83 FR 8509 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 83, Issue 39 (February 27, 2018)
Federal Register Volume 83, Issue 39 (February 27, 2018)
| Page Range | 8509-8523 | |
| FR Document | 2018-03727 |
[Federal Register Volume 83, Number 39 (Tuesday, February 27, 2018)] [Notices] [Pages 8509-8523] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2018-03727] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2018-0031] Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations 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: [email protected]. 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. 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: [email protected]. 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 [email protected]. 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 Sec. 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 [[Page 8510]] issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued. As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest. In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures. Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from 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 [[Page 8511]] accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [email protected], 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 [email protected], 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 ([deg]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. [[Page 8512]] Technical Specifications currently allow for operation at greater than 200[emsp14][deg]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[emsp14][deg]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[emsp14][deg]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 [[Page 8513]] 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 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. [[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 [[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. [[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 [[Page 8517]] ventilation 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. [[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. [[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 Chief: 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. [[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: [[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. [[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. [[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. 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. [FR Doc. 2018-03727 Filed 2-26-18; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8509
Dated: February 21, 2018. Commission, Washington, DC 20555– The NRC does not routinely edit
Sherry P. Hale, 0001. comment submissions to remove
Staff Assistant, National Endowment for the For additional direction on obtaining identifying or contact information.
Arts. information and submitting comments, If you are requesting or aggregating
[FR Doc. 2018–03913 Filed 2–26–18; 8:45 am] see ‘‘Obtaining Information and comments from other persons for
BILLING CODE 7537–01–P Submitting Comments’’ in the submission to the NRC, then you should
SUPPLEMENTARY INFORMATION section of inform those persons not to include
this document. identifying or contact information that
FOR FURTHER INFORMATION CONTACT: they do not want to be publicly
NUCLEAR REGULATORY Paula Blechman, Office of Nuclear disclosed in their comment submission.
COMMISSION Reactor Regulation, U.S. Nuclear Your request should state that the NRC
Regulatory Commission, Washington, does not routinely edit comment
[NRC–2018–0031] submissions to remove such information
DC 20555–0001; telephone: 301–415–
2242, email: Paula.Blechman@nrc.gov. before making the comment
Biweekly Notice; Applications and
SUPPLEMENTARY INFORMATION:
submissions available to the public or
Amendments to Facility Operating
entering the comment into ADAMS.
Licenses and Combined Licenses
I. Obtaining Information and
Involving No Significant Hazards II. Notice of Consideration of Issuance
Submitting Comments
Considerations of Amendments to Facility Operating
A. Obtaining Information Licenses and Combined Licenses and
AGENCY: Nuclear Regulatory Proposed No Significant Hazards
Commission. Please refer to Docket ID NRC–2018–
0031, facility name, unit number(s), Consideration Determination
ACTION: Biweekly notice.
plant docket number, application date, The Commission has made a
SUMMARY: Pursuant to Section 189a.(2) and subject when contacting the NRC proposed determination that the
of the Atomic Energy Act of 1954, as about the availability of information for following amendment requests involve
amended (the Act), the U.S. Nuclear this action. You may obtain publicly- no significant hazards consideration.
Regulatory Commission (NRC) is available information related to this Under the Commission’s regulations in
publishing this regular biweekly notice. action by any of the following methods: § 50.92 of title 10 of the Code of Federal
The Act requires the Commission to • Federal Rulemaking website: Go to Regulations (10 CFR), this means that
publish notice of any amendments http://www.regulations.gov and search operation of the facility in accordance
issued, or proposed to be issued, and for Docket ID NRC–2018–0031. with the proposed amendment would
grants the Commission the authority to • NRC’s Agencywide Documents not (1) involve a significant increase in
issue and make immediately effective Access and Management System the probability or consequences of an
any amendment to an operating license (ADAMS): You may obtain publicly- accident previously evaluated, or (2)
or combined license, as applicable, available documents online in the create the possibility of a new or
upon a determination by the ADAMS Public Documents collection at different kind of accident from any
Commission that such amendment http://www.nrc.gov/reading-rm/ accident previously evaluated; or (3)
involves no significant hazards adams.html. To begin the search, select involve a significant reduction in a
consideration, notwithstanding the ‘‘ADAMS Public Documents’’ and then margin of safety. The basis for this
pendency before the Commission of a select ‘‘Begin Web-based ADAMS proposed determination for each
request for a hearing from any person. Search.’’ For problems with ADAMS, amendment request is shown below.
please contact the NRC’s Public The Commission is seeking public
This biweekly notice includes all
Document Room (PDR) reference staff at comments on this proposed
notices of amendments issued, or
1–800–397–4209, 301–415–4737, or by determination. Any comments received
proposed to be issued, from January 30,
email to pdr.resource@nrc.gov. The within 30 days after the date of
2018, to February 12, 2018. The last
ADAMS accession number for each publication of this notice will be
biweekly notice was published on
document referenced (if it is available in considered in making any final
February 13, 2018.
DAMS) is provided the first time that it determination.
DATES: Comments must be filed by Normally, the Commission will not
is mentioned in this document.
March 29, 2018. A request for a hearing • NRC’s PDR: You may examine and issue the amendment until the
must be filed by April 30, 2018. purchase copies of public documents at expiration of 60 days after the date of
ADDRESSES: You may submit comments the NRC’s PDR, Room O1–F21, One publication of this notice. The
by any of the following methods (unless White Flint North, 11555 Rockville Commission may issue the license
this document describes a different Pike, Rockville, Maryland 20852. amendment before expiration of the 60-
method for submitting comments on a day period provided that its final
specific subject): B. Submitting Comments determination is that the amendment
• Federal Rulemaking website: Go to Please include Docket ID NRC–2018– involves no significant hazards
http://www.regulations.gov and search 0031, facility name, unit number(s), consideration. In addition, the
for Docket ID NRC–2018–0031. Address plant docket number, application date, Commission may issue the amendment
questions about NRC dockets to Jennifer and subject in your comment prior to the expiration of the 30-day
Borges; telephone: 301–287–9127; submission. comment period if circumstances
email: Jennifer.Borges@nrc.gov. For The NRC cautions you not to include change during the 30-day comment
daltland on DSKBBV9HB2PROD with NOTICES
technical questions, contact the identifying or contact information that period such that failure to act in a
individual listed in the FOR FURTHER you do not want to be publicly timely way would result, for example in
INFORMATION CONTACT section of this disclosed in your comment submission. derating or shutdown of the facility. If
document. The NRC will post all comment the Commission takes action prior to the
• Mail comments to: May Ma, Office submissions at http:// expiration of either the comment period
of Administration, Mail Stop: TWFN–7– www.regulations.gov as well as enter the or the notice period, it will publish in
A60M, U.S. Nuclear Regulatory comment submissions into ADAMS. the Federal Register a notice of
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Technical Specifications currently allow Duke Energy Carolinas, LLC, Docket accidents since the remaining DGs
for operation at greater than 200 °F while Nos. 50–413 and 50–414, Catawba supporting the redundant engineered safety
imposing MODE 4 requirements in addition Nuclear Station, Units 1 and 2 (CNS), feature systems would continue to be
to the secondary containment requirements available to perform the accident mitigation
York County, South Carolina functions. Thus, allowing a DG to be
required to be met. Extending the activities
that can apply this allowance will not Date of amendment request: May 2, inoperable for an additional 11 days for
2017, as supplemented by letters dated performance of maintenance or testing does
adversely impact the probability or
July 20 and November 21, 2017. not increase the probability of a previously
consequences of an accident previously evaluated accident.
evaluated. Therefore, the proposed change Publicly-available versions are in Deterministic and probabilistic risk
does not involve a significant increase in the ADAMS under Accession Nos. assessment techniques evaluated the effect of
probability or consequences of an accident ML17122A116, ML17201Q132, and the proposed TS change to extend the CT for
previously evaluated. ML17325A588, respectively. an inoperable DG on the availability of an
Criterion 2: The proposed change does not Description of amendment request: electrical power supply to the plant
create the possibility of a new or different The amendments would modify CNS emergency safeguards feature systems. These
kind of accident from any accident Technical Specifications (TSs) to extend assessments concluded that the proposed
previously evaluated. CNS [. . .] TS change does not involve a
the Completion Time (CT) of TS 3.8.1,
Technical Specifications currently allow significant increase in the risk of power
‘‘AC [Alternating Current] Sources— supply unavailability.
for operation at greater than 200 °F while Operating,’’ Required Action B.6
imposing MODE 4 requirements in addition There is a small incremental risk
(existing Required Action B.4, associated with continued operation for an
to the secondary containment requirements
numbered as B.6) for an inoperable additional 11 days with one DG inoperable;
required to be met. No new operational however, the calculated impact provides risk
conditions beyond those currently allowed
emergency diesel generator (DG) from
72 hours to 14 days. A conforming metrics consistent with the acceptance
by LCO 3.10.1 are introduced. The changes guidelines contained in Regulatory Guides
do not involve a physical alteration of the change is also proposed to extend the
1.177 and 1.174. The remaining operable DGs
plant (i.e., no new or different type of maximum CT of TS 3.8.1 Required and paths are adequate to supply electrical
equipment will be installed) or a change in Actions A.3 and B.4. To support this power to the onsite emergency power
the methods governing normal plant request, the licensee will add a distribution system. A DG is required to
operation. In addition, the changes do not supplemental power source (i.e., two operate only if both offsite power sources fail
impose any new or different requirements or supplemental diesel generators (SDGs) and there is an event which requires
eliminate any existing requirements. The per station) with the capability to power operation of the plant engineered safety
changes do not alter assumptions made in the any emergency bus. The SDGs will have features such as a design basis accident. The
probability of a design basis accident
safety analysis. The proposed changes are the capacity to bring the affected unit to occurring during this period is low.
consistent with the safety analysis cold shutdown. Additionally, the The consequences of previously evaluated
assumptions and current plant operating amendments would modify TS 3.8.1 to accidents will remain the same during the
practice. Therefore, the proposed change add new two limiting conditions for proposed 14 day CT as during the current
does not create the possibility of a new or operation (LCOs), TS LCO 3.8.1.c and CNS [. . .] 72 hour CT. The ability of the
different kind of accident from any accident TS LCO 3.8.1.d, to ensure that at least remaining TS required DGs to mitigate the
previously evaluated. one train of shared components has an consequences of an accident will not be
Criterion 3: The proposed change does not affected since no additional failures are
operable emergency power supply.
involve a significant reduction in a margin of postulated while equipment is inoperable
safety.
Basis for proposed no significant within the TS CT.
Technical Specifications currently allow hazards consideration determination: Regarding the proposed change to add
for operation at greater than 200 °F while As required by 10 CFR 50.91(a), the Required Action to ensure that at least one
imposing MODE 4 requirements in addition licensee has provided its analysis of the train of shared components has an operable
to the secondary containment requirements issue of no significant hazards emergency power supply, there is no change
required to be met. Extending the activities consideration, which is presented to how or under what conditions offsite
that can apply this allowance will not below: circuits or DGs are operated nor are there any
adversely impact any margin of safety. changes to acceptable operating parameters.
1. Does the proposed amendment involve Power source operability requirements for
Allowing completion of inspections and a significant increase in the probability or shared components are being moved from the
testing and supporting completion of scram consequences of an accident previously TS Bases to TS with the proposed change.
time testing initiated in conjunction with an evaluated? The proposed change will ensure that at least
inservice leak or hydrostatic test prior to Response: No. one train of shared components has an
power operation results in enhanced safe The proposed change involves extending operable emergency power supply whenever
operations by eliminating unnecessary the TS CT for an inoperable DG at CNS a DG is inoperable.
maneuvers to control reactor temperature and [. . .]. The proposed change also involves Therefore, the proposed change does not
pressure. Therefore, the proposed change adding a new Required Action to TSs to involve a significant increase in the
does not involve a significant reduction in a ensure that at least one train of shared probability or consequences of an accident
margin of safety. components at CNS [. . .] has an operable previously evaluated.
emergency power supply whenever one DG 2. Does the proposed amendment create
The NRC staff has reviewed the above is inoperable. The DGs at both stations are the possibility of a new or different kind of
analysis and, based on this review, it safety related components which provide a accident from any accident previously
appears that the three standards of 10 backup electrical power supply to the onsite evaluated?
CFR 50.92(c) are satisfied. Therefore, the emergency power distribution system. The Response: No.
proposed change does not affect the design The proposed change involves extending
NRC staff proposes to determine that the of the DGs, the operational characteristics or the TS CT for an inoperable DG at CNS
amendment request involves no
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function of the DGs, the interfaces between [. . .]. The proposed change also involves
significant hazards consideration. the DGs and other plant systems or the adding a new Required Action to TSs to
Attorney for licensee: Jon P. reliability of the DGs. The DGs are not ensure that at least one train of shared
Christinidis, DTE Energy, Expert accident initiators; the DGs are designed to components at CNS [. . .] has an operable
Attorney—Regulatory, 688 WCB, One mitigate the consequences of previously emergency power supply whenever one DG
evaluated accidents including a loss of offsite is inoperable.
Energy Plaza, Detroit, MI 48226–1279. power. Extending the CT for a single DG The proposed change does not involve a
NRC Branch Chief: David J. Wrona. would not affect the previously evaluated change in the CNS [. . .] plant design, plant
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![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8513
configuration, system operation or scenarios is not changed by this proposed (existing Required Action B.4,
procedures involved with the DGs. The amendment; there is no change to the DG numbered as B.6) for an inoperable
proposed change allows a DG to be operating parameters. In the extended CT, as emergency diesel generator (DG) from
inoperable for additional time. Equipment in the existing CT, the remaining operable
72 hours to 14 days. A conforming
will be operated in the same configuration DGs and paths are adequate to supply
and manner that is currently allowed and electrical power to the onsite emergency change is also proposed to extend the
designed for. The functional demands on power distribution system. The proposed maximum CT of TS 3.8.1 Required
credited equipment is unchanged. There are change to extend the CT for an inoperable DG Actions A.3 and B.4. To support this
no new failure modes or mechanisms created does not alter a design basis safety limit; request, the licensee will add a
due to plant operation for an extended period therefore, it does not significantly reduce the supplemental power source (i.e., two
to perform DG maintenance or testing. margin of safety. The DGs will continue to supplemental diesel generators (SDGs)
Extended operation with an inoperable DG operate per the existing design and regulatory per station) with the capability to power
does not involve any modification to the requirements. any emergency bus. The SDGs will have
operational limits or physical design of plant The proposed TS changes (i.e., the
systems. There are no new accident the capacity to bring the affected unit to
inoperable DG CT extension request and
precursors generated due to the extended CT. proposed change to add Required Action to cold shutdown. Additionally, the
Regarding the proposed change to add ensure that at least one train of shared amendments would modify TS 3.8.1 to
Required Action to ensure that at least one components has an operable emergency add new two limiting conditions for
train of shared components has an operable power supply) do not alter the plant design operation (LCOs), TS LCO 3.8.1.c and
emergency power supply, there is no change nor do they change the assumptions TS LCO 3.8.1.d, to ensure that at least
to how or under what conditions offsite contained in the safety analyses. The standby one train of shared components has an
circuits or DGs are operated nor are there any AC power system is designed with sufficient operable emergency power supply.
changes to acceptable operating parameters. redundancy such that a DG may be removed Basis for proposed no significant
Power source operability requirements for from service for maintenance or testing. The
shared components are being moved from the hazards consideration determination:
remaining DGs are capable of carrying
TS Bases to TS with the proposed change. As required by 10 CFR 50.91(a), the
sufficient electrical loads to satisfy the
The proposed change will ensure that at least Updated Final Safety Analysis Report licensee has provided its analysis of the
one train of shared components has an requirements for accident mitigation or unit issue of no significant hazards
operable emergency power supply whenever safe shutdown. The proposed change does consideration, which is presented
a DG is inoperable. This change does not not impact the redundancy or availability below:
alter the nature of events postulated in the requirements of offsite power circuits or
Updated Final Safety Analysis Report nor 1. Does the proposed amendment involve
change the ability of the plant to cope with a significant increase in the probability or
does it introduce any unique precursor a station blackout. Therefore, based on the consequences of an accident previously
mechanisms. considerations given above, the proposed evaluated?
Therefore, the proposed change does not changes do not involve a significant Response: No.
create the possibility of a new or different reduction in the margin of safety. The proposed change involves extending
kind of accident from any accident Therefore, the proposed change does not the TS CT for an inoperable DG at [. . .]
previously evaluated. involve a significant reduction in a margin of MNS. The proposed change also involves
3. Does the proposed amendment involve safety. adding a new Required Action to TSs to
a significant reduction in the margin of ensure that at least one train of shared
safety? The NRC staff has reviewed the
components at [. . .] MNS has an operable
Response: No. licensee’s analysis and, based on this
emergency power supply whenever one DG
The proposed change involves extending review, it appears that the three is inoperable. The DGs at both stations are
the TS CT for an inoperable DG at CNS standards of 10 CFR 50.92(c) are safety related components which provide a
[. . .]. The proposed change also involves satisfied. Therefore, the NRC staff backup electrical power supply to the onsite
adding a new Required Action to TSs to proposes to determine that the emergency power distribution system. The
ensure that at least one train of shared amendment request involves no proposed change does not affect the design
components at CNS [. . .] has an operable of the DGs, the operational characteristics or
emergency power supply whenever one DG
significant hazards consideration.
Attorney for licensee: Kate B. Nolan, function of the DGs, the interfaces between
is inoperable. the DGs and other plant systems or the
Currently, if an inoperable DG is not Deputy General Counsel, Duke Energy
reliability of the DGs. The DGs are not
restored to operable status within 72 hours at Carolinas, LLC, 550 South Tryon accident initiators; the DGs are designed to
CNS [. . .], TS 3.8.1, requires the units to be Street—DEC45A, Charlotte, NC 28202– mitigate the consequences of previously
in Mode 3 (i.e., Hot Standby) within a CT of 1802. evaluated accidents including a loss of offsite
6 hours, and to be in Mode 5 (i.e., Cold NRC Branch Chief: Michael T. power. Extending the CT for a single DG
Shutdown) within a CT of 36 hours. The Markley. would not affect the previously evaluated
proposed TS changes will allow steady state accidents since the remaining DGs
plant operation at 100 percent power for an Duke Energy Carolinas, LLC, Docket supporting the redundant engineered safety
additional 11 days for performance of DG Nos. 50–369 and 50–370, McGuire feature systems would continue to be
planned reliability improvements and Nuclear Station, Units 1 and 2 (MNS), available to perform the accident mitigation
preventive and corrective maintenance. Mecklenburg County, North Carolina functions. Thus, allowing a DG to be
Deterministic and probabilistic risk inoperable for an additional 11 days for
assessment techniques evaluated the effect of Date of amendment request: May 2, performance of maintenance or testing does
the proposed TS change to extend the CT for 2017, as supplemented by letters dated not increase the probability of a previously
an inoperable DG on the availability of an July 20 and November 21, 2017. evaluated accident.
electrical power supply to the plant Publicly-available versions are in Deterministic and probabilistic risk
emergency safeguards feature systems. These ADAMS under Accession Nos. assessment techniques evaluated the effect of
assessments concluded that the proposed ML17122A116, ML17201Q132, and the proposed TS change to extend the
daltland on DSKBBV9HB2PROD with NOTICES
CNS [. . .] TS change does not involve a ML17325A588, respectively. [completion time] CT for an inoperable DG
significant increase in the risk of power on the availability of an electrical power
Description of amendment request:
supply unavailability. supply to the plant emergency safeguards
The DGs continue to meet their design The amendments would modify MNS feature systems. These assessments
requirements; there is no reduction in Technical Specifications (TSs) to extend concluded that the proposed [. . .] MNS TS
capability or change in design configuration. the Completion Time (CT) of TS 3.8.1, change does not involve a significant
The DG response to loss of offsite power, loss ‘‘AC [Alternating Current] Sources— increase in the risk of power supply
of coolant accident, station blackout or fire Operating,’’ Required Action B.6 unavailability.
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![8514 Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices
There is a small incremental risk train of shared components has an operable power supply) do not alter the plant design
associated with continued operation for an emergency power supply, there is no change nor do they change the assumptions
additional 11 days with one DG inoperable; to how or under what conditions offsite contained in the safety analyses. The standby
however, the calculated impact provides risk circuits or DGs are operated nor are there any AC power system is designed with sufficient
metrics consistent with the acceptance changes to acceptable operating parameters. redundancy such that a DG may be removed
guidelines contained in Regulatory Guides Power source operability requirements for from service for maintenance or testing. The
1.177 and 1.174. shared components are being moved from the remaining DGs are capable of carrying
The remaining operable DGs and paths are TS Bases to TS with the proposed change. sufficient electrical loads to satisfy the
adequate to supply electrical power to the The proposed change will ensure that at least Updated Final Safety Analysis Report
onsite emergency power distribution system. one train of shared components has an requirements for accident mitigation or unit
A DG is required to operate only if both operable emergency power supply whenever safe shutdown. The proposed change does
offsite power sources fail and there is an a DG is inoperable. This change does not not impact the redundancy or availability
event which requires operation of the plant alter the nature of events postulated in the requirements of offsite power circuits or
engineered safety features such as a design Updated Final Safety Analysis Report nor change the ability of the plant to cope with
basis accident. The probability of a design does it introduce any unique precursor a station blackout. Therefore, based on the
basis accident occurring during this period is mechanisms. considerations given above, the proposed
low. Therefore, the proposed change does not changes do not involve a significant
The consequences of previously evaluated create the possibility of a new or different reduction in the margin of safety.
accidents will remain the same during the kind of accident from any accident Therefore, the proposed change does not
proposed 14 day CT as during the current previously evaluated. involve a significant reduction in a margin of
[. . .] MNS 72 hour CT. The ability of the 3. Does the proposed amendment involve safety.
remaining TS required DGs to mitigate the a significant reduction in the margin of
consequences of an accident will not be safety? The NRC staff has reviewed the
affected since no additional failures are Response: No. licensee’s analysis and, based on this
postulated while equipment is inoperable The proposed change involves extending review, it appears that the three
within the TS CT. the TS CT for an inoperable DG at [. . .] standards of 10 CFR 50.92(c) are
Regarding the proposed change to add MNS. The proposed change also involves satisfied. Therefore, the NRC staff
Required Action to ensure that at least one adding a new Required Action to TSs to
train of shared components has an operable proposes to determine that the
ensure that at least one train of shared amendment request involves no
emergency power supply, there is no change components at [. . .] MNS has an operable
to how or under what conditions offsite significant hazards consideration.
emergency power supply whenever one DG
circuits or DGs are operated nor are there any Attorney for licensee: Kate B. Nolan,
is inoperable.
changes to acceptable operating parameters.
Currently, if an inoperable DG is not Deputy General Counsel, Duke Energy
Power source operability requirements for Carolinas, LLC, 550 South Tryon
restored to operable status within 72 hours at
shared components are being moved from the Street—DEC45A Charlotte, NC 28202–
[. . .] MNS, TS 3.8.1, requires the units to be
TS Bases to TS with the proposed change.
in Mode 3 (i.e., Hot Standby) within a CT of 1802.
The proposed change will ensure that at least
6 hours, and to be in Mode 5 (i.e., Cold NRC Branch Chief: Michael T.
one train of shared components has an
operable emergency power supply whenever
Shutdown) within a CT of 36 hours. The Markley.
a DG is inoperable. proposed TS changes will allow steady state
plant operation at 100 percent power for an Entergy Operations, Inc., Docket No. 50–
Therefore, the proposed change does not 368, Arkansas Nuclear One, Unit 2
involve a significant increase in the additional 11 days for performance of DG
probability or consequences of an accident planned reliability improvements and (ANO–2), Pope County, Arkansas
previously evaluated. preventive and corrective maintenance. Date of amendment request:
2. Does the proposed amendment create Deterministic and probabilistic risk
assessment techniques evaluated the effect of
November 20, 2017. A publicly-
the possibility of a new or different kind of available version is in ADAMS under
accident from any accident previously the proposed TS change to extend the CT for
an inoperable DG on the availability of an Accession No. ML17326A387.
evaluated? Description of amendment request:
Response: No. electrical power supply to the plant
The proposed change involves extending emergency safeguards feature systems. These The amendment would revise the
the TS CT for an inoperable DG at [. . .] assessments concluded that the proposed Technical Specifications (TSs) to
MNS. The proposed change also involves [. . .] MNS TS change does not involve a replace the current pressure-
adding a new Required Action to TSs to significant increase in the risk of power temperature limits for heatup,
ensure that at least one train of shared supply unavailability. cooldown, and the inservice leak
components at [. . .] MNS has an operable The DGs continue to meet their design
hydrostatic tests for the reactor coolant
emergency power supply whenever one DG requirements; there is no reduction in
capability or change in design configuration. system presented in TS 3.4.9 that expire
is inoperable. at 32 Effective Full Power Years (EFPY)
The proposed change does not involve a The DG response to loss of offsite power, loss
change in the [. . .] MNS plant design, plant of coolant accident, station blackout or fire with limitations that extend out to 54
configuration, system operation or scenarios is not changed by this proposed EFPY.
procedures involved with the DGs. The amendment; there is no change to the DG Basis for proposed no significant
proposed change allows a DG to be operating parameters. In the extended CT, as hazards consideration determination:
inoperable for additional time. Equipment in the existing CT, the remaining operable As required by 10 CFR 50.91(a), the
will be operated in the same configuration DGs and paths are adequate to supply licensee has provided its analysis of the
and manner that is currently allowed and electrical power to the onsite emergency issue of no significant hazards
designed for. The functional demands on power distribution system. The proposed
consideration, which is presented
credited equipment is unchanged. There are change to extend the CT for an inoperable DG
does not alter a design basis safety limit; below:
no new failure modes or mechanisms created
due to plant operation for an extended period therefore, it does not significantly reduce the 1. Does the proposed change involve a
daltland on DSKBBV9HB2PROD with NOTICES
to perform DG maintenance or testing. margin of safety. The DGs will continue to significant increase in the probability or
Extended operation with an inoperable DG operate per the existing design and regulatory consequences of an accident previously
does not involve any modification to the requirements. evaluated?
operational limits or physical design of plant The proposed TS changes (i.e., the Response: No.
systems. There are no new accident inoperable DG CT extension request and The proposed change will revise the
precursors generated due to the extended CT. proposed change to add Required Action to pressure-temperature (P–T) limits for heatup,
Regarding the proposed change to add ensure that at least one train of shared cooldown, and inservice leak hydrostatic test
Required Action to ensure that at least one components has an operable emergency limitations for the Reactor Coolant System
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![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8515
(RCS) to a maximum of 54 Effective Full ensure that protective actions are initiated alter the design assumptions, conditions, and
Power Years (EFPY) in accordance with 10 and the operability requirements for configuration of the facility or the manner in
CFR 50, Appendix G. This is the end of the equipment assumed to operate for accident which the plant is operated and maintained.
period of extended operation for the renewed mitigation are not affected. The removal and addition of specific
ANO–2 operating License. The P–T limits Therefore, this change does not involve a instrumentation within ANO–2 TS 3.3.3.6 is
were developed in accordance with the significant reduction in a margin of safety. consistent with the ANO–2 SAR [Safety
requirements of 10 CFR 50, Appendix G, Analysis Report], Table 7.5–3 RG 1.97
utilizing the analytical methods and flaw The NRC staff has reviewed the variables classified as Type A or Category 1
acceptance criteria of Topical Report WCAP– licensee’s analysis and, based on this variables. Modifications to the TS Actions
14040, Revision 4, and American Society of review, it appears that the three associated with inoperable instrumentation
Mechanical Engineers (ASME) Code, Section standards of 10 CFR 50.92(c) are are consistent with the current ANO–2
XI, Appendix G. These methods and criteria satisfied. Therefore, the NRC staff licensing basis or act to improve consistency
are the previously NRC approved standards proposes to determine that the with NUREG 1432. The proposed change
for the preparation of P–T limits. Updating does not adversely affect the ability of
amendment request involves no
the P–T limits for additional EFPYs structures, systems, and components (SSCs)
significant hazards consideration. to perform the associated intended safety
maintains the level of assurance that reactor
coolant pressure boundary integrity will be
Attorney for licensee: Ms. Anna function to mitigate the consequences of an
maintained, as specified in 10 CFR 50, Vinson Jones, Senior Counsel, Entergy initiating event within the assumed
Appendix G. Services, Inc., 101 Constitution Avenue acceptance limits. The proposed change does
The proposed changes do not adversely NW, Suite 200 East, Washington, DC not affect the source term, containment
affect accident initiators or precursors, and 20001. isolation, or radiological release assumptions
do not alter the design assumptions, NRC Branch Chief: Robert J. used in evaluating the radiological
conditions, or configuration of the plant or consequences of any accident previously
Pascarelli.
the manner in which the plant is operated evaluated. Further, the proposed change does
and maintained. The ability of structures, Entergy Operations, Inc., Docket No. 50– not increase the types and amounts of
systems, and components to perform their 368, Arkansas Nuclear One, Unit 2 radioactive effluent that may be released
intended safety functions is not altered or offsite, nor significantly increase individual
(ANO–2), Pope County, Arkansas
prevented by the proposed changes, and the or cumulative occupational/public radiation
assumptions used in determining the Date of amendment request: exposures.
radiological consequences of previously December 14, 2017. A publicly-available Instrumentation that does not meet the RG
evaluated accidents are not affected. version is in ADAMS under Accession 1.97 inclusion criteria as established in
No. ML17348A150. NUREG–1432 are removed from the TS;
Therefore, this change does not involve a
however, the instrumentation remains
significant increase in the probability or Description of amendment request: applicable to other RG 1.97 criteria and is
consequences of an accident previously The amendment would revise ANO–2 maintained accordingly. Instrumentation
evaluated. Technical Specification (TS) 3.3.3.6, added to the ANO–2 PAM TS does not
2. Does the proposed change create the ‘‘Post-Accident Instrumentation,’’ to change the manner in which the
possibility of a new or different kind of
ensure that both Category 1 and Type A instrumentation is currently maintained
accident from any accident previously since these instruments are currently
evaluated? Regulatory Guide (RG) 1.97, Revision 3,
‘‘Instrumentation for Light-Water- designated as Type A and/or Category 1
Response: No. variables in the ANO–2 SAR. However,
The proposed changes implement Cooled Nuclear Power Plants to Assess
including these instruments within the TSs
methodologies that have been approved by Plant and Environs Conditions During will now require different mitigating actions
the NRC (provided that any conditions/ and Following an Accident,’’ during periods of inoperability, which may
limitations are satisfied). The P–T limits will instrumentation is included in the include a plant shutdown, establishment of
ensure the protection consistent with specification (unless already addressed alternate monitoring methods, and/or
assuring the integrity of the reactor coolant within another specification) and gains submittal of a special report to the NRC.
pressure boundary as was previously Therefore, this change does not involve a
greater consistency with NUREG–1432,
evaluated. Reactor coolant pressure boundary significant increase in the probability or
integrity will continue to be maintained in Revision 4, ‘‘Standard Technical
Specifications for Combustion consequences of an accident previously
accordance with 10 CFR 50, Appendix G, and evaluated.
the assumed accident performance of plant Engineering Plants.’’ 2. Does the proposed change create the
structures, systems and components will not Basis for proposed no significant possibility of a new or different kind of
be affected. These changes do not involve hazards consideration determination: accident from any accident previously
any physical alteration of the plant (i.e., no As required by 10 CFR 50.91(a), the evaluated?
new or different type of equipment will be licensee has provided its analysis of the Response: No.
installed), and installed equipment is not issue of no significant hazards The proposed change does not result in a
being operated in a new or different manner. change in the manner in which the plant is
consideration, which is presented
Thus, no new failure modes are introduced. operated during post-accident conditions and
Therefore, this change does not create the below:
does not change the established mitigating
possibility of a new or different kind of 1. Does the proposed change involve a actions associated with any necessary
accident from an accident previously significant increase in the probability or response to a DBA [design-basis accident].
evaluated. consequences of an accident previously The proposed change continues to ensure
3. Does the proposed change involve a evaluated? important instrumentation remains available
significant reduction in a margin of safety? Response: No. to station operators such that currently
Response: No. The PAM [Post-Accident Monitoring] established mitigating actions are not
The proposed changes do not affect the instrumentation is not an initiator of any impacted. The change does not involve a
function of the reactor coolant pressure design basis accident or event and, therefore, physical alteration of the plant (i.e., no new
boundary or its response during plant the proposed change does not increase the or different type of equipment will be
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transients. By calculating the P–T limits probability of any accident previously installed) or a change in the methods
using NRC-approved methodology, adequate evaluated. The proposed change ensures governing normal or post-accident plant
margins of safety relating to reactor coolant required instrumentation is included in and operation. The change does not alter
pressure boundary integrity are maintained. controlled by the station TSs and does not assumptions made in the safety analysis.
The proposed changes do not alter the change the response of the plant to any Therefore, this change does not create the
manner in which safety limits, limiting safety accidents. possibility of a new or different kind of
system settings, or limiting conditions for The proposed change does not adversely accident from an accident previously
operation are determined. These changes will affect accident initiators or precursors, nor evaluated.
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![8516 Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices
3. Does the proposed change involve a licensee has provided its analysis of the The existing Updated Final Safety Analysis
significant reduction in a margin of safety? issue of no significant hazards Report accident analysis will continue to
Response: No. consideration, which is presented below meet requirements for the scope and type of
The proposed change does not alter the with NRC: accidents that require analysis.
manner in which safety limits, limiting safety Therefore, the proposed amendment will
system settings, or limiting conditions for 1. Will operation of the facility in not create the possibility of a new or different
operation are determined. The safety analysis accordance with this proposed change kind of accident than those previously
acceptance criteria and assumptions are not involve a significant increase in the evaluated.
impacted by the proposed change. The probability or consequences of an accident 3. Will operation of the facility in
proposed change will not result in plant previously evaluated? accordance with this proposed change
operation in a configuration outside the Response: No. involve a significant reduction in a margin of
design basis. The proposed change ensures The proposed amendment is to incorporate safety?
appropriate PAM instrumentation is the TMRE methodology into the GGNS Response: No.
controlled by the station TSs and that UFSAR. The TMRE methodology is an The proposed amendment is to incorporate
specified remedial action will be taken when alternative methodology for determining the TMRE methodology into the GGNS
required instrumentation is inoperable. The whether protection from tornado-generated UFSAR. The TMRE methodology is an
proposed change continues to support the missiles is required. The methodology can alternative methodology for determining
operator ability to monitor and control vital only be applied to discovered conditions whether protection from tornado-generated
systems during post-accident conditions. where tornado missile protection was not missiles is required. The methodology can
Therefore, this change does not involve a provided, and cannot be used to avoid only be applied to discovered conditions
significant reduction in a margin of safety. providing tornado missile protection in the where tornado missile protection was not
plant modification process. provided, and cannot be used to avoid
The NRC staff has reviewed the The proposed amendment does not involve providing tornado missile protection in the
licensee’s analysis and, based on this an increase in the probability of an accident plant modification process.
review, it appears that the three previously evaluated. The relevant accident The change does not exceed or alter any
standards of 10 CFR 50.92(c) are previously evaluated is a Design Basis controlling numerical value for a parameter
satisfied. Therefore, the NRC staff Tornado impacting the GGNS site. The established in the UFSAR or elsewhere in the
proposes to determine that the probability of a Design Basis Tornado is GGNS licensing basis related to design basis
driven by external factors and is not affected or safety limits. The change does not impact
amendment request involves no
by the proposed amendment. There are no any UFSAR Chapter 6 or 15 Safety Analyses,
significant hazards consideration. changes required to any of the previously and those analyses remain valid. The change
Attorney for licensee: Ms. Anna evaluated accidents in the UFSAR. does not reduce diversity or redundancy as
Vinson Jones, Senior Counsel, Entergy The proposed amendment does not involve required by regulation or credited in the
Services, Inc., 101 Constitution Avenue a significant increase in the consequences or UFSAR. The change does not reduce defense-
NW, Suite 200 East, Washington, DC a Design Basis Tornado. [The methodology as in-depth as described in the UFSAR.
20001. proposed does not alter any input Therefore, the changes associated with this
NRC Branch Chief: Robert J. assumptions or results of the accident license amendment request do not involve a
Pascarelli. analyses. Instead, it reflects a methodology to significant reduction in the margin of safety.
more realistically evaluate the probability of
Entergy Operations, Inc., System Energy unacceptable consequences of a Design Basis The NRC staff has reviewed the
Resources, Inc., Cooperative Energy, A Tornado. As such, there is no significant licensee’s modified analysis and, based
Mississippi Electric Cooperative, and increase in the consequence of an accident on this review, it appears that the three
Entergy Mississippi, Inc., Docket No. 50– previously evaluated. A similar consideration standards of 10 CFR 50.92(c) are
would apply in the event additional non- satisfied. Therefore, the NRC staff
416, Grand Gulf Nuclear Station, Unit 1
conforming conditions are discovered in the proposes to determine that the
(GGNS), Claiborne County, Mississippi future.] amendment request involves no
Date of amendment request: 2. Will operation of the facility in
significant hazards consideration.
November 3, 2017, as supplemented by accordance with this proposed change create
the possibility of a new or different kind of
Attorney for licensee: William B.
letters dated December 6, 2017, and Glew, Associate General Counsel,
accident from any accident previously
January 22, 2018. Publicly-available Entergy Services, Inc., 440 Hamilton
evaluated?
versions are in ADAMS under Response: No. Avenue, White Plains, New York 10601.
Accession Nos. ML17307A440, The proposed amendment is to incorporate NRC Branch Chief: Douglas A.
ML17340B025, and ML18022A598, the TMRE methodology into the GGNS Broaddus.
respectively. UFSAR. The TMRE methodology is an
Description of amendment request: alternative methodology for determining Florida Power & Light Company, Docket
The amendment would revise the GGNS whether protection from tornado-generated Nos. 50–250 and 50–251, Turkey Point
Updated Final Safety Analysis Report missiles is required. The methodology can Nuclear Generating Unit Nos. 3 and 4,
(UFSAR) to incorporate the Tornado only be applied to discovered conditions Miami-Dade County, Florida
where tornado missile protection was not
Missile Risk Evaluator (TMRE) Date of amendment request:
provided, and cannot be used to avoid
methodology contained in Nuclear providing tornado missile protection in the December 21, 2017. A publicly-available
Energy Institute (NEI) 17–02, Revision 1, plant modification process. version is in ADAMS under Accession
‘‘Tornado Missile Risk (TMRE) Industry The proposed amendment will involve no No. ML17355A184.
Guidance Document,’’ September 2017 physical changes to the existing plant, so no Description of amendment request:
(ADAMS Accession No. ML17268A036). new malfunctions could create the possibility The amendments would revise the
This methodology can only be applied of a new or different kind of accident. The Technical Specifications (TSs)
to discovered conditions where tornado proposed amendment makes no changes to pertaining to the Engineered Safety
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missile protection is not currently conditions external to the plant that could Features Actuation System
create the possibility of a new or different
provided, and cannot be used to avoid instrumentation to resolve non-
kind of accident. The proposed change will
providing tornado missile protection in not create the possibility of a new or different conservative actions associated with the
the plant modification process. kind of accident due to new accident containment ventilation isolation and
Basis for proposed no significant precursors, failure mechanisms, the control room ventilation isolation
hazards consideration determination: malfunctions, or accident initiators not functions. In addition, the amendments
As required by 10 CFR 50.91(a), the considered in the design and licensing bases. would revise the control room
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![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8517
ventilation isolation function to no Control Room isolation on Containment high- As required by 10 CFR 50.91(a), the
longer credit containment radiation radiation instrumentation function which is licensee has provided its analysis of the
monitoring instrumentation, eliminate supported by revised accident analyses issue of no significant hazards
which demonstrate that the radiological
redundant radiation monitoring consequences remain within applicable
consideration, which is presented
instrumentation requirements, eliminate regulatory limits. The elimination of core below:
select core alterations applicability alterations applicability requirements do not 1. Does the proposed change involve a
requirements, relocate radiation impact the outcome of any applicable significant increase in the probability or
monitoring and reactor coolant system postulated accident since none result in fuel consequences of an accident previously
leakage detection requirements within cladding damage. No physical changes are evaluated?
the TSs to align with their respective made to the plant, so no new causal Response: No.
mechanisms are introduced. Therefore, the The proposed change addresses conditions
functions, and relocate the spent fuel
proposed changes do not create the during which the secondary containment SR
pool area monitoring requirements to possibility of a new or different kind of
licensee-controlled documents. is not met. The secondary containment is not
accident from any accident previously an initiator of any accident previously
Basis for proposed no significant evaluated. evaluated. As a result, the probability of any
hazards consideration determination: 3. Does the proposed amendment involve accident previously evaluated is not
As required by 10 CFR 50.91(a), the a significant reduction in a margin of safety? increased. The consequences of an accident
licensee has provided its analysis of the Response: No. previously evaluated while utilizing the
issue of no significant hazards The ability of any operable SSC to perform proposed changes are no different than the
consideration, which is presented its designated safety function is unaffected by consequences of an accident while utilizing
the proposed changes. The proposed change the existing four-hour Completion Time for
below: do not revise any safety limits or limiting an inoperable secondary containment. As a
1. Does the proposed amendment involve safety system settings. The proposed changes result, the consequences of an accident
a significant increase in the probability or revises safety analyses assumptions and the previously evaluated are not significantly
consequences of an accident previously method of operating the plant with regard to increased.
evaluated? the Control Room isolation on Containment Therefore, the proposed change does not
Response: No. high-radiation instrumentation function. The
The instrumentation associated with the involve a significant increase in the
changes are supported by revised accident
proposed changes to the technical probability or consequences of an accident
analyses which demonstrate that no adverse
specifications (TS) is not an initiator of any previously evaluated.
impact will result to either the plant
accidents previously evaluated, so the 2. Does the proposed change create the
operating margins or the reliability of
probability of accidents previously evaluated possibility of a new or different kind of
equipment credited in the safety analyses.
is unaffected by the proposed changes. There accident from any previously evaluated?
The existing margin in dose assessment
is no change to any equipment response or Response: No.
currently afforded Control Room operators
accident scenario, with the exception of the The proposed change does not alter the
during any design basis accident is
Control Room isolation on Containment high- protection system design, create new failure
maintained. No other safety margins are
radiation instrumentation function which impacted by the proposed changes. modes, or change any modes of operation.
impose no additional challenges to fission Therefore, the proposed changes do not The proposed change does not involve a
product barrier integrity. The exception is involve a significant reduction in the margin physical alteration of the plant; and no new
supported by revised radiological analyses of safety. or different kind of equipment will be
which demonstrate that the Control Room air installed. Consequently, there are no new
intake radioactivity monitoring The NRC staff has reviewed the initiators that could result in a new or
instrumentation provides timely automatic licensee’s analysis and, based on this different kind of accident.
isolation of the Control Room ventilation review, it appears that the three Therefore, the proposed change does not
system and thereby limits Control Room standards of 10 CFR 50.92(c) are create the possibility of a new or different
operator doses to within regulatory limits for satisfied. Therefore, the NRC staff kind of accident from any previously
any design basis accident. The proposed evaluated.
proposes to determine that the
changes also eliminate limitations imposed 3. Does the proposed change involve a
amendment request involves no significant reduction in a margin of safety?
on Containment and Control Room
ventilation instrumentation during CORE significant hazards consideration. Response: No.
ALTERATIONS since the applicable Attorney for licensee: William S. The proposed change addresses conditions
postulated accidents do not result in fuel Blair, Managing Attorney—Nuclear, during which the secondary containment SR
cladding integrity damage. Hence, the Florida Power & Light Company, 700 is not met. The allowance for both an inner
capability of any TS-required SSC [structure, Universe Blvd. MS LAW/JB, Juno and outer secondary containment door to be
system, or component] to perform its Beach, FL 33408–0420. open simultaneously for entry and exit does
specified safety function is not impacted by NRC Branch Chief: Undine Shoop. not affect the safety function of the secondary
the proposed changes and the outcomes of containment as the doors are promptly closed
accidents previously evaluated are NextEra Energy Duane Arnold, LLC, after entry or exit, thereby restoring the
unaffected. Therefore, the proposed changes Docket No. 50–331, Duane Arnold secondary containment boundary.
do not result in a significant increase in the Energy Center, Linn County, Iowa Therefore, the proposed change does not
probability or consequences of an accident involve a significant reduction in a margin of
previously evaluated. Date of amendment request: safety.
2. Does the proposed amendment create November 10, 2017. A publicly-
the possibility of a new or different kind of available version is in ADAMS under The NRC staff has reviewed the
accident from any accident previously Accession No. ML17318A240. licensee’s analysis and, based on this
evaluated? Description of amendment request: review, it appears that the three
Response: No. The proposed amendment revises standards of 10 CFR 50.92(c) are
No new accident scenarios, failure Technical Specification (TS) 3.6.4.1, satisfied. Therefore, the NRC staff
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mechanisms, or limiting single failures are ‘‘Secondary Containment,’’ Surveillance proposes to determine that the
introduced as a result of the proposed amendment request involves no
Requirement (SR) 3.6.4.1.2. The SR is
changes. The changes do not challenge the
integrity or performance of any safety-related modified to acknowledge that secondary significant hazards consideration.
systems. No plant equipment is installed or containment access openings may be Attorney for licensee: William Blair,
removed, and the changes do not alter the open for entry and exit. P. O. Box 14000, Juno Beach, FL 33408–
design, configuration, or method of operation Basis for proposed no significant 0420.
of any plant SSC with the exception of the hazards consideration determination: NRC Branch Chief: David J. Wrona.
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![8518 Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices
Northern States Power Company— 3. Does the proposed amendment involve As required by 10 CFR 50.91(a), licensee
Minnesota (NSPM), Docket No. 50–263, a significant reduction in a margin of safety? has provided its analysis of the issue on
Monticello Nuclear Generating Plant, Response: No. no significant hazards consideration
The design, operation, testing methods,
Wright County, Minnesota and acceptance criteria for systems,
determination, which is presented
structures, and components (SSCs), specified below:
Date of amendment request:
December 19, 2017. A publicly-available in applicable codes and standards (or 1. Does the proposed amendment involve
alternatives approved for use by the NRC) a significant increase in the probability or
version is in ADAMS under Accession will continue to be met as described in the
No. ML17353A189. consequences of an accident previously
plant licensing basis (including the final evaluated?
Description of amendment request: safety analysis report and bases to TS), since Response: No.
The proposed amendment would adopt these are not affected by changes to the The proposed consistency and editorial
Technical Specifications Task Force surveillance frequencies. Similarly, there is changes to COL Appendix C (and associated
(TSTF) traveler TSTF–425, ‘‘Relocate no impact to safety analysis acceptance plant-specific Tier 1) and Tier 2 and Tier 2*
Surveillance Frequencies to Licensee criteria as described in the plant licensing information in the UFSAR do not involve a
basis. To evaluate a change in the relocated
Control—RITSTF [Risk-Informed technical change, (e.g. there is no design
surveillance frequency, [NSPM] will perform
Technical Specifications Task Force parameter or requirement, calculation,
a probabilistic risk evaluation using the
Initiative 5b,’’ Revision 3. analysis, function or qualification change).
guidance contained in NRC approved NEI
04–10, Rev. 1 in accordance with the TS No structure, requirement, calculation,
Basis for proposed no significant analysis, function or qualification change).
hazards consideration determination: SFCP. NEI 04–10, Rev. 1, methodology
provides reasonable acceptance guidelines No structure, system, or component (SSC)
As required by 10 CFR 50.91(a), the design or function would be affected. No
and methods for evaluating the risk increase
licensee has provided its analysis of the of proposed changes to surveillance design or safety analysis would be affected.
issue of no significant hazards frequencies consistent with Regulatory Guide The proposed changes do not affect any
consideration, which is presented 1.177. accident initiating event or component
below: Therefore, the proposed changes do not failure, thus the probabilities of the accidents
involve a significant reduction in a margin of previously evaluated are not affected. No
1. Does the proposed amendment involve
safety. function used to mitigate a radioactive
a significant increase in the probability or
material release and no radioactive material
consequences of an accident previously The NRC staff has reviewed the release source term is involved, thus the
evaluated? licensee’s analysis and, based on this radiological releases in the accident analyses
Response: No. review, it appears that the three are not affected.
The proposed change relocates the standards of 10 CFR 50.92(c) are Therefore, the proposed amendment does
specified frequencies for periodic
satisfied. Therefore, the NRC staff not involve a significant increase in the
surveillance requirements to licensee control
under a new SFCP [Surveillance Frequency proposes to determine that the probability or consequences of an accident
amendment request involves no previously evaluated.
Control Program]. Surveillance frequencies
are not an initiator to any accident previously significant hazards consideration. 2. Does the proposed amendment create
evaluated. As a result, the probability of any Attorney for licensee: Peter M. Glass, the possibility of a new or different kind of
accident previously evaluated is not Assistant General Counsel, Xcel Energy accident from any accident previously
significantly increased. The systems and Services, Inc., 414 Nicollet Mall, evaluated?
components required by the technical Response: No.
Minneapolis, MN 55401.
specifications for which the surveillance The proposed consistency and editorial
NRC Branch Chief: David J. Wrona.
frequencies are relocated are still required to changes to COL Appendix C (and associated
be operable, meet the acceptance criteria for Southern Nuclear Operating Company, plant specific Tier 1) and Tier 2 and Tier 2*
the surveillance requirements, and be Docket Nos. 52–025 and 52–026, Vogtle information in the UFSAR do not change the
capable of performing any mitigation Electric Generating Plant (VEGP), Units design or functionality of safety-related SSCs.
function assumed in the accident analysis. 3 and 4, Burke County, Georgia The proposed change does not affect plant
As a result, the consequences of any accident electrical systems, and does not affect the
previously evaluated are not significantly
Date of amendment request: design function, support, design, or operation
increased. November 30, 2017. A publicly- of mechanical and fluid systems. The
Therefore, the proposed change does not available version is in ADAMS under proposed change does not result in a new
involve a significant increase in the Accession No. ML17334B211. failure mechanism or introduce any new
probability or consequences of an accident Description of amendment request: accident precursors. No design function
previously evaluated. The proposed changes include changes described in the UFSAR is affected by the
2. Does the proposed amendment create to the Updated Final Safety Analysis proposed changes.
the possibility of a new or different kind of Report (UFSAR) in the form of Therefore, the proposed amendment does
accident from any previously evaluated? not create the possibility of a new or different
departures from the incorporated plant- kind of accident from any accident
Response: No. specific Design Control Document
No new or different accidents result from previously evaluated.
utilizing the proposed change. The changes
(DCD) Tier 2* and Tier 1 information 3. Does the proposed amendment involve
do not involve a physical alteration of the and related changes to the VEGP Units a significant reduction in a margin of safety?
plant (i.e., no new or different type of 3 and 4 Combined License (COL) Response: No.
equipment will be installed) or a change in Appendix C information. Pursuant to The proposed consistency and editorial
the methods governing normal plant the provisions of 10 CFR 52.63(b)(1), an changes to COL Appendix C (and associated
operation. In addition, the changes do not exemption from the elements of the plant specific Tier 1) and Tier 2 and Tier 2*
impose any new or different requirements. design as certified in 10 CFR part 52, information in the UFSAR do not involve any
The changes do not alter assumptions made Appendix D, design certification rule is change to the design as described in the COL.
daltland on DSKBBV9HB2PROD with NOTICES
in the safety analysis. The proposed changes There would be no change to an existing
also requested for the plant-specific Tier design basis, design function, regulatory
are consistent with the safety analysis
assumptions and current plant operating
1 material departures. This submittal criterion, or analysis. No safety analysis or
practice. requests approval of the license design basis acceptance limit/criterion is
Therefore, the proposed changes do not amendment, necessary to implement involved.
create the possibility of a new or different these changes. Therefore, the proposed amendment does
kind of accident from any accident Basis for proposed no significant not involve a significant reduction in a
previously evaluated. hazards consideration determination: margin of safety.
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![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8519
The NRC staff has reviewed the The changes do not impact the support, Therefore, the proposed changes do not
licensee’s analysis and, based on this design, or operation of mechanical and fluid involve a significant reduction in a margin of
review, it appears that the three systems. The changes do not impact the safety.
standards of 10 CFR 50.92(c) are
support, design, or operation of any safety- The NRC staff has reviewed the
related structures. There is no change to licensee’s analysis and, based on this
satisfied. Therefore, the NRC staff plant systems or the response of systems to
proposes to determine that the review, it appears that the three
postulated accident conditions. There is no standards of 10 CFR 50.92(c) are
amendment request involves no change to the predicted radioactive releases
significant hazards consideration. due to normal operation or postulated satisfied. Therefore, the NRC staff
accident conditions. The plant response to proposes to determine that the
Attorney for licensee: M. Stanford
previously evaluated accidents or external amendment request involves no
Blanton, Balch & Bingham LLP, 1710
events is not adversely affected, nor do the significant hazards consideration.
Sixth Avenue North, Birmingham, AL proposed changes create any new accident Attorney for licensee: M. Stanford
35203–2015. precursors. Blanton, Balch & Bingham LLP, 1710
NRC Branch Chief: Jennifer Dixon- Therefore, the proposed amendment does Sixth Avenue North, Birmingham, AL
Herrity. not involve a significant increase in the 35203–2015.
probability or consequences of an accident
Southern Nuclear Operating Company, NRC Branch Chief: Jennifer Dixon-
previously evaluated.
Docket Nos. 52–025 and 52–026, Vogtle 2. Does the proposed amendment create Herrity.
Electric Generating Plant, Units 3 and 4, the possibility of a new or different kind of Southern Nuclear Operating Company,
Burke County, Georgia accident from any accident previously Inc., Docket Nos. 52–025 and 52–026,
evaluated?
Date of amendment request: February Response: No. Vogtle Electric Generating Plant, Units 3
1, 2018. A publicly-available version is The proposed changes are to relax the and 4, Burke County, Georgia
in ADAMS under Accession No. minimum gap requirement above grade Date of amendment request: January
ML18032A359. between the nuclear island and the annex 31, 2018. A publicly-available version is
Description of amendment request: building/turbine building from a 4 inch gap
in ADAMS under Accession No.
to a 3 inch gap. The proposed changes
The requested amendment proposes modify and clarify the gap requirements ML18031B142.
changes to relax the minimum gap between the nuclear island and the annex Description of amendment request:
requirement above grade between the building/turbine building and radwaste The requested amendment proposes to
nuclear island and the annex building/ building, respectively. The proposed changes include changes to Combined License
turbine building and removing the delete the gap requirement for the radwaste (COL) Appendix A, Technical
minimum gap requirement for the building from the ITAAC in COL Appendix Specifications related to fuel
radwaste building from the Inspections, C. The proposed changes do not affect the management. Specifically, the requested
operation of any systems or equipment that
Tests, Analyses and Acceptance amendment proposes improvements to
may initiate a new or different kind of
Criteria. Pursuant to the provisions of 10 accident, or alter any SSC such that a new the technical specifications for the Rod
CFR 52.63(b)(1), an exemption from accident initiator or initiating sequence of Position Indication, the Control Rod
elements of the design as certified in the events is created. Drive Mechanism, Power Range Neutron
10 CFR part 52, Appendix D, design The proposed changes do not adversely Flux Channels and the Mechanical
certification rule is also requested for affect the design function of the nuclear Shim Augmentation.
the plant-specific Design Control island and adjoining buildings’ SSC design Basis for proposed no significant
Document Tier 1 material departures. functions or methods of operation in a hazards consideration determination:
manner that results in a new failure mode, As required by 10 CFR 50.91(a), the
Basis for proposed no significant malfunction, or sequence of events that affect
hazards consideration determination: licensee has provided its analysis of the
safety-related or non-safety-related
As required by 10 CFR 50.91(a), the equipment. This activity does not allow for issue of no significant hazards
licensee has provided its analysis of the a new fission product release path, result in consideration, which is presented
issue of no significant hazards a new fission product barrier failure mode, or below:
consideration, which is presented below create a new sequence of events that result 1. Does the proposed amendment involve
with NRC staff edits in square brackets: in significant fuel cladding failures. a significant increase in the probability or
Therefore, the proposed amendment does consequences of an accident previously
1. Does the proposed amendment involve not create the possibility of a new or different evaluated?
a significant increase in the probability or kind of accident from any accident Response: No.
consequences of an accident previously previously evaluated. The proposed changes are to clarify proper
evaluated? 3. Does the proposed amendment involve operation and methodology associated with
Response: No. a significant reduction in a margin of safety? the DRPI [Digital Rod Position Indication],
The proposed changes are to relax the Response: No. Control Rod Gripper Coils, instrumentation
minimum gap requirement above grade The proposed changes maintain existing associated with Quadrant Power Tilt Ratio, or
between the nuclear island and the annex safety margin and provide adequate Control or Gray Rods. These changes do not
building/turbine building from a 4 inch gap protection through continued application of affect the operation of this equipment and
to a 3 inch gap. The proposed changes the existing requirements in the UFSAR have no adverse impact on their design
modify and clarify the gap requirements [Updated Final Safety Analysis Report]. The functions.
between the nuclear island and the annex proposed changes satisfy the same design The changes do not involve an interface
building/turbine building and radwaste functions in accordance with the same codes with any structure, system, or component
building, respectively. The proposed change and standards as stated in the UFSAR. These (SSC) accident initiator or initiating sequence
deletes the gap requirement for the radwaste changes do not adversely affect any design of events, and thus, the probabilities of the
building from the Inspections, Tests, code, function, design analysis, safety accidents evaluated in the plant-specific
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Analyses and Acceptance Criteria (ITAAC) in analysis input or result, or design/safety Updated Final Safety Analysis Report
(COL) [Combined License] Appendix C. The margin. No safety analysis or design basis (UFSAR) are not affected. The proposed
proposed changes do not affect the operation acceptance limit/criterion is challenged or changes do not adversely affect any
of any systems or equipment that may initiate exceeded by the proposed changes. mitigation sequence or the predicted
a new or different kind of accident, or alter Because no safety analysis or design basis radiological releases due to postulated
any structure, system or component (SSC) acceptance limit/criterion is challenged or accident conditions, thus, the consequences
such that a new accident initiator or exceeded by these changes, no significant of the accidents evaluated in the UFSAR are
initiating sequence of events is created. margin of safety is reduced. not affected.
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![8520 Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices
Therefore, the proposed amendment does Sixth Avenue North, Birmingham, AL Response: No.
not involve a significant increase in the 35203–2015. The proposed change does not alter the
probability or consequences of an accident NRC Branch Chief: Jennifer Dixon- protection system design, create new failure
previously evaluated. Herrity. modes, or change any modes of operation.
2. Does the proposed amendment create The proposed change does not involve a
the possibility of a new or different kind of Susquehanna Nuclear, LLC, Docket Nos. physical alteration of the plant; and no new
accident from any accident previously 50–387 and 50–388, Susquehanna or different kind of equipment will be
evaluated? Steam Electric Station, Units 1 and 2, installed. Consequently, there are no new
Response: No. initiators that could result in a new or
Luzerne County, Pennsylvania different kind of accident.
The proposed changes verify and maintain
the capabilities of the DRPI, Control Rod Date of amendment request: Therefore, the proposed change does not
Gripper Coils, instrumentation associated December 14, 2017. A publicly-available create the possibility of a new or different
with Quadrant Power Tilt Ratio, and Control version is in ADAMS under Accession kind of accident from any previously
and Gray Rods to perform their design No. ML17348B097. evaluated.
functions. The proposed changes do not 3. Does the proposed change involve a
Description of amendment request: significant reduction in a margin of safety?
affect the operation of any systems or The amendments would revise
equipment that may initiate a new or Response: No.
different kind of accident, or alter any SSC Technical Specification (TS) 3.6.4.1, The proposed change addresses conditions
such that a new accident initiator or ‘‘Secondary Containment,’’ Surveillance during which the secondary containment SR
initiating sequence of events is created. Requirement (SR) 3.6.4.1.1. The SR is not met. Conditions in which the
The proposed changes do not affect any would be revised to address conditions secondary containment vacuum is less than
other SSC design functions or methods of during which the secondary the required vacuum are acceptable provided
operation in a manner that results in a new containment pressure may not meet the the conditions do not affect the ability of the
failure mode, malfunction, or sequence of SGT [Standby Gas Treatment] System to
SR pressure requirements. The proposed establish the required secondary containment
events that affect safety-related or nonsafety changes are based on Technical
related equipment. Therefore, this activity vacuum under post-accident conditions
does not allow for a new fission product Specifications Task Force (TSTF) within the time assumed in the accident
release path, result in a new fission product Traveler TSTF–551, Revision 3, ‘‘Revise analysis. This condition is incorporated in
barrier failure mode, or create a new Secondary Containment Surveillance the proposed change by requiring an analysis
sequence of events that result in significant Requirements.’’ Also, the editorial note of actual environmental and secondary
fuel cladding failures. These changes are to in SR 3.6.4.1.3 is removed because it is containment pressure conditions to confirm
clarify proper operation and methodology the capability of the SGT System is
redundant to the SR itself and does not maintained within the assumptions of the
associated with this equipment and have no alter the requirement.
adverse impact on their design functions. accident analysis. Therefore, the safety
Basis for proposed no significant function of the secondary containment is not
Therefore, the proposed amendment does
hazards consideration determination: affected.
not create the possibility of a new or different
kind of accident from any accident As required by 10 CFR 50.91(a), the Therefore, the proposed change does not
previously evaluated. licensee has provided its analysis of the involve a significant reduction in a margin of
3. Does the proposed amendment involve issue of no significant hazards safety.
a significant reduction in a margin of safety? consideration, which is presented The NRC staff has reviewed the
Response: No. below, along with NRC edits in square licensee’s analysis and, based on this
The proposed changes do not affect brackets: review, it appears that the three
existing safety margins. The proposed
changes verify and maintain the capabilities 1. Does the proposed change involve a standards of 10 CFR 50.92(c) are
of the DRPI, Control Rod Gripper Coils, significant increase in the probability or satisfied. Therefore, the NRC staff
instrumentation associated with Quadrant consequences of an accident previously proposes to determine that the
Power Tilt Ratio, and Control and Gray Rods evaluated? amendment request involves no
to perform their design functions. Therefore, Response: No. significant hazards consideration.
the proposed changes satisfy the same design The proposed change addresses conditions Attorney for licensee: Damon D. Obie,
functions in accordance with the same codes during which the secondary containment SR
is not met. The secondary containment is not
Associate General Counsel, Talen
and standards as stated in the UFSAR. These Energy Supply, LLC, 835 Hamilton St.,
changes do not affect any design code, an initiator of any accident previously
evaluated. As a result, the probability of any Suite 150, Allentown, PA 18101.
function, design analysis, safety analysis NRC Branch Chief: James G. Danna.
input or result, or design/safety margin. accident previously evaluated is not
The proposed changes would not affect any increased. The consequences of an accident
Tennessee Valley Authority, Docket
safety-related design code, function, design previously evaluated while utilizing the
proposed changes are no different than the Nos. 50–327 and 50–328, Sequoyah
analysis, safety analysis input or result, or Nuclear Plant (SQN), Units 1 and 2,
existing design/safety margin. Because no consequences of an accident while utilizing
the existing four hour Completion Time for Hamilton County, Tennessee
safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by an inoperable secondary containment. In Date of amendment request:
addition, the proposed Note for SR 3.6.4.1.1
the requested changes, no margin of safety is September 29, 2017. A publicly-
significantly reduced. provides an alternative means to ensure the
secondary containment safety function is available version is in ADAMS under
Therefore, the proposed amendment does Accession No. ML17272A940.
met. Additionally, the Note removed from SR
not involve a significant reduction in a Description of amendment request:
3.6.4.1.3 is editorial because it is redundant
margin of safety. The amendments would make changes
to the SR itself and does not alter the
The NRC staff has reviewed the requirement. As a result, the consequences of to the SQN Emergency Plan to extend
licensee’s analysis and, based on this an accident previously evaluated are not staff augmentation times for Emergency
review, it appears that the three significantly increased. Response Organization (ERO) functions.
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standards of 10 CFR 50.92(c) are Therefore, the proposed change does not Basis for proposed no significant
involve a significant increase in the hazards consideration determination:
satisfied. Therefore, the NRC staff
probability or consequences of an accident
proposes to determine that the previously evaluated.
As required by 10 CFR 50.91(a), the
amendment request involves no 2. Does the proposed change create the licensee has provided its analysis of the
significant hazards consideration. possibility of a new or different kind of issue of no significant hazards
Attorney for licensee: Mr. M. Stanford accident from any accident previously consideration, which is presented
Blanton, Balch & Bingham LLP, 1710 evaluated? below:
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![8522 Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices
Duke Energy Carolinas, LLC, Docket consideration determination as consideration determination as
Nos. 50–413 and 50–414, Catawba published in the Federal Register. published in the Federal Register.
Nuclear Station, Units 1 and 2, York The Commission’s related evaluation The Commission’s related evaluation
County, South Carolina of the amendment is contained in a of the amendments is contained in a
Duke Energy Carolinas, LLC, Docket safety evaluation dated February 1, Safety Evaluation dated January 31,
Nos. 50–369 and 50–370, McGuire 2018. 2018.
Nuclear Station, Units 1 and 2, No significant hazards consideration No significant hazards consideration
Mecklenburg County, North Carolina comments received: No. comments received: No.
Duke Energy Carolinas, LLC, Docket Duke Energy Carolinas, LLC, Docket Exelon Generation Company, LLC,
Nos. 50–269, 50–270, and 50–287, Nos. 50–369 and 50–370, McGuire Docket Nos. 50–373 and 50–374, LaSalle
Oconee Nuclear Station, Units 1, 2, and Nuclear Station, Units 1 and 2 County Station (LSCS), Units 1 and 2,
3, Oconee County, South Carolina (McGuire), Mecklenburg County, North LaSalle County, Illinois
Carolina Date of amendment request: August
Duke Energy Progress, LLC, Docket No.
50–400, Shearon Harris Nuclear Power Date of amendment requests: 29, 2017, as supplemented by letter
Plant, Unit 1, Wake County, North December 19, 2016, as supplemented by dated January 25, 2018.
Carolina letters dated May 25, 2017, and Brief description of amendment: The
December 12, 2017. amendments revised the LSCS technical
Duke Energy Progress, LLC, Docket No. Brief description of amendments: The specification (TS) 2.1.1, ‘‘Reactor Core
50–261, H. B. Robinson Steam Electric amendments modified Technical SLs [Safety Limits].’’ Specifically, this
Plant, Unit No. 2, Darlington County, Specification 5.5.2, ‘‘Containment change incorporates revised LSCS, Units
South Carolina Leakage Rate Testing Program,’’ by 1 and 2, safety limits for minimum
Date of amendment request: July 18, replacing the reference to Regulatory critical power ratio for two circulation
2017, as supplemented by letter dated Guide 1.163, ‘‘Performance-Based loop minimum critical power ratio
October 12, 2017. Containment Leak-Test Program,’’ with (MCPR) and single circulation loop
Brief description of amendments: The a reference to Nuclear Energy Institute MCPR values for Unit 1 and Unit 2
amendments revised the technical (NEI) Topical Report NEI 94–01, based on the results of the cycle-specific
specifications (TSs) based on Technical Revision 3–A, ‘‘Industry Guideline for analyses performed by Global Nuclear
Specifications Task Force (TSTF) Implementing Performance-Based Fuel (GNF) for LSCS Unit 1, Cycle 17,
Traveler TSTF–529, ‘‘Clarify Use and Option of 10 CFR part 50, Appendix J,’’ and LSCS Unit 2, Cycle 17.
Application Rules.’’ The changes revise dated July 2012 and the conditions and Date of issuance: February 6, 2018.
and clarify the TS usage rules for limitations specified in NEI 94–01, Effective date: As of the date of
completion times, limiting conditions Revisions 2–A, ‘‘Industry Guideline for issuance and shall be implemented as
for operation, and surveillance Implementing Performance-Based follows:
requirements. Option of 10 CFR part 50, Appendix J,’’ Unit 1: Prior to startup from the
Date of issuance: February 1, 2018. dated October 2008, as the February 2018 refueling outage for Unit
Effective date: As of the date of implementation documents used by 1 (i.e., L1R17) for operation starting in
issuance and shall be implemented McGuire to implement the performance- Cycle 18.
within 120 days of issuance. based leakage testing program in Unit 2: Prior to startup from the
Amendment Nos.: 298 and 294, for accordance with Option B of 10 CFR February 2018 refueling outage for Unit
the Catawba Nuclear Station Units 1 and part 50, Appendix J. The proposed 1 (i.e., L1R17). This will be a mid-Cycle
2; 307 and 286, for the McGuire Nuclear change would also delete the listing of 17 implementation for Unit 2.
Station, Units 1 and 2; 407, 409, and one-time exceptions previously granted Amendment No.: Unit 1–227; Unit 2–
408, for the Oconee Nuclear Station, to Integrated Leak Rate Test frequency. 213. A publicly-available version is in
Units 1, 2, and 3; 162, for the Shearon Date of issuance: January 31, 2018. ADAMS under Accession No.
Harris Nuclear Power Plant, Unit 1; and Effective date: As of its date of ML18008A123; documents related to
256, for the H. B. Robinson Steam issuance and shall be implemented these amendments are listed in the
Electric Plant, Unit No. 2. A publicly- within 120 days of issuance. Safety Evaluation enclosed with the
available version is in ADAMS under Amendment Nos.: 306 (Unit 1) and amendments.
Accession No. ML17340A720; 285 (Unit 2). A publicly-available Renewed Facility Operating License
documents related to these amendments version is in ADAMS under Accession Nos. NPF–11 and NPF–18: The
are listed in the Safety Evaluation No. ML18009A842; documents related amendments revised the Renewed
enclosed with the amendments. to these amendments are listed in the Facility Operating Licenses and TSs.
Renewed Facility Operating License Safety Evaluation enclosed with the Date of initial notice in Federal
Nos. NPF–35, NPF–52, NPF–9, NPF–17, amendments. Register: December 5, 2017 (82 FR
DPR–38, DPR–47, DPR–55, NPF–63, and Renewed Facility Operating License 57482). The supplemental letter dated
DPR–23: Amendments revised the Nos. NPF–9 and NPF–17: Amendments January 25, 2018, provided additional
Renewed Facility Operating Licenses revised the Renewed Facility Operating information that clarified the
and TSs. Licenses and Technical Specifications. application, did not expand the scope of
Date of initial notice in Federal Date of initial notice in Federal the application as originally noticed,
Register: August 29, 2017 (82 FR Register: May 9, 2017 (82 FR 21557). and did not change the NRC staff’s
daltland on DSKBBV9HB2PROD with NOTICES
41067). The supplemental letter dated The supplemental letters dated May 25 original proposed no significant hazards
October 12, 2017, provided additional and December 12, 2017, provided consideration determination as
information that clarified the additional information that clarified the published in the Federal Register.
application, did not expand the scope of application, did not expand the scope of The Commission’s related evaluation
the application as originally noticed, the application as originally noticed, of the amendments is contained in a
and did not change the NRC staff’s and did not change the NRC staff’s Safety Evaluation dated February 6,
original proposed no significant hazards original proposed no significant hazards 2018.
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![Federal Register / Vol. 83, No. 39 / Tuesday, February 27, 2018 / Notices 8523
No significant hazards consideration No. ML17296A133; documents related original proposed no significant hazards
comments received: No. to these amendments are listed in the consideration determination as
Safety Evaluation (SE) enclosed with the published in the Federal Register.
Southern Nuclear Operating Company,
amendments. The Commission’s related evaluation
Inc., Docket Nos. 50–424 and 50–425, Facility Operating License Nos. DPR– of the amendment is contained in a
Vogtle Electric Generating Plant, Units 1 77 and DPR–79: The amendments Safety Evaluation dated February 9,
and 2, Burke County, Georgia revised the Renewed Facility Operating 2018.
Date of amendment request: May 24, Licenses and TSs. No significant hazards consideration
2017, as supplemented by letter dated Date of initial notice in Federal comments received: No.
August 17, 2017. Register: July 5, 2017 (82 FR 31102). Dated at Rockville, Maryland, on February
Brief description of amendments: The The supplemental letter dated August 7, 20, 2018.
amendments revised Surveillance 2017, provided additional information For the Nuclear Regulatory Commission.
Requirement 3.3.1.3 to change the that clarified the application, did not
thermal power at which the surveillance Kathryn M. Brock,
expand the scope of the application as
may be performed. originally noticed, and did not change Deputy Director, Division of Operating
Date of issuance: February 7, 2018. Reactor Licensing, Office of Nuclear Reactor
the NRC staff’s original proposed no Regulation.
Effective date: As of the date of significant hazards consideration
issuance and shall be implemented [FR Doc. 2018–03727 Filed 2–26–18; 8:45 am]
determination as published in the
within 90 days of issuance. Federal Register. BILLING CODE 7590–01–P
Amendment Nos.: 194 (Unit 1) and The Commission’s related evaluation
177 (Unit 2). A publicly-available of the amendments is contained in an
version is in ADAMS under Accession SE dated February 2, 2018. OVERSEAS PRIVATE INVESTMENT
No. ML18012A068; documents related No significant hazards consideration CORPORATION
to these amendments are listed in the comments received: No.
Safety Evaluation enclosed with the Sunshine Act Cancellation Notice—
amendments. Virginia Electric and Power Company, OPIC’S March 8, 2018 Annual Public
Renewed Facility Operating License Docket Nos. 50–280 and 50–281, Surry Hearing
No. NPF–68 and NPF–81: Amendments Power Station, Unit Nos. 1 and 2, Surry
County, Virginia OPIC’s Sunshine Act notice of its
revised the Renewed Facility Operating Annual Public Hearing was published
Licenses and Technical Specifications. Date of amendment request: January
in the Federal Register (Volume 83,
Date of initial notice in Federal 20, 2017, as supplemented by letter
Number 13, Page 2823) on January 19,
Register: July 18, 2017 (82 FR 32883). dated September 7, 2017.
Brief description of amendments: The 2018. No requests were received to
The supplemental letter dated August
amendments revised the Technical provide testimony or submit written
17, 2017, provided additional
Specification (TS) 3.5, ‘‘Residual Heat statements for the record; therefore,
information that clarified the
Removal (RHR) System,’’ requirements, OPIC’s Annual Public Hearing
application, did not expand the scope of
as well as the TS 3.13, ‘‘Component scheduled for 10 a.m., March 8, 2018
the application as originally noticed,
Cooling System,’’ RHR support has been cancelled.
and did not change the NRC staff’s
original proposed no significant hazards requirements for consistency with the CONTACT PERSON FOR INFORMATION:
consideration determination as design basis of the RHR system. In Information on the hearing cancellation
published in the Federal Register. addition, an RHR surveillance may be obtained from Catherine F.I.
The Commission’s related evaluation requirement is added in TS Table 4.1– Andrade at (202) 336–8768, or via email
of the amendment is contained in a 2A, ‘‘Minimum Frequency for at Catherine.Andrade@opic.gov.
Safety Evaluation dated February 7, Equipment Tests,’’ to test the RHR Dated: February 22, 2018.
2018. system in accordance with the inservice Catherine F.I. Andrade,
No significant hazards consideration testing program, since a TS surveillance OPIC Corporate Secretary.
comments received: No. does not currently exist for this system. [FR Doc. 2018–04037 Filed 2–23–18; 11:15 am]
Date of issuance: February 9, 2018.
Tennessee Valley Authority, Docket Effective date: As of the date of BILLING CODE 3210–01–P
Nos. 50–327 and 50–328, Sequoyah issuance and shall be implemented
Nuclear Plant (SQN), Units 1 and 2, within 60 days of issuance.
Hamilton County, Tennessee OVERSEAS PRIVATE INVESTMENT
Amendment Nos.: 291 and 291. A
CORPORATION
Date of amendment requests: March publicly-available version is in ADAMS
13, 2017, as supplemented by letter under Accession No. ML17326A225; Sunshine Act Cancellation Notice—
dated August 7, 2017. documents related to these amendments OPIC February 28, 2018 Public Hearing
Brief description of amendments: The are listed in the Safety Evaluation
amendments deleted the Note enclosed with the amendments. OPIC’s Sunshine Act notice of its
associated with Technical Specification Renewed Facility Operating License Public Hearing in Conjunction with
(TS) Surveillance Requirement (SR) No. DPR–32 and DPR–37: Amendments each Board meeting was published in
3.8.1.17 to allow the performance of the revised the Renewed Facility Operating the Federal Register (Volume 83,
SR in Modes 1 through 4. Licenses and TSs. Number 25, Page 5284) on Tuesday,
Date of issuance: February 2, 2018. Date of initial notice in Federal February 6, 2018. No requests were
daltland on DSKBBV9HB2PROD with NOTICES
Effective date: As of its date of Register: March 14, 2017 (82 FR received to provide testimony or submit
issuance and shall be implemented no 13672). The supplemental letter dated written statements for the record;
later than 60 days from the date of September 7, 2017, provided additional therefore, OPIC’s public hearing
issuance. information that clarified the scheduled for 2 p.m., February 28, 2018
Amendment Nos.: 340 (Unit 1) and application, did not expand the scope of in conjunction with OPIC’s March 8,
333 (Unit 2). A publicly-available the application as originally noticed, 2018 Board of Directors meeting has
version is in ADAMS under Accession and did not change the NRC staff’s been cancelled.
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Document Created: 2018-02-27 01:14:52
Document Modified: 2018-02-27 01:14:52
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by March 29, 2018. A request for a hearing must be filed by April 30, 2018. | |
| Contact | Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: [email protected] | |
| FR Citation | 83 FR 8509 |
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