circuits or DGs are operated nor are there any changes to acceptable operating parameters. Power source operability requirements for shared components are being moved from the TS Bases to TS with the proposed change. The proposed change will ensure that at least one train of shared components has an operable emergency power supply whenever a DG is inoperable. This change does not alter the nature of events postulated in the Updated Final Safety Analysis Report nor does it introduce any unique precursor mechanisms.

    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 the margin of safety?

    Response: No.

    The proposed change involves extending the TS CT for an inoperable DG at [. . .] MNS. The proposed change also involves adding a new Required Action to TSs to ensure that at least one train of shared components at [. . .] MNS has an operable emergency power supply whenever one DG is inoperable.

    Currently, if an inoperable DG is not restored to operable status within 72 hours at [. . .] MNS, TS 3.8.1, requires the units to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The proposed TS changes will allow steady state plant operation at 100 percent power for an additional 11 days for performance of DG planned reliability improvements and preventive and corrective maintenance.

    Deterministic and probabilistic risk assessment techniques evaluated the effect of the proposed TS change to extend the CT for an inoperable DG on the availability of an electrical power supply to the plant emergency safeguards feature systems. These assessments concluded that the proposed [. . .] MNS TS change does not involve a significant increase in the risk of power supply unavailability.

    The DGs continue to meet their design requirements; there is no reduction in capability or change in design configuration. The DG response to loss of offsite power, loss of coolant accident, station blackout or fire scenarios is not changed by this proposed amendment; there is no change to the DG operating parameters. In the extended CT, as in the existing CT, the remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. The proposed change to extend the CT for an inoperable DG does not alter a design basis safety limit; therefore, it does not significantly reduce the margin of safety. The DGs will continue to operate per the existing design and regulatory requirements.

    The proposed TS changes (i.e., the inoperable DG CT extension request and proposed change to add Required Action to ensure that at least one train of shared components has an operable emergency power supply) do not alter the plant design nor do they change the assumptions contained in the safety analyses. The standby AC power system is designed with sufficient redundancy such that a DG may be removed from service for maintenance or testing. The remaining DGs are capable of carrying sufficient electrical loads to satisfy the Updated Final Safety Analysis Report requirements for accident mitigation or unit safe shutdown. The proposed change does not impact the redundancy or availability requirements of offsite power circuits or change the ability of the plant to cope with a station blackout. Therefore, based on the considerations given above, the proposed changes do not involve a significant reduction in the margin of safety.

    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: Kate B. Nolan, Deputy General Counsel, Duke Energy Carolinas, LLC, 550 South Tryon Street—DEC45A Charlotte, NC 28202-1802.

    NRC Branch Chief: Michael T. Markley.

    Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2 (ANO-2), Pope County, Arkansas

    Date of amendment request: November 20, 2017. A publicly-available version is in ADAMS under Accession No. ML17326A387.

    Description of amendment request: The amendment would revise the Technical Specifications (TSs) to replace the current pressure-temperature limits for heatup, cooldown, and the inservice leak hydrostatic tests for the reactor coolant system presented in TS 3.4.9 that expire at 32 Effective Full Power Years (EFPY) with limitations that extend out to 54 EFPY.

    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 will revise the pressure-temperature (P-T) limits for heatup, cooldown, and inservice leak hydrostatic test limitations for the Reactor Coolant System Start Printed Page 8515(RCS) to a maximum of 54 Effective Full Power Years (EFPY) in accordance with 10 CFR 50, Appendix G. This is the end of the period of extended operation for the renewed ANO-2 operating License. The P-T limits were developed in accordance with the requirements of 10 CFR 50, Appendix G, utilizing the analytical methods and flaw acceptance criteria of Topical Report WCAP-14040, Revision 4, and American Society of Mechanical Engineers (ASME) Code, Section XI, Appendix G. These methods and criteria are the previously NRC approved standards for the preparation of P-T limits. Updating the P-T limits for additional EFPYs maintains the level of assurance that reactor coolant pressure boundary integrity will be maintained, as specified in 10 CFR 50, Appendix G.

    The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

    Therefore, this 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 changes implement methodologies that have been approved by the NRC (provided that any conditions/limitations are satisfied). The P-T limits will ensure the protection consistent with assuring the integrity of the reactor coolant pressure boundary as was previously evaluated. Reactor coolant pressure boundary integrity will continue to be maintained in accordance with 10 CFR 50, Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. These changes do not involve any physical alteration of the plant (i.e., no new or different type of equipment will be installed), and installed equipment is not being operated in a new or different manner. Thus, no new failure modes are introduced.

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

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

    Response: No.

    The proposed changes do not affect the function of the reactor coolant pressure boundary or its response during plant transients. By calculating the P-T limits using NRC-approved methodology, adequate margins of safety relating to reactor coolant pressure boundary integrity are maintained. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. These changes will ensure that protective actions are initiated and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

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

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

    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.

    NRC Branch Chief: Robert J. Pascarelli.

    Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2 (ANO-2), Pope County, Arkansas

    Date of amendment request: December 14, 2017. A publicly-available version is in ADAMS under Accession No. ML17348A150.

    Description of amendment request: The amendment would revise ANO-2 Technical Specification (TS) 3.3.3.6, “Post-Accident Instrumentation,” to ensure that both Category 1 and Type A Regulatory Guide (RG) 1.97, Revision 3, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,” instrumentation is included in the specification (unless already addressed within another specification) and gains greater consistency with NUREG-1432, Revision 4, “Standard Technical Specifications for Combustion Engineering Plants.”

    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 PAM [Post-Accident Monitoring] instrumentation is not an initiator of any design basis accident or event and, therefore, the proposed change does not increase the probability of any accident previously evaluated. The proposed change ensures required instrumentation is included in and controlled by the station TSs and does not change the response of the plant to any accidents.

    The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The removal and addition of specific instrumentation within ANO-2 TS 3.3.3.6 is consistent with the ANO-2 SAR [Safety Analysis Report], Table 7.5-3 RG 1.97 variables classified as Type A or Category 1 variables. Modifications to the TS Actions associated with inoperable instrumentation are consistent with the current ANO-2 licensing basis or act to improve consistency with NUREG 1432. The proposed change does not adversely affect the ability of structures, systems, and components (SSCs) to perform the associated intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed change does not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

    Instrumentation that does not meet the RG 1.97 inclusion criteria as established in NUREG-1432 are removed from the TS; however, the instrumentation remains applicable to other RG 1.97 criteria and is maintained accordingly. Instrumentation added to the ANO-2 PAM TS does not change the manner in which the instrumentation is currently maintained since these instruments are currently designated as Type A and/or Category 1 variables in the ANO-2 SAR. However, including these instruments within the TSs will now require different mitigating actions during periods of inoperability, which may include a plant shutdown, establishment of alternate monitoring methods, and/or submittal of a special report to the NRC.

    Therefore, this 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 does not result in a change in the manner in which the plant is operated during post-accident conditions and does not change the established mitigating actions associated with any necessary response to a DBA [design-basis accident]. The proposed change continues to ensure important instrumentation remains available to station operators such that currently established mitigating actions are not impacted. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal or post-accident plant operation. The change does not alter assumptions made in the safety analysis.

    Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.Start Printed Page 8516

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

    Response: No.

    The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria and assumptions are not impacted by the proposed change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change ensures appropriate PAM instrumentation is controlled by the station TSs and that specified remedial action will be taken when required instrumentation is inoperable. The proposed change continues to support the operator ability to monitor and control vital systems during post-accident conditions.

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

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

    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.

    NRC Branch Chief: Robert J. Pascarelli.

    Entergy Operations, Inc., System Energy Resources, Inc., Cooperative Energy, A Mississippi Electric Cooperative, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi

    Date of amendment request: November 3, 2017, as supplemented by letters dated December 6, 2017, and January 22, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML17307A440, ML17340B025, and ML18022A598, respectively.

    Description of amendment request: The amendment would revise the GGNS Updated Final Safety Analysis Report (UFSAR) to incorporate the Tornado Missile Risk Evaluator (TMRE) methodology contained in Nuclear Energy Institute (NEI) 17-02, Revision 1, “Tornado Missile Risk (TMRE) Industry Guidance Document,” September 2017 (ADAMS Accession No. ML17268A036). This methodology can only be applied to discovered conditions where tornado missile protection is not currently provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.

    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 with NRC:

    1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

    Response: No.

    The proposed amendment is to incorporate the TMRE methodology into the GGNS UFSAR. The TMRE methodology is an alternative methodology for determining whether protection from tornado-generated missiles is required. The methodology can only be applied to discovered conditions where tornado missile protection was not provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.

    The proposed amendment does not involve an increase in the probability of an accident previously evaluated. The relevant accident previously evaluated is a Design Basis Tornado impacting the GGNS site. The probability of a Design Basis Tornado is driven by external factors and is not affected by the proposed amendment. There are no changes required to any of the previously evaluated accidents in the UFSAR.

    The proposed amendment does not involve a significant increase in the consequences or a Design Basis Tornado. [The methodology as proposed does not alter any input assumptions or results of the accident analyses. Instead, it reflects a methodology to more realistically evaluate the probability of unacceptable consequences of a Design Basis Tornado. As such, there is no significant increase in the consequence of an accident previously evaluated. A similar consideration would apply in the event additional non-conforming conditions are discovered in the future.]

    2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

    Response: No.

    The proposed amendment is to incorporate the TMRE methodology into the GGNS UFSAR. The TMRE methodology is an alternative methodology for determining whether protection from tornado-generated missiles is required. The methodology can only be applied to discovered conditions where tornado missile protection was not provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.

    The proposed amendment will involve no physical changes to the existing plant, so no new malfunctions could create the possibility of a new or different kind of accident. The proposed amendment makes no changes to conditions external to the plant that could create the possibility of a new or different kind of accident. The proposed change will not create the possibility of a new or different kind of accident due to new accident precursors, failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. The existing Updated Final Safety Analysis Report accident analysis will continue to meet requirements for the scope and type of accidents that require analysis.

    Therefore, the proposed amendment will not create the possibility of a new or different kind of accident than those previously evaluated.

    3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

    Response: No.

    The proposed amendment is to incorporate the TMRE methodology into the GGNS UFSAR. The TMRE methodology is an alternative methodology for determining whether protection from tornado-generated missiles is required. The methodology can only be applied to discovered conditions where tornado missile protection was not provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.

    The change does not exceed or alter any controlling numerical value for a parameter established in the UFSAR or elsewhere in the GGNS licensing basis related to design basis or safety limits. The change does not impact any UFSAR Chapter 6 or 15 Safety Analyses, and those analyses remain valid. The change does not reduce diversity or redundancy as required by regulation or credited in the UFSAR. The change does not reduce defense-in-depth as described in the UFSAR.

    Therefore, the changes associated with this license amendment request do not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's modified 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: William B. Glew, Associate General Counsel, Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601.

    NRC Branch Chief: Douglas A. Broaddus.

    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: December 21, 2017. A publicly-available version is in ADAMS under Accession No. ML17355A184.

    Description of amendment request: The amendments would revise the Technical Specifications (TSs) pertaining to the Engineered Safety Features Actuation System instrumentation to resolve non-conservative actions associated with the containment ventilation isolation and the control room ventilation isolation functions. In addition, the amendments would revise the control room Start Printed Page 8517ventilation isolation function to no longer credit containment radiation monitoring instrumentation, eliminate redundant radiation monitoring instrumentation requirements, eliminate select core alterations applicability requirements, relocate radiation monitoring and reactor coolant system leakage detection requirements within the TSs to align with their respective functions, and relocate the spent fuel pool area monitoring requirements to licensee-controlled documents.

    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 instrumentation associated with the proposed changes to the technical specifications (TS) is not an initiator of any accidents previously evaluated, so the probability of accidents previously evaluated is unaffected by the proposed changes. There is no change to any equipment response or accident scenario, with the exception of the Control Room isolation on Containment high-radiation instrumentation function which impose no additional challenges to fission product barrier integrity. The exception is supported by revised radiological analyses which demonstrate that the Control Room air intake radioactivity monitoring instrumentation provides timely automatic isolation of the Control Room ventilation system and thereby limits Control Room operator doses to within regulatory limits for any design basis accident. The proposed changes also eliminate limitations imposed on Containment and Control Room ventilation instrumentation during CORE ALTERATIONS since the applicable postulated accidents do not result in fuel cladding integrity damage. Hence, the capability of any TS-required SSC [structure, system, or component] to perform its specified safety function is not impacted by the proposed changes and the outcomes of accidents previously evaluated are unaffected. Therefore, the proposed changes do not result in 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.

    No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, configuration, or method of operation of any plant SSC with the exception of the Control Room isolation on Containment high-radiation instrumentation function which is supported by revised accident analyses which demonstrate that the radiological consequences remain within applicable regulatory limits. The elimination of core alterations applicability requirements do not impact the outcome of any applicable postulated accident since none result in fuel cladding damage. No physical changes are made to the plant, so no new causal mechanisms are introduced. 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 amendment involve a significant reduction in a margin of safety?

    Response: No.

    The ability of any operable SSC to perform its designated safety function is unaffected by the proposed changes. The proposed change do not revise any safety limits or limiting safety system settings. The proposed changes revises safety analyses assumptions and the method of operating the plant with regard to the Control Room isolation on Containment high-radiation instrumentation function. The changes are supported by revised accident analyses which demonstrate that no adverse impact will result to either the plant operating margins or the reliability of equipment credited in the safety analyses. The existing margin in dose assessment currently afforded Control Room operators during any design basis accident is maintained. No other safety margins are impacted by the proposed changes. 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.

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

    NRC Branch Chief: Undine Shoop.

    NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

    Date of amendment request: November 10, 2017. A publicly-available version is in ADAMS under Accession No. ML17318A240.

    Description of amendment request: The proposed amendment revises Technical Specification (TS) 3.6.4.1, “Secondary Containment,” Surveillance Requirement (SR) 3.6.4.1.2. The SR is modified to acknowledge that secondary containment access openings may be open for entry and exit.

    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 addresses conditions during which the secondary containment SR is not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four-hour Completion Time for an inoperable secondary containment. As a result, the consequences of an accident previously evaluated are not significantly increased.

    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 previously evaluated?

    Response: No.

    The proposed change does not alter the protection system design, create new failure modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident.

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

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

    Response: No.

    The proposed change addresses conditions during which the secondary containment SR is not met. The allowance for both an inner and outer secondary containment door to be open simultaneously for entry and exit does not affect the safety function of the secondary containment as the doors are promptly closed after entry or exit, thereby restoring the secondary containment boundary.

    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: William Blair, P. O. Box 14000, Juno Beach, FL 33408-0420.

    NRC Branch Chief: David J. Wrona.Start Printed Page 8518

    Northern States Power Company—Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: December 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17353A189.

    Description of amendment request: The proposed amendment would adopt Technical Specifications Task Force (TSTF) traveler TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk-Informed Technical Specifications Task Force Initiative 5b,” Revision 3.

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

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

    Response: No.

    The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new SFCP [Surveillance Frequency Control Program]. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

    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 previously evaluated?

    Response: No.

    No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

    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 amendment involve a significant reduction in a margin of safety?

    Response: No.

    The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, [NSPM] will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

    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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

    NRC Branch Chief: David J. Wrona.

    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: November 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17334B211.

    Description of amendment request: The proposed changes include changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2* and Tier 1 information and related changes to the VEGP Units 3 and 4 Combined License (COL) Appendix C information. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from the elements of the design as certified in 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Tier 1 material departures. This submittal requests approval of the license amendment, necessary to implement these changes.

    Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), licensee has provided its analysis of the issue on no significant hazards consideration determination, 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 consistency and editorial changes to COL Appendix C (and associated plant-specific Tier 1) and Tier 2 and Tier 2* information in the UFSAR do not involve a technical change, (e.g. there is no design parameter or requirement, calculation, analysis, function or qualification change). No structure, requirement, calculation, analysis, function or qualification change). No structure, system, or component (SSC) design or function would be affected. No design or safety analysis would be affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not affected.

    Therefore, the proposed amendment 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 consistency and editorial changes to COL Appendix C (and associated plant specific Tier 1) and Tier 2 and Tier 2* information in the UFSAR do not change the design or functionality of safety-related SSCs. The proposed change does not affect plant electrical systems, and does not affect the design function, support, design, or operation of mechanical and fluid systems. The proposed change does not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is affected by the proposed changes.

    Therefore, the proposed amendment 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 consistency and editorial changes to COL Appendix C (and associated plant specific Tier 1) and Tier 2 and Tier 2* information in the UFSAR do not involve any change to the design as described in the COL. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit/criterion is involved.

    Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

    Start Printed Page 8519

    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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

    NRC Branch Chie f: Jennifer Dixon-Herrity.

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

    Date of amendment request: February 1, 2018. A publicly-available version is in ADAMS under Accession No. ML18032A359.

    Description of amendment request: The requested amendment proposes changes to relax the minimum gap requirement above grade between the nuclear island and the annex building/turbine building and removing the minimum gap requirement for the radwaste building from the Inspections, Tests, Analyses and Acceptance Criteria. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

    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 with NRC staff edits in square brackets:

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

    Response: No.

    The proposed changes are to relax the minimum gap requirement above grade between the nuclear island and the annex building/turbine building from a 4 inch gap to a 3 inch gap. The proposed changes modify and clarify the gap requirements between the nuclear island and the annex building/turbine building and radwaste building, respectively. The proposed change deletes the gap requirement for the radwaste building from the Inspections, Tests, Analyses and Acceptance Criteria (ITAAC) in (COL) [Combined License] Appendix C. The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any structure, system or component (SSC) such that a new accident initiator or initiating sequence of events is created.

    The changes do not impact the support, design, or operation of mechanical and fluid systems. The changes do not impact the support, design, or operation of any safety-related structures. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors.

    Therefore, the proposed amendment 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 changes are to relax the minimum gap requirement above grade between the nuclear island and the annex building/turbine building from a 4 inch gap to a 3 inch gap. The proposed changes modify and clarify the gap requirements between the nuclear island and the annex building/turbine building and radwaste building, respectively. The proposed changes delete the gap requirement for the radwaste building from the ITAAC in COL Appendix C. The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created.

    The proposed changes do not adversely affect the design function of the nuclear island and adjoining buildings' SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or non-safety-related equipment. This activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

    Therefore, the proposed amendment 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 changes maintain existing safety margin and provide adequate protection through continued application of the existing requirements in the UFSAR [Updated Final Safety Analysis Report]. The proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes.

    Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by these changes, no significant margin of safety is reduced.

    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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

    NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: January 31, 2018. A publicly-available version is in ADAMS under Accession No. ML18031B142.

    Description of amendment request: The requested amendment proposes to include changes to Combined License (COL) Appendix A, Technical Specifications related to fuel management. Specifically, the requested amendment proposes improvements to the technical specifications for the Rod Position Indication, the Control Rod Drive Mechanism, Power Range Neutron Flux Channels and the Mechanical Shim Augmentation.

    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 are to clarify proper operation and methodology associated with the DRPI [Digital Rod Position Indication], Control Rod Gripper Coils, instrumentation associated with Quadrant Power Tilt Ratio, or Control or Gray Rods. These changes do not affect the operation of this equipment and have no adverse impact on their design functions.

    The changes do not involve an interface with any structure, system, or component (SSC) accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the plant-specific Updated Final Safety Analysis Report (UFSAR) are not affected. The proposed changes do not adversely affect any mitigation sequence or the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected.Start Printed Page 8520

    Therefore, the proposed amendment 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 changes verify and maintain the capabilities of the DRPI, Control Rod Gripper Coils, instrumentation associated with Quadrant Power Tilt Ratio, and Control and Gray Rods to perform their design functions. The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created.

    The proposed changes do not affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures. These changes are to clarify proper operation and methodology associated with this equipment and have no adverse impact on their design functions.

    Therefore, the proposed amendment 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 changes do not affect existing safety margins. The proposed changes verify and maintain the capabilities of the DRPI, Control Rod Gripper Coils, instrumentation associated with Quadrant Power Tilt Ratio, and Control and Gray Rods to perform their design functions. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

    The proposed changes would not affect any safety-related design code, function, design analysis, safety analysis input or result, or existing design/safety margin. Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes, no margin of safety is significantly reduced.

    Therefore, the proposed amendment 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: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

    NRC Branch Chief: Jennifer Dixon-Herrity.

    Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: December 14, 2017. A publicly-available version is in ADAMS under Accession No. ML17348B097.

    Description of amendment request: The amendments would revise Technical Specification (TS) 3.6.4.1, “Secondary Containment,” Surveillance Requirement (SR) 3.6.4.1.1. The SR would be revised to address conditions during which the secondary containment pressure may not meet the SR pressure requirements. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, “Revise Secondary Containment Surveillance Requirements.” Also, the editorial note in SR 3.6.4.1.3 is removed because it is redundant to the SR itself and does not alter the requirement.

    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, along with NRC edits in square brackets:

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

    Response: No.

    The proposed change addresses conditions during which the secondary containment SR is not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four hour Completion Time for an inoperable secondary containment. In addition, the proposed Note for SR 3.6.4.1.1 provides an alternative means to ensure the secondary containment safety function is met. Additionally, the Note removed from SR 3.6.4.1.3 is editorial because it is redundant to the SR itself and does not alter the requirement. As a result, the consequences of an accident previously evaluated are not significantly increased.

    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 does not alter the protection system design, create new failure modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident.

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

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

    Response: No.

    The proposed change addresses conditions during which the secondary containment SR is not met. Conditions in which the secondary containment vacuum is less than the required vacuum are acceptable provided the conditions do not affect the ability of the SGT [Standby Gas Treatment] System to establish the required secondary containment vacuum under post-accident conditions within the time assumed in the accident analysis. This condition is incorporated in the proposed change by requiring an analysis of actual environmental and secondary containment pressure conditions to confirm the capability of the SGT System is maintained within the assumptions of the accident analysis. Therefore, the safety function of the secondary containment is not affected.

    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: Damon D. Obie, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101.

    NRC Branch Chief: James G. Danna.

    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: September 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17272A940.

    Description of amendment request: The amendments would make changes to the SQN Emergency Plan to extend staff augmentation times for Emergency Response Organization (ERO) functions.

    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:

    Start Printed Page 8521

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

    Response: No.

    The proposed removal of maintenance personnel from shift and extension in staff augmentation times has no effect on normal plant operation or on any accident initiator or precursor and does not affect the function of plant structures, systems, or components (SCCs). The proposed changes do not alter or prevent the ability of the ERO to perform their intended functions to mitigate the consequences of an accident or event. The ability of the ERO to respond adequately to radiological emergencies has been demonstrated as acceptable through a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9.

    Therefore, the proposed changes do 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 changes do not affect the accident analyses. The 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 changes do not alter assumptions made in the safety analysis. This proposed change removes maintenance personnel from shift and extends the staff augmentation response times in the SQN Emergency Plan, which are demonstrated as acceptable through a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9. The proposed changes do not alter or prevent the ability of the ERO to perform their intended functions to mitigate the consequences of an accident or event.

    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 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 change is associated with the SQN Emergency Plan staffing and does not affect operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses are affected by the proposed changes. Safety analysis acceptance criteria are not affected by this proposed change. A staffing analysis and a functional analysis were performed for the proposed changes on the timeliness of performing major tasks for the functional areas of the SQN Emergency Plan. The analysis concluded that removal of maintenance personnel from shift and an extension in staff augmentation times would not significantly affect the ability to perform the required Emergency Plan tasks.

    Therefore, the proposed changes are determined to not adversely affect the ability to meet 10 CFR 50.54(q)(2), the requirements of 10 CFR 50 Appendix E, and the emergency planning standards as described in 10 CFR 50.47(b).

    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 Branch Chief: Undine Shoop.

    III. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

    For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

    Florida Power & Light Company, Docket Nos. 50-250, Turkey Point Nuclear Generating Unit No. 3, Miami-Dade County, Florida

    Date of amendment request: December 18, 2017. A publicly-available version is in ADAMS under ML17353A492.

    Brief description of amendment request: Revise the Technical Specifications to allow a one-time extension of the allowable outage time for the Unit 3 Containment Spray System from 72 hours to 14 days.

    Date of publication of individual notice in Federal Register: January 30, 2018 (83 FR 4285).

    Expiration date of individual notice: March 1, 2018 (Public comments); April 2, 2018 (Hearing requests).

    IV. 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.Start Printed Page 8522

    Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake County, North Carolina

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

    Date of amendment request: July 18, 2017, as supplemented by letter dated October 12, 2017.

    Brief description of amendments: The amendments revised the technical specifications (TSs) based on Technical Specifications Task Force (TSTF) Traveler TSTF-529, “Clarify Use and Application Rules.” The changes revise and clarify the TS usage rules for completion times, limiting conditions for operation, and surveillance requirements.

    Date of issuance: February 1, 2018.

    Effective date: As of the date of issuance and shall be implemented within 120 days of issuance.

    Amendment Nos.: 298 and 294, for the Catawba Nuclear Station Units 1 and 2; 307 and 286, for the McGuire Nuclear Station, Units 1 and 2; 407, 409, and 408, for the Oconee Nuclear Station, Units 1, 2, and 3; 162, for the Shearon Harris Nuclear Power Plant, Unit 1; and 256, for the H. B. Robinson Steam Electric Plant, Unit No. 2. A publicly-available version is in ADAMS under Accession No. ML17340A720; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-17, DPR-38, DPR-47, DPR-55, NPF-63, and DPR-23: Amendments revised the Renewed Facility Operating Licenses and TSs.

    Date of initial notice in Federal Register: August 29, 2017 (82 FR 41067). The supplemental letter dated October 12, 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 February 1, 2018.

    No significant hazards consideration comments received: No.

    Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (McGuire), Mecklenburg County, North Carolina

    Date of amendment requests: December 19, 2016, as supplemented by letters dated May 25, 2017, and December 12, 2017.

    Brief description of amendments: The amendments modified Technical Specification 5.5.2, “Containment Leakage Rate Testing Program,” by replacing the reference to Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program,” with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” dated July 2012 and the conditions and limitations specified in NEI 94-01, Revisions 2-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” dated October 2008, as the implementation documents used by McGuire to implement the performance-based leakage testing program in accordance with Option B of 10 CFR part 50, Appendix J. The proposed change would also delete the listing of one-time exceptions previously granted to Integrated Leak Rate Test frequency.

    Date of issuance: January 31, 2018.

    Effective date: As of its date of issuance and shall be implemented within 120 days of issuance.

    Amendment Nos.: 306 (Unit 1) and 285 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18009A842; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

    Date of initial notice in Federal Register: May 9, 2017 (82 FR 21557). The supplemental letters dated May 25 and December 12, 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 a Safety Evaluation dated January 31, 2018.

    No significant hazards consideration comments received: No.

    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: August 29, 2017, as supplemented by letter dated January 25, 2018.

    Brief description of amendment: The amendments revised the LSCS technical specification (TS) 2.1.1, “Reactor Core SLs [Safety Limits].” Specifically, this change incorporates revised LSCS, Units 1 and 2, safety limits for minimum critical power ratio for two circulation loop minimum critical power ratio (MCPR) and single circulation loop MCPR values for Unit 1 and Unit 2 based on the results of the cycle-specific analyses performed by Global Nuclear Fuel (GNF) for LSCS Unit 1, Cycle 17, and LSCS Unit 2, Cycle 17.

    Date of issuance: February 6, 2018.

    Effective date: As of the date of issuance and shall be implemented as follows:

    Unit 1: Prior to startup from the February 2018 refueling outage for Unit 1 (i.e., L1R17) for operation starting in Cycle 18.

    Unit 2: Prior to startup from the February 2018 refueling outage for Unit 1 (i.e., L1R17). This will be a mid-Cycle 17 implementation for Unit 2.

    Amendment No.: Unit 1-227; Unit 2-213. A publicly-available version is in ADAMS under Accession No. ML18008A123; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Renewed Facility Operating Licenses and TSs.

    Date of initial notice in Federal Register: December 5, 2017 (82 FR 57482). The supplemental letter dated January 25, 2018, 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 a Safety Evaluation dated February 6, 2018.Start Printed Page 8523

    No significant hazards consideration comments received: No.

    Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of amendment request: May 24, 2017, as supplemented by letter dated August 17, 2017.

    Brief description of amendments: The amendments revised Surveillance Requirement 3.3.1.3 to change the thermal power at which the surveillance may be performed.

    Date of issuance: February 7, 2018.

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

    Amendment Nos.: 194 (Unit 1) and 177 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18012A068; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License No. NPF-68 and NPF-81: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

    Date of initial notice in Federal Register: July 18, 2017 (82 FR 32883). The supplemental letter dated August 17, 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 February 7, 2018.

    No significant hazards consideration comments received: No.

    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment requests: March 13, 2017, as supplemented by letter dated August 7, 2017.

    Brief description of amendments: The amendments deleted the Note associated with Technical Specification (TS) Surveillance Requirement (SR) 3.8.1.17 to allow the performance of the SR in Modes 1 through 4.

    Date of issuance: February 2, 2018.

    Effective date: As of its date of issuance and shall be implemented no later than 60 days from the date of issuance.

    Amendment Nos.: 340 (Unit 1) and 333 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17296A133; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments.

    Facility Operating License Nos. DPR-77 and DPR-79: The amendments revised the Renewed Facility Operating Licenses and TSs.

    Date of initial notice in Federal Register: July 5, 2017 (82 FR 31102). The supplemental letter dated August 7, 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 February 2, 2018.

    No significant hazards consideration comments received: No.

    Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: January 20, 2017, as supplemented by letter dated September 7, 2017.

    Brief description of amendments: The amendments revised the Technical Specification (TS) 3.5, “Residual Heat Removal (RHR) System,” requirements, as well as the TS 3.13, “Component Cooling System,” RHR support requirements for consistency with the design basis of the RHR system. In addition, an RHR surveillance requirement is added in TS Table 4.1-2A, “Minimum Frequency for Equipment Tests,” to test the RHR system in accordance with the inservice testing program, since a TS surveillance does not currently exist for this system.

    Date of issuance: February 9, 2018.

    Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.

    Amendment Nos.: 291 and 291. A publicly-available version is in ADAMS under Accession No. ML17326A225; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

    Renewed Facility Operating License No. DPR-32 and DPR-37: Amendments revised the Renewed Facility Operating Licenses and TSs.

    Date of initial notice in Federal Register : March 14, 2017 (82 FR 13672). The supplemental letter dated September 7, 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 February 9, 2018.

    No significant hazards consideration comments received: No.

    Start Signature

    Dated at Rockville, Maryland, on February 20, 2018.

    For the Nuclear Regulatory Commission.

    Kathryn M. Brock,

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

    End Signature End Supplemental Information

    [FR Doc. 2018-03727 Filed 2-26-18; 8:45 am]

    BILLING CODE 7590-01-P

Document Information

Published:
02/27/2018
Department:
Nuclear Regulatory Commission
EntryType:
Notice
Action:
Biweekly notice.
Document Number:
2018-03727
Dates:
Comments must be filed by March 29, 2018. A request for a hearing must be filed by April 30, 2018.
Pages:
8509-8523 (15 pages)
Docket Numbers:
NRC-2018-0031
SectionNoes:
PDF File:
2018-03727.pdf