82 FR 23615 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 82, Issue 98 (May 23, 2017)
Federal Register Volume 82, Issue 98 (May 23, 2017)
| Page Range | 23615-23632 | |
| FR Document | 2017-10570 |
[Federal Register Volume 82, Number 98 (Tuesday, May 23, 2017)] [Notices] [Pages 23615-23632] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-10570] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0120] 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 April 25, 2017, to May 8, 2017. The last biweekly notice was published on May 9, 2017. DATES: Comments must be filed by June 22, 2017. A request for a hearing must be filed by July 24, 2017. ADDRESSES: You may submit comments by any of the following methods:Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0120. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: T-8-D36M, 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-2017-0120, 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0120. 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 ADAMS) is provided the first time that it is mentioned in this document. NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2017-0120, 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 [[Page 23616]] 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 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 Web site 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 by July 24, 2017. 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 [[Page 23617]] 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's Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [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 Web site 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 Web site 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 Web site 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 [[Page 23618]] additional direction on obtaining information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station (CNS), Units 1 and 2, York County, South Carolina Date of amendment request: December 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16350A422. Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.4, ``Residual Heat Removal (RHR) and Coolant Circulation--High Water Level,'' and TS 3.9.5, ``Residual Heat Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A of TS 3.9.4 applies when RHR requirements are not met, and includes four required actions. Required Action A.4 requires, within 4 hours, the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes revise Required Action A.4 and add new Required Actions A.5, A.6.1, and A.6.2 to clarify that the intent of the required actions is to establish containment closure. Each of these required actions will have a completion time of 4 hours. Condition B of TS 3.9.5 applies when no RHR loop is in operation, and includes three required actions. Required Action B.3 requires the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes are the same as the proposed changes to TS 3.9.4, consisting of a revision to Required Action B.3 and the addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When Shutdown Cooling Requirements Are Not Met.'' 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 revise the CNS TS to ensure that the appropriate actions are taken to establish containment closure in the event that Residual Heat Removal requirements are not met during refueling operations. Containment closure would be appropriate for mitigation of a loss of shutdown cooling accident, but it does not affect the initiation of the accident. The containment purge system isolation valves will be capable of being closed automatically on a high containment radiation signal, such that there will be no significant increase in the radiological consequences of a loss of shutdown cooling. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal. 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 the margin of safety? Response: No. Currently the Technical Specifications are vague and overly restrictive concerning the requirement for containment closure when shutdown cooling is lost. The proposed changes eliminate unclear requirements and provide a clear way to establish containment closure that meets the [TS] Bases description, which is to prevent radioactive gas from being released from the containment during a loss of shutdown cooling incident. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal. 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: 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, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification 3.1.2, ``Core Reactivity,'' to revise the Completion Times of Required Action A.1 and A.2 from 72 hours to 7 days. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the Completion Time when the Core Reactivity Balance is Not Within Limit.'' 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 extend the Completion Time to take the Required Actions when measured core reactivity is not within the specified limit of the predicted values. The Completion Time to respond to a difference between predicted and measured core reactivity is not an initiator to any accident previously evaluated. The radiological consequences of an accident during the proposed Completion Time are no different from the consequences of an accident during the existing Completion Time. Therefore, the proposed changes do not involved a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. The proposed changes provide additional time to investigate and to implement appropriate operating restrictions when [[Page 23619]] measured core reactivity is not within the specified limit of the predicted values. The additional time will not have a significant effect on plant safety due to the conservatisms used in designing the reactor core and performing the safety analyses, and the low probability of an accident or transient which would approach the core design limits during the additional time. 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: Kate B. Nolan, Deputy General Counsel, Duke Energy Corporation, 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 (MNS), Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.5, ``Residual Heat Removal (RHR) and Coolant Circulation--High Water Level,'' and TS 3.9.6, ``Residual Heat Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A of TS 3.9.5 applies when RHR requirements are not met, and includes four required actions. Required Action A.4 requires, within 4 hours, the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes revise Required Action A.4 and add new Required Actions A.5, A.6.1, and A.6.2 to clarify that the intent of the required actions is to establish containment closure. Each of these required actions will have a completion time of 4 hours. Condition B of TS 3.9.6 applies when no RHR loop is in operation, and includes three required actions. Required Action B.3 requires the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes are the same as the proposed changes to TS 3.9.5, consisting of a revision to Required Action B.3 and the addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When Shutdown Cooling Requirements Are Not Met.'' 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 revise the MNS TS to ensure that the appropriate actions are taken to establish containment closure in the event that Residual Heat Removal requirements are not met during refueling operations. Containment closure would be appropriate for mitigation of a loss of shutdown cooling accident, but it does not affect the initiation of the accident. The containment purge system isolation valves will be capable of being closed automatically on a high containment radiation signal, such that there will be no significant increase in the radiological consequences of a loss of shutdown cooling. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal. 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 the margin of safety? Response: No. Currently the Technical Specifications are vague and overly restrictive concerning the requirement for containment closure when shutdown cooling is lost. The proposed changes eliminate unclear requirements and provide a clear way to establish containment closure that meets the [TS] Bases description, which is to prevent radioactive gas from being released from the containment during a loss of shutdown cooling incident. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal. 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: 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 (MNS), Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,'' to add a Note to TS Limiting Condition for Operation (LCO) 3.6.3 Required Actions A.2, C.2 and E.2 to allow isolation devices that are locked, sealed or otherwise secured to be verified by use of administrative means. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-269-A, Revision 2, ``Allow Administrative Means of Position Verification for Locked or Sealed Valves.'' 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 modify MNS TS 3.6.3, ``Containment Isolation Valves''. This TS currently includes actions that require penetrations to be isolated and periodically verified to be isolated. A Note is proposed to be added to TS 3.6.3 Required Actions A.2, C.2, and E.2, to allow isolation devices that are locked, sealed, or otherwise secured to be verified by use of administrative means. The proposed changes do not affect any plant equipment, test methods, or plant operation, and is not an initiator of any analyzed accident sequence. The inoperable containment penetrations will continue to be isolated, and hence perform their isolation [[Page 23620]] function. Operation in accordance with the proposed TSs will ensure that all analyzed accidents will continue to be mitigated as previously analyzed. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. The proposed changes will not affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. Affected containment penetrations will continue to be isolated as required by the existing TS. 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: 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, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' to allow greater flexibility in performing Surveillance Requirements (SRs) by modifying Mode restriction notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction Notes.'' 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 modify Mode restriction Notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19 to allow performance of the Surveillance in whole or in part to reestablish Diesel Generator (DG) Operability, and to allow the crediting of unplanned events that satisfy the Surveillance(s) [Requirements]. The emergency diesel generators and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. To manage any increase in risk, the proposed changes require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the current prohibited Modes. The radiological consequences of an accident previously evaluated during the period that the DG is being tested to reestablish operability are no different from the radiological consequences of an accident previously evaluated while the DG is inoperable. As a result, the consequences of any accident previously evaluated are not increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. The purpose of Surveillances is to verify that equipment is capable of performing its assumed safety function. The proposed changes will only allow the performance of the Surveillances to reestablish operability, and the proposed changes may not be used to remove a DG from service. In addition, the proposed changes will potentially shorten the time that a DG is unavailable because testing to reestablish operability can be performed without a plant shutdown. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the current prohibited Modes. 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: 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, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.4.12, ``Low Temperature Overpressure Protection (LTOP) System,'' to increase the time allowed for swapping charging pumps to 1 hour. Additionally, an existing note in the Applicability section of TS 3.4.12 is being reworded and relocated to the Limiting Condition for Operation section of TS 3.4.12 as Note 2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-285-A, Revision 1, ``Charging Pump Swap LTOP Allowance.'' 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 increase the time allowed for swapping charging pumps from 15 minutes to one hour, and make several other associated administrative changes and clarifications to the TS. These changes do not affect event initiators or precursors. Thus, the proposed changes do not involve a [[Page 23621]] significant increase in the probability of an accident previously evaluated. In addition, the proposed changes do not alter any assumptions previously made in the radiological consequence evaluations nor affect mitigation of the radiological consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR). As such, the consequences of accidents previously evaluated in the UFSAR will not be increased and no additional radiological source terms are generated. Therefore, there will be no reduction in the capability of those structures, systems, and components (SSCs) in limiting the radiological consequences of previously evaluated accidents, and reasonable assurance that there is no undue risk to the health and safety of the public will continue to be provided. Thus, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not involve physical changes to analyzed SSCs or changes to the modes of plant operation defined in the technical specification. The proposed changes do not involve the addition or modification of plant equipment (no new or different type of equipment will be installed) nor do they alter the design or operation of any plant systems. No new accident scenarios, accident or transient initiators or precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed changes do not cause the malfunction of safety-related equipment assumed to be operable in accident analyses. No new or different mode of failure has been created and no new or different equipment performance requirements are imposed for accident mitigation. As such, the proposed changes have no effect on previously evaluated accidents. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. The proposed changes do not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. 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: 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, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.1.8, ``PHYSICS TESTS Exceptions,'' to allow the numbers of channels required by the Limiting Condition for Operation (LCO) section of TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' to be reduced from ``4'' to ``3'' to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-315-A, Revision 0, ``Reduce Plant Trips Due to Spurious Signals to the Nuclear Instrumentation System (NIS) During Physics Testing.'' 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 revise TS 3.1.8, ``PHYSICS TESTS Exceptions,'' to allow the number of channels required by LCO 3.3.1, ``RTS Instrumentation,'' to be reduced from ``4'' to ``3'', to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. A reduction in the number of required nuclear instrumentation channels is not an initiator to any accident previously evaluated. With the nuclear instrumentation channel placed in bypass instead of in trip, reactor protection is still provided by the nuclear instrumentation system operating in a two-out-of-three channel logic. As a result, the ability to mitigate any accident previously evaluated is not significantly affected. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. The proposed changes reduce the probability of a spurious reactor trip during physics testing. The reactor trip system continues to be capable of protecting the reactor utilizing the power range neutron flux trips operating in a two-out-of-three trip logic. As a result, the reactor is protected and the probability of a spurious reactor trip is significantly 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: 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, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to expand the TS 3.7.5 Limiting [[Page 23622]] Condition for Operation, Condition A to include the situation when one turbine driven AFW pump is operable in MODE 3, immediately following a refueling outage (if MODE 2 has not been entered), with a 7-day Completion Time. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-340-A, Revision 3, ``Allow 7 Day Completion Time for a Turbine-Driven AFW Pump Inoperable.'' 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 revise TS 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to allow a 7 day Completion Time to restore an inoperable AFW turbine-driven pump in MODE 3 immediately following a refueling outage, if MODE 2 has not been entered. An inoperable AFW turbine-driven pump is not an initiator of any accident previously evaluated. The ability of the plant to mitigate an accident is no different while in the extended Completion Time than during the existing Completion Time. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. The proposed changes revise TS 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to allow a 7 day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3, immediately following a refueling outage, if Mode 2 has not been entered. In Mode 3 immediately following a refueling outage, core decay heat is low and the need for AFW is also diminished. The two operable motor driven AFW pumps are available and there are alternate means of decay heat removal if needed. As a result, the risk presented by the extended Completion Time is minimal. 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: 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 (MNS), Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS 3.4.12, ``Low Temperature Overpressure Protection (LTOP) System,'' and TS 3.7.4, ``Steam Generator Power Operated Relief Valves (SG PORVs),'' to revise the Completion Times for Limiting Condition for Operation (LCO) 3.4.10 Required Action B.2, and LCO 3.7.4 Required Action C.2 from 12 to 24 hours and LCO 3.4.12 Required Action G.1 from 8 to 12 hours. The proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-352-A, Revision 1, ``Provide Consistent Completion Time to Reach MODE 4.'' 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 allow a more reasonable time to plan and execute required actions, and will 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 proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the MNS Updated Final Safety Analysis Report (UFSAR). The applicable radiological dose acceptance criteria will continue to be met. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. There are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC performs its safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed changes will not alter any assumptions made in the safety analyses. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment. Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The proposed changes will not have any impact on these barriers. No accident mitigating equipment will be adversely impacted. Therefore, existing safety margins will be preserved. None of the proposed changes will 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 [[Page 23623]] 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 (MNS), Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069. Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.6, ``Residual Heat Removal (RHR) and Coolant Circulation--Low Water Level,'' to add Note 1 to the Limiting Condition for Operation (LCO) Section of TS 3.9.6 to allow the securing of the operating train of RHR for up to 15 minutes to support switching operating trains. The allowance is restricted to three conditions: (a) The core outlet temperature is maintained greater than 10 degrees Fahrenheit below saturation temperature; (b) no operations are permitted that would cause an introduction of coolant into the Reactor Coolant System (RCS) with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1; and (c) no draining operations to further reduce RCS water volume are permitted. Additionally, the amendments would modify the LCO Section of TS 3.9.6 to add Note 2 which would allow one required RHR loop to be inoperable for up to 2 hours for surveillance testing, provided that the other RHR loop is operable and in operation. These proposed changes are consistent with Technical Specification Task Force (TSTF) Travelers TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation,'' TSTF-361-A, Revision 2, ``Allow Standby SDC [Shutdown Cooling]/RHR/DHR [Decay Heat Removal] Loop to be Inoperable to Support Testing,'' and TSTF-438-A, Revision 0, ``Clarify Exception Notes to be Consistent with the Requirement Being Excepted.'' 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 add two notes to MNS TS LCO 3.9.6. Note 1 would allow securing the operating train of Residual Heat Removal (RHR) for up to 15 minutes to support switching operating trains, subject to certain restrictions. Note 2 to would allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is Operable and in operation. These provisions are operational allowances. Neither operational allowance is an initiator to any accident previously evaluated. In addition, the proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any 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 do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 the margin of safety? Response: No. An operational allowance is proposed which would allow securing the operating train of RHR for up to 15 minutes to support switching operating trains, subject to certain restrictions. Considering these restrictions, combined with the short time frame allowed to swap operating RHR trains, and the ability to start an operating RHR train, if needed, the occurrence of an event that would require immediate operation of an RHR train is extremely remote. An operational allowance is also proposed which would allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation. A similar allowance currently appears in MNS TS 3.4.7, ``Reactor Coolant System (RCS) Loops--MODE 5, Loops Filled,'' and MNS TS 3.4.8, ``RCS Loops--MODE 5, Loops Not Filled,'' and the conditions under which the operational allowance would be applied in TS 3.9.6 are not significantly different from those specifications. This operational allowance provides the flexibility to perform surveillance testing, while ensuring that there is reasonable time for operators to respond to and mitigate any expected failures. The purpose of the RHR System is to remove decay and sensible heat from the Reactor Coolant System, to provide mixing of borated coolant, and to prevent boron stratification. Removal of system components from service as described above, and with limitations in place to maintain the ability of the RHR System to perform its safety function, does not significantly impact the margin of safety. Operators will continue to have adequate time to respond to any off-normal events. Removing the system from service, for a limited period of time, with other operational restrictions, limits the consequences to those already assumed in the Updated Final Safety Analysis Report (UFSAR). 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: 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 Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP), Van Buren County, Michigan Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17089A380. Description of amendment request: The proposed amendment would revise the PNP Cyber Security Plan (CSP) Milestone 8 full implementation date from December 15, 2017, to May 31, 2020. This amendment request is in support of PNP's transition, starting on October 1, 2018, from an operating power plant to a decommissioned plant. 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 changes to the CSP implementation schedule is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any [[Page 23624]] plant modifications which affect the performance capability of the structures, system, and components relied upon to mitigate the consequences of postulated accidents, and has no impact on 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. 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 to the CSP implementation schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed changes to the CSP implementation schedule is administrative in nature. In addition, the milestone date delay for full implementation of the CSP has no substantive impact because other measures have been taken which provide adequate protection during this period of time. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a 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: Jeanne Cho, Senior Counsel, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 10601. NRC Branch Chief: David J. Wrona. Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17101A608. Description of amendment request: The amendment would revise the renewed facility operating license Paragraph 3.G, ``Physical Protection.'' The amendment would revise the Pilgrim Nuclear Power Station Cyber Security Plan (CSP) implementation schedule for Milestone 8 full implementation date from December 15, 2017, to December 31, 2020. 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 to the CSP implementation schedule is administrative in nature. The change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents, and has no impact on 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. 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 to the CSP implementation schedule is administrative in nature. The proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents, and does not create the possibility of a new or different kind of accident from any accident previously evaluated. 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. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the CSP implementation schedule is administrative in nature. In addition, the milestone date delay for full implementation of the CSP has no substantive impact because other measures have been taken which provide adequate protection during this period of time. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Douglas A. Broaddus. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc. (the licensees), Docket Nos. 50-416 and 72-50, Grand Gulf Nuclear Station, Unit 1 (Grand Gulf), and Independent Spent Fuel Storage Installation (ISFSI), Claiborne County, Mississippi Date of amendment request: March 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17093A729. Description of amendment request: The proposed amendment would make an administrative change to the name of South Mississippi Electric Power Association, one of the licensees for Grand Gulf and its ISFSI. Effective November 10, 2016, South Mississippi Electric Power Association changed its corporate name from ``South Mississippi Electric Power Association'' to ``Cooperative Energy, a Mississippi Electric Cooperative.'' The corporate name was changed for commercial reasons. The changes proposed herein to the Grand Gulf operating license solely reflects the changed licensee name. This name change is purely administrative in nature. This request does not involve a transfer of control or of an interest in the license. [[Page 23625]] Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes ``involve a significant increase in the probability or consequences of an accident previously evaluated''? Response: No. The proposed amendments simply change the name of a licensee. The name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes ``create the possibility of a new or different kind of accident from any accident previously evaluated''? Response: No. The proposed amendments simply change the name of a licensee. The proposed name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed changes ``involve a significant reduction in the margin of safety''? Response: No. The proposed amendments simply change the name of a licensee. The name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected. 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 B. Glew, Jr., Associate General Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: December 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16364A338. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) for GGNS. The amendment would allow for a one cycle extension to the 10-year frequency of the GGNS containment integrated leakage rate test (ILRT) or Type A test and the drywell bypass leak rate test (DWBT). These tests are required by TS 5.5.12, ``10 CFR part 50, Appendix J [Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors], Testing Program,'' and TS Surveillance Requirement 3.6.5.1.1, respectively. The proposed change would permit the existing ILRT and DWBT frequency to be extended from 10 years to 11.5 years. 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 edits in [brackets]: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment to the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. Type B and C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type B and C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The assessment of the [leak- tightness] of the drywell will continue to be performed at least once each operating cycle. The proposed amendment will not change the leakage rate acceptance requirements. As such, the containment will continue to perform its design function as a barrier to fission product releases. In addition, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment and the assessment of the [leak-tightness] of the drywell exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. Therefore, the proposed change does not result in 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 amendment to the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. 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 the margin of safety? Response: No. The proposed amendment to the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS 10 CFR part 50, Appendix J, Testing Program for containment leak rate testing exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves the extension of the interval for only the Type A containment leakage rate test and the drywell bypass leakage rate test for GGNS. The proposed surveillance interval extension is bounded by the 15-year Type A test interval currently authorized within NEI 94-01, Revision 3-A. The design, operation, testing methods, and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met with the [[Page 23626]] acceptance of this proposed change, since these are not affected by the proposed changes to the Type A test interval. In addition to the scheduled performance of DWBT GGNS will continue to monitor the drywell for significant leakage during operation. 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 B. Glew, Jr., Associate General Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601. NRC Branch Chief: Robert J. Pascarelli. Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey Date of amendment request: April 10, 2017. A publicly-available version is available in ADAMS under Accession No. ML17100A844. Description of amendment request: The amendment would revise the OCNGS Cyber Security Plan (CSP) Milestone 8 (MS8) full implementation completion date, as set forth in the CSP implementation schedule, and revise the physical protection license condition in the renewed facility operating license. The licensee proposes to revise the CSP MS8 completion date from December 31, 2017, to August 31, 2021. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review 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 amendment request proposes a change to the OCNGS CSP MS8 completion date as set forth in the CSP implementation schedule and associated regulatory commitments. The NRC staff has concluded that the proposed change: (1) Does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected; (2) does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents; and (3) has no impact on the probability or consequences of an accident previously evaluated. In addition, the NRC staff has concluded that the proposed change to the CSP implementation schedule is administrative in nature. 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 NRC staff has concluded the proposed change: (1) Does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected; and (2) does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated. In addition, the NRC staff has concluded that the proposed change to the OCNGS CSP MS8 implementation schedule is administrative in nature. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The delay of the full implementation date for the OCNGS CSP MS8 has no substantive impact because other measures have been taken which provide adequate protection for the plant during this period of time. Therefore, the NRC staff has concluded that there is no significant reduction in a margin of safety. In addition, the NRC staff has concluded that the proposed change to the OCNGS CSP MS8 implementation schedule is administrative in nature. Therefore, the proposed change does not involve a significant reduction in the margin of safety. Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Douglas A. Broaddus. Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: March 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17087A012. Description of amendment request: The proposed changes would modify Technical Specifications (TS) Section 3.7.2, ``Steam Generator Stop Valves (SGSVs),'' to incorporate the SGSV actuator trains into the Limiting Condition for Operation and provide associated Conditions and Required Actions. In addition, Surveillance Requirement (SR) 3.7.2.2 would be revised to clearly identify that the SGSV actuator trains are required to be tested in accordance with the SR. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes provide requirements for SGSVs that have dual actuators which receive signals from separate instrumentation trains. The design and functional performance requirements, operational characteristics, and reliability of the SGSVs and actuator trains are unchanged. There is no impact on the design safety function of the SGSVs to close (as an accident mitigator), nor is there any change with respect to inadvertent closure of an SGSV (as a potential transient initiator). Since no failure mode or initiating condition that could cause an accident (including any plant transient) is created or affected, the change cannot involve a significant increase in the probability of an accident previously evaluated. With regard to the consequences of an accident and the equipment required for mitigation of the accident, the proposed changes involve no design or physical changes to the SGSVs or any other equipment required for accident mitigation. With respect to SGSV actuator train Completion Times, the consequences of an accident are independent of equipment Completion Times as long as adequate equipment availability is maintained. The proposed SGSV actuator Completion Times take into account the redundancy of the actuator trains and are limited in extent consistent with other Completion Times specified in the TS. Adequate equipment availability would therefore continue to be required by the TS. On this basis, the consequences of applicable, analyzed accidents are not significantly affected by the proposed changes. [[Page 23627]] Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to incorporate requirements for the SGSV actuator trains in TS 3.7.2 do not involve any design or physical changes to the facility, including the SGSVs and actuator trains themselves. No physical alteration of the plant is involved, as no new or different type of equipment is to be installed. The proposed changes do not alter any assumptions made in the safety analyses, nor do they involve any changes to plant procedures for ensuring that the plant is operated within analyzed limits. As such, no new failure modes or mechanisms that could cause a new or different kind of accident from any previously evaluated are being introduced. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes to incorporate requirements for the SGSV actuator trains do not alter the manner in which safety limits or limiting safety system settings are determined. No changes to instrument/system actuation setpoints are involved. The safety analysis acceptance criteria are not affected by this change and the proposed changes will not permit plant operation in a configuration outside the design basis. 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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: David J. Wrona. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa Date of amendment request: March 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17086A442. Description of amendment request: The proposed change would relocate cycle specific minimum critical power ratio (MCPR) values to the DAEC core operating limits report (COLR). The proposed amendment would revise the DAEC technical specifications (TS) to modify TS Table 3.3.2.1-1, ``Control Rod Block Instrumentation,'' Footnotes (a) through (e), and would relocate cycle specific MCPR values previously specified in TS Table 3.3.2.1-1, Footnotes (a) through (e) to TS 5.6.5(a)(4) by reference to the DAEC COLR. 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 is an administrative change that does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on, nor does it contribute in any way to the probability or consequences of transients or accidents. The COLR will continue to be controlled by the NextEra programs and procedures that comply with TS 5.6.5. Transient analyses addressed in the Final Safety Analysis Report will continue to be performed in the same manner with respect to changes in the cycle-dependent parameters obtained from the use of NRC-approved reload design methodologies, which ensures that the transient evaluation of new reloads are bounded by previously accepted analyses. Therefore, the proposed TS change does not involve an increase in the probability or consequences of a previously evaluated accident. 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 administrative change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on new or different kind of accidents. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The margin of safety is not affected by the relocation of cycle- specific Control Rod Block Instrumentation MCPR values from the TS to the COLR. Appropriate measures exist to control the values of these cycle-specific limits since it is required by TS that only NRC-approved methods be used to determine the limits. The proposed change continues to require operation within the core thermal limits as obtained from NRC-approved reload design methodologies and the actions to be taken if a limit is exceeded remain unchanged, again, in accordance with existing TS. Therefore, the proposed change has no impact to 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 Blair, P.O. Box 14000, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17093A688. Description of amendment request: The amendments would revise technical specification requirements to operate ventilation systems with charcoal filters from 10 hours to 15 minutes in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month.'' 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 replaces an existing Surveillance Requirement to operate the CREMAFS [Control Room Emergency Makeup Air and Filtration System], FSBEACS [Fuel Storage Building Emergency Air Cleaning System], and SBVS [Shield Building Ventilation System] equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP [Surveillance Frequency Control Program] with a requirement to operate the systems for 15 continuous minutes with heaters operating. [[Page 23628]] These systems are not accident initiators and therefore, these changes do not involve a significant increase in the probability of an accident. The proposed system and filter testing changes are consistent with current regulatory guidance for these systems and will continue to assure that these systems perform their design function which may include mitigating accidents. Thus, the change does not involve a significant increase in the consequences of an accident. Therefore, it is concluded that 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 replaces an existing Surveillance Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP with a requirement to operate the systems for 15 continuous minutes with heaters operating. The change proposed for these ventilation systems does not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are capable of performing their intended safety functions. The change does not create new failure modes or mechanisms and no new accident precursors are generated. Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change replaces an existing Surveillance Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP with a requirement to operate the systems for 15 continuous minutes with heaters operating. The design basis for the ventilation systems' heaters is to heat the incoming air which reduces the relative humidity. The heater testing change proposed will continue to demonstrate that the heaters are capable of heating the air and will perform their design function. The proposed change is consistent with regulatory guidance. Therefore, it is concluded that 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: William Blair, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408- 0420. NRC Branch Chief: James G. Danna. Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50- 306, Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP), Goodhue County, Minnesota Date of amendment request: March 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17094A565. Brief description of amendment request: The proposed amendments would revise the current emergency action levels (EAL) scheme used at PINGP to the EAL scheme contained in NEI 99-01, Revision 6, ``Development of Emergency Action Levels.'' 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 to the PINGP EAL scheme does not impact the physical function of plant structures, systems or components (SSC) or the manner in which the SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions. Therefore, the proposed change does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an event. The Emergency Plan, including the associated EALs, is implemented when an event occurs and cannot increase the probability of an accident. Further, the proposed change does not reduce the effectiveness of the Emergency Plan to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR part 50, Appendix E. Therefore, the proposed EAL scheme 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 involve any physical alteration to the plant, that is, no new or different type of equipment will be installed. The proposed change also does not change the method of plant operation and does not alter assumptions made in the safety analysis. Therefore, the proposed change will not create new failure modes or mechanisms that could result in a new or different kind of accident. The emergency plan, including the associated EAL scheme, is implemented when an event occurs and is not an accident initiator. Therefore, the proposed EAL scheme change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Margin of safety is provided by the ability of accident mitigation SCCs to perform at their analyzed capability. The change proposed in this license amendment request does not modify any plant equipment and there is no impact to the capability of the equipment to perform its intended accident mitigation function. The proposed change does not impact operation of the plant or its response to transients or accidents. Additionally, the proposed changes will not change any criteria used to establish safety limits or any safety system settings. The applicable requirements of 10 CFR 50.47 and 10 CFR part 50, Appendix E will continue to be met. Therefore, the proposed EAL scheme 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 requests involve 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. PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: March 27, 2017, as supplemented by letter dated April 28, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17086A364 and ML17118A092, respectively. Description of amendment request: The amendment would amend the Hope Creek Generating Station (Hope Creek) Technical Specifications (TSs) to revise and relocate the pressure-temperature (P-T) limit curves to a licensee-controlled pressure and temperature limits report (PTLR). The request was submitted in accordance with guidance provided in NRC Generic Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protections System Limits,'' dated January 31, 1996. 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 23629]] 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed license amendment adopts the NRC approved methodology described in Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044), ``Pressure Temperature Limits Report Methodology for Boiling Water Reactors.'' The Hope Creek PTLR was developed based on the methodology and template provided in the BWROG LTR. 10 CFR part 50, Appendix G establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Implementing this NRC approved methodology does not reduce the ability to protect the RCPB as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions. 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, 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 change in methodology for calculating P-T limits and the relocation of those limits to the PTLR do not alter or involve any design basis accident initiators. RCPB integrity will continue to be maintained in accordance with 10 CFR part 50, Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner. 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 changes involve a significant reduction in a margin of safety? Response: No. The proposed changes do not affect the function of the RCPB or its response during plant transients. Calculating the Hope Creek P-T limits using the NRC approved SI methodology ensures adequate margins of safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation P-T limits for the RCPB are determined. There are no changes to the setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected. Therefore, it is concluded that 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: James G. Danna. South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: April 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17102B032. Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix C (and plant-specific Tier 1) and Updated Final Safety Analysis Report (UFSAR) Tier 2 that describe: (1) The inspection and analysis of, and specifies the maximum calculated flow resistance acceptance criteria for, the fourth-stage automatic depressurization system loops; (2) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the allowable flow resistance acceptance criteria for, the in-containment refueling water storage tank (IRWST) injection line; (3) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the maximum flow resistance acceptance criteria for, the containment recirculation line; (4) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that specifies acceptance criteria for the maximum flow resistance between the IRWST drain line and the containment; and 5) removes licensing basis text from UFSAR Tier 2 that discusses the operation of swing check valves in current operating 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes do not adversely affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. The proposed changes do not adversely affect the physical design and operation of the in-containment refueling water storage tank (IRWST) injection, drain, containment recirculation, or fourth-stage automatic depressurization system (ADS) valves, including as-installed inspections and maintenance requirements as described in the Updated Final Safety Analysis Report (UFSAR). Inadvertent operation or failure of the fourth-stage ADS valves are considered as an accident initiator or part of an initiating sequence of events for an accident previously evaluated. However, the proposed change to the test methodology and calculated flow resistance for the fourth-stage ADS lines does not adversely affect the probability of inadvertent operation or failure. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected. The proposed changes do not adversely affect the ability of IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The designs of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria. The proposed changes do not adversely affect the prevention and mitigation of other abnormal events, e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR 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 changes do not affect the operation of any systems or equipment that might 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 [[Page 23630]] not adversely affect the physical design and operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves, including as-installed inspections, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the IRWST injection, drain, containment recirculation, and fourth- stage ADS valves is not adversely affected. These proposed changes do not adversely 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. 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 margins. The proposed changes verify and maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while updating the acceptance criteria for verifying the design features necessary to ensure the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves perform the design functions required to meet the existing safety margins in the safety analyses. 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 adversely affect any design code function, design analysis, safety analysis input or result, or design/safety margin. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 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. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the 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. Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50- 529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS), Maricopa County, Arizona Date of amendment request: April 1, 2016, as supplemented by letters dated July 21, September 9, and October 26, 2016. Description of amendment request: The amendments revised the Technical Specifications (TSs) for PVNGS, by modifying the requirements regarding the degraded and loss of voltage relays that are planned to be modified to be more aligned with designs generally implemented in the industry. Specifically, the licensing basis for degraded voltage protection will be changed from reliance on a TS initial condition that ensures adequate post-trip voltage support of accident mitigation equipment to crediting automatic actuation of the degraded and loss of voltage relays to ensure proper equipment performance. Date of issuance: April 27, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: Unit 1-201, Unit 2-201, and Unit 3-201. A publicly- available version is in ADAMS under Accession No. ML17090A164; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Operating Licenses and TSs. Date of initial notice in Federal Register: May 24, 2016 (81 FR 32803). The supplements dated July 21, September 9, and October 26, 2016, 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 April 27, 2017. No significant hazards consideration comments received: No. Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324; Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina Duke Energy Progress, Inc., Docket No. 50-400; Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Wake County, North Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (MNS), 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 (ONS), Oconee County, South Carolina Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP), Darlington County, South Carolina Date of amendment request: June 23, 2016. Brief description of amendments: The amendments modified the technical specification (TS) requirements for unavailable barriers by adding Limiting [[Page 23631]] Condition for Operation (LCO) 3.0.9 to the TS for BSEP, ONS, and RNP. The same changes were added as LCO 3.0.10 to the TS for CNS and MNS. For HNP, TS requirements for unavailable barriers were modified by adding LCO 3.0.6 to the TS. The changes are consistent with Technical Specification Task Force Traveler (TSTF)-427, Revision 2, ``Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY,'' subject to stated variations. Date of issuance: April 26, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos: 274/302 (BSEP), 288/284 (CNS), 155 (HNP), 295/274 (MNS), 402/404/403 (ONS), and 251 (RNP). A publicly-available version is in ADAMS under Accession No. ML17066A374; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-71 and DPR-62 (BSEP), NPF-35 and NPF-52 (CNS), NPF-63 (HNP), NPF-9 and NPF-17 (MNS), DPR-38, DPR-47, DPR-55 (ONS), and DPR-23 (RNP): Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: August 16, 2016 (81 FR 54614). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 26, 2017. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina 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 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: September 27, 2016, as supplemented by letters dated November 22, 2016, and April 20, 2017. Description of amendment request: The amendments revised Technical Specification Surveillance Requirements to require operating ventilation systems with charcoal filters for 15 continuous minutes every 31 days or at a frequency controlled in accordance with the Surveillance Frequency Control Program. The amendments are consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month,'' as published in the Federal Register on September 20, 2012 (77 FR 58428), with variations due to plant-specific differences. Date of issuance: May 8, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: 275 (Unit 1) and 303 (Unit 2) for the Brunswick Steam Electric Plant; 289 (Unit 1) and 285 (Unit 2) for the Catawba Nuclear Station; 296 (Unit 1) and 275 (Unit 2) for the McGuire Nuclear Station; 156 (Unit 1) for the Shearon Harris Nuclear Power Plant; and 252 (Unit No. 2) for the H. B. Robinson Steam Electric Plant. A publicly-available version is in ADAMS under Accession No. ML17055A647; documents related to these amendments are listed in the Safety Evaluations enclosed with the amendments. Renewed Facility Operating License Nos. DPR-71 and DPR-62, for the Brunswick Steam Electric Plant, Units 1 and 2; NPF-35 and NPF-52, for the Catawba Nuclear Station, Units 1 and 2; NPF-9 and NPF-17, for the McGuire Nuclear Station, Units 1 and 2; NPF-63, for the Shearon Harris Nuclear Power Plant, Unit 1; and DPR-23, for the H. B. Robinson Steam Electric Plant, Unit No. 2: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: January 17, 2017 (82 FR 4929). The supplemental letter dated April 20, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluations of the amendments are contained in Safety Evaluations dated May 8, 2017. No significant hazards consideration comments received: No. Northern States Power Company--Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: February 10, 2016, as supplemented by letters dated October 10 and December 16, 2016; and January 31, February 7, February 16, and March 29, 2017. Brief description of amendment: The amendment revised Technical Specification 5.5.11, ``Primary Containment Leakage Rate Testing Program,'' to increase the containment integrated leakage rate test program Test A interval from 10 to 15 years. Date of issuance: April 25, 2017. Effective date: As of the date of issuance and shall be implemented prior to the startup from the 2017 refueling outage. Amendment No.: 193. A publicly-available version is in ADAMS under Accession No. ML17103A235; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-22: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: April 26, 2016 (81 FR 24663). The supplemental letters dated October 10 and December 16, 2016; and January 31, February 7, February 16, and March 29, 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 April 25, 2017. No significant hazards consideration comments received: No. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo County, California Date of application for amendments: June 17, 2015, as supplemented by letters dated August 31, October 22, November 2, November 6, and December 17, 2015; and February 1, February 10, April 21, June 9, September 15, October 6, and December 27, 2016. Brief description of amendments: The amendments revised the licensing bases to adopt the alternative source term (AST) as allowed by 10 CFR 50.67, ``Accident source term.'' The AST [[Page 23632]] methodology, as established in NRC Regulatory Guide 1.183, ``Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession No. ML003716792), is used to calculate the offsite and control room radiological consequences of postulated accidents for DCPP, Units 1 and 2. The amendments revised Technical Specification (TS) 1.1, ``Definitions,'' for the definition of Dose Equivalent I-131; TS 3.4.16, ``RCS [Reactor Coolant System] Specific Activity,'' to revise the noble gas activity limit; TS 3.6.3, ``Containment Isolation Valves,'' to require the 48-inch containment purge supply and exhaust valves to be sealed closed during Modes 1, 2, 3, and 4; TS 5.5.11, ``Ventilation Filter Testing Program (VFTP),'' to change the allowable methyl iodide penetration testing criteria for the auxiliary building system charcoal filter; TS 5.5.19, ``Control Room Habitability Program,'' to replace ``whole body or its equivalent to any part of the body,'' with ``Total Effective Dose Equivalent,'' which is the dose criteria specified in 10 CFR 50.67, and Appendix D, ``Additional Conditions,'' for Facility Operating License Nos. DPR-80 and DPR-82 for DCPP, Units 1 and 2, to add additional license conditions. Date of issuance: April 27, 2017. Effective date: As of its date of issuance and shall be implemented within 365 days from the date of issuance. Amendment Nos.: Unit 1-230; Unit 2-232. A publicly-available version is in ADAMS under Accession No. ML17012A246; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: The license amendment request was originally noticed in the Federal Register on October 13, 2015 (80 FR 61486). As a result of the supplemental letters dated October 22, November 2, November 6, and December 17, 2015; and February 1, February 10, April 21, June 9, and September 15, 2016, the notice was reissued in its entirety to include the revised scope, description of the amendment request, and proposed no significant hazards consideration determination on November 8, 2016 (81 FR 78664). The supplemental letters dated October 6 and December 27, 2016, 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 April 27, 2017. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of amendment request: June 7, 2016. Brief description of amendment: The amendment modified the Technical Specifications to allow the use of Component Cooling System (CCS) pump 2B-B to support Train 1B operability when the normally aligned CCS pump C-S is removed from service. Date of issuance: April 27, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 113. A publicly-available version is in ADAMS under Accession No. ML17081A263; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-90: Amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: September 13, 2016 (81 FR 62932). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 27, 2017. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 11th day of May 2017. For the Nuclear Regulatory Commission. Kathryn M. Brock, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2017-10570 Filed 5-22-17; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices 23615
SUMMARY: In accordance with the NUCLEAR REGULATORY SUPPLEMENTARY INFORMATION:
Federal Advisory Committee Act, as COMMISSION I. Obtaining Information and
amended, the National Aeronautics and [NRC–2017–0120] Submitting Comments
Space Administration (NASA)
announces a meeting of the Ad Hoc Biweekly Notice; Applications and A. Obtaining Information
Task Force on Big Data. This task force Amendments to Facility Operating Please refer to Docket ID NRC–2017–
reports to the NASA Advisory Council’s Licenses and Combined Licenses 0120, facility name, unit number(s),
Science Committee. The meeting will be Involving No Significant Hazards plant docket number, application date,
held for the purpose of soliciting and Considerations and subject when contacting the NRC
discussing, from the scientific AGENCY: Nuclear Regulatory about the availability of information for
community and other persons, scientific Commission. this action. You may obtain publicly-
and technical information relevant to available information related to this
ACTION: Biweekly notice.
big data. action by any of the following methods:
SUMMARY: Pursuant to Section 189a.(2) • Federal rulemaking Web site: Go to
DATES: Thursday, June 22, 2017, 11:00 http://www.regulations.gov and search
of the Atomic Energy Act of 1954, as
a.m.–6:00 p.m., and Friday, June 23, for Docket ID NRC–2017–0120.
amended (the Act), the U.S. Nuclear
2017, 11:00 a.m.–6:00 p.m., Eastern Regulatory Commission (NRC) is • NRC’s Agencywide Documents
Daylight Time. (EDT). publishing this regular biweekly notice. Access and Management System
FOR FURTHER INFORMATION CONTACT: Ms. The Act requires the Commission to (ADAMS): You may obtain publicly-
Karshelia Henderson, Science Mission publish notice of any amendments available documents online in the
Directorate, NASA Headquarters, issued, or proposed to be issued, and ADAMS Public Documents collection at
Washington, DC 20546, (202) 358–2355, grants the Commission the authority to http://www.nrc.gov/reading-rm/
issue and make immediately effective adams.html. To begin the search, select
fax (202) 358–2779, or khenderson@
any amendment to an operating license ‘‘ADAMS Public Documents’’ and then
nasa.gov.
or combined license, as applicable, select ‘‘Begin Web-based ADAMS
SUPPLEMENTARY INFORMATION: The upon a determination by the Search.’’ For problems with ADAMS,
meeting will be open to the public Commission that such amendment please contact the NRC’s Public
telephonically and via WebEx. You involves no significant hazards Document Room (PDR) reference staff at
must use a touch tone phone to consideration, notwithstanding the 1–800–397–4209, 301–415–4737, or by
participate in this meeting. Any pendency before the Commission of a email to pdr.resource@nrc.gov. The
interested person may call the USA toll request for a hearing from any person. ADAMS accession number for each
This biweekly notice includes all document referenced (if it is available in
free conference call number 1–888–324–
notices of amendments issued, or ADAMS) is provided the first time that
9653 or toll number 1–312–470–7237, proposed to be issued, from April 25, it is mentioned in this document.
passcode 3883300 followed by the # 2017, to May 8, 2017. The last biweekly • NRC’s PDR: You may examine and
sign, to participate in this meeting by notice was published on May 9, 2017. purchase copies of public documents at
telephone on both days. The WebEx link DATES: Comments must be filed by June the NRC’s PDR, Room O1–F21, One
is https://nasa.webex.com/; the meeting 22, 2017. A request for a hearing must White Flint North, 11555 Rockville
number is 991 071 373 and the be filed by July 24, 2017. Pike, Rockville, Maryland 20852.
password is BDTFmtg#5 (case ADDRESSES: You may submit comments
sensitive). The agenda for the meeting B. Submitting Comments
by any of the following methods:
includes the following topics: • Federal Rulemaking Web site: Go to Please include Docket ID NRC–2017–
http://www.regulations.gov and search 0120, facility name, unit number(s),
—NASA Data Science Program
for Docket ID NRC–2017–0120. Address plant docket number, application date,
—NASA Science Mission Directorate questions about NRC dockets to Carol and subject in your comment
Data Archives Assessment Gallagher; telephone: 301–415–3463; submission.
—NASA’s Participation in Federal Big email: Carol.Gallagher@nrc.gov. For The NRC cautions you not to include
Data Initiatives technical questions, contact the identifying or contact information that
individual listed in the FOR FURTHER you do not want to be publicly
It is imperative that the meeting be INFORMATION CONTACT section of this disclosed in your comment submission.
held on these dates to accommodate the document. The NRC will post all comment
scheduling priorities of the key • Mail comments to: Cindy Bladey, submissions at http://
participants. Office of Administration, Mail Stop: T– www.regulations.gov as well as enter the
Patricia D. Rausch,
8–D36M, U.S. Nuclear Regulatory comment submissions into ADAMS.
Commission, Washington, DC 20555– The NRC does not routinely edit
Advisory Committee Management Officer, 0001. comment submissions to remove
National Aeronautics and Space For additional direction on obtaining identifying or contact information.
Administration. information and submitting comments, If you are requesting or aggregating
[FR Doc. 2017–10499 Filed 5–22–17; 8:45 am] see ‘‘Obtaining Information and comments from other persons for
BILLING CODE 7510–13–P Submitting Comments’’ in the submission to the NRC, then you should
SUPPLEMENTARY INFORMATION section of inform those persons not to include
sradovich on DSK3GMQ082PROD with NOTICES
this document. identifying or contact information that
FOR FURTHER INFORMATION CONTACT: they do not want to be publicly
Paula Blechman, Office of Nuclear disclosed in their comment submission.
Reactor Regulation, U.S. Nuclear Your request should state that the NRC
Regulatory Commission, Washington, does not routinely edit comment
DC 20555–0001; telephone: 301–415– submissions to remove such information
2242, email: Paula.Blechman@nrc.gov. before making the comment
VerDate Sep<11>2014 21:15 May 22, 2017 Jkt 241001 PO 00000 Frm 00095 Fmt 4703 Sfmt 4703 E:\FR\FM\23MYN1.SGM 23MYN1](https://thefederalregister.org/doc/82_FR_23713/page/1.svg)
![23618 Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices
additional direction on obtaining event that Residual Heat Removal Duke Energy Carolinas, LLC, Docket
information related to this document, requirements are not met during refueling Nos. 50–369 and 50–370, McGuire
see the ‘‘Obtaining Information and operations. Containment closure would be Nuclear Station, Units 1 and 2,
Submitting Comments’’ section of this appropriate for mitigation of a loss of Mecklenburg County, North Carolina
document. shutdown cooling accident, but it does not
affect the initiation of the accident. The Date of amendment request: January
Duke Energy Carolinas, LLC, Docket containment purge system isolation valves 11, 2017. A publicly-available version is
Nos. 50–413 and 50–414, Catawba will be capable of being closed automatically in ADAMS under Accession No.
Nuclear Station (CNS), Units 1 and 2, on a high containment radiation signal, such ML17025A069.
York County, South Carolina that there will be no significant increase in Description of amendment request:
the radiological consequences of a loss of The amendments would modify
Date of amendment request:
shutdown cooling. Technical Specification 3.1.2, ‘‘Core
December 15, 2016. A publicly-available
Therefore, the proposed changes do not Reactivity,’’ to revise the Completion
version is in ADAMS under Accession involve a significant increase in the Times of Required Action A.1 and A.2
No. ML16350A422. probability or consequences of any accident
Description of amendment request: from 72 hours to 7 days. This proposed
previously evaluated. change is consistent with Technical
The amendments would modify 2. Does the proposed amendment create
Technical Specification (TS) 3.9.4, Specification Task Force (TSTF)
the possibility of a new or different kind of Traveler TSTF–142–A, Revision 0,
‘‘Residual Heat Removal (RHR) and accident from any accident previously
Coolant Circulation—High Water ‘‘Increase the Completion Time when
evaluated? the Core Reactivity Balance is Not
Level,’’ and TS 3.9.5, ‘‘Residual Heat Response: No.
Removal (RHR) and Coolant Within Limit.’’
The proposed changes do not involve a
Circulation—Low Water Level.’’ Basis for proposed no significant
physical alteration to the plant (i.e., no new
Condition A of TS 3.9.4 applies when or different type of equipment will be hazards consideration determination:
RHR requirements are not met, and installed) or a change to the methods As required by 10 CFR 50.91(a), the
includes four required actions. Required governing normal plant operation. The licensee has provided its analysis of the
Action A.4 requires, within 4 hours, the containment purge system isolation valves issue of no significant hazards
closure of all containment penetrations will remain capable of being closed consideration, which is presented
providing direct access from automatically on a high containment below:
containment atmosphere to outside radiation signal. 1. Does the proposed amendment involve
Therefore, the proposed changes do not a significant increase in the probability or
atmosphere. The proposed changes
create the possibility of a new or different consequences of an accident previously
revise Required Action A.4 and add new kind of accident from any accident evaluated?
Required Actions A.5, A.6.1, and A.6.2 previously evaluated. Response: No.
to clarify that the intent of the required 3. Does the proposed amendment involve The proposed changes extend the
actions is to establish containment a significant reduction in the margin of Completion Time to take the Required
closure. Each of these required actions safety? Actions when measured core reactivity is not
will have a completion time of 4 hours. Response: No. within the specified limit of the predicted
Condition B of TS 3.9.5 applies when no Currently the Technical Specifications are values. The Completion Time to respond to
RHR loop is in operation, and includes vague and overly restrictive concerning the a difference between predicted and measured
three required actions. Required Action requirement for containment closure when core reactivity is not an initiator to any
B.3 requires the closure of all shutdown cooling is lost. The proposed accident previously evaluated. The
changes eliminate unclear requirements and radiological consequences of an accident
containment penetrations providing
provide a clear way to establish containment during the proposed Completion Time are no
direct access from containment different from the consequences of an
atmosphere to outside atmosphere. The closure that meets the [TS] Bases description,
which is to prevent radioactive gas from accident during the existing Completion
proposed changes are the same as the Time.
being released from the containment during
proposed changes to TS 3.9.4, consisting Therefore, the proposed changes do not
a loss of shutdown cooling incident. The
of a revision to Required Action B.3 and involved a significant increase in the
containment purge system isolation valves
the addition of new Required Actions probability or consequences of any accident
will remain capable of being closed previously evaluated.
B.4, B.5.1, and B.5.2. These proposed automatically on a high containment
changes are consistent with Technical 2. Does the proposed amendment create
radiation signal. the possibility of a new or different kind of
Specification Task Force (TSTF) Therefore, the proposed changes do not accident from any accident previously
Traveler TSTF–197–A, Revision 2, involve a significant reduction in a margin of evaluated?
‘‘Require Containment Closure When safety. Response: No.
Shutdown Cooling Requirements Are The proposed changes do not involve a
Not Met.’’ The NRC staff has reviewed the physical alteration to the plant (i.e., no new
Basis for proposed no significant licensee’s analysis and, based on this or different type of equipment will be
hazards consideration determination: review, it appears that the three installed) or a change to the methods
As required by 10 CFR 50.91(a), the standards of 10 CFR 50.92(c) are governing normal plant operation. The
licensee has provided its analysis of the satisfied. Therefore, the NRC staff changes do not alter the assumptions made
issue of no significant hazards proposes to determine that the in the safety analysis.
amendment request involves no Therefore, the proposed changes do not
consideration, which is presented
create the possibility of a new or different
below: significant hazards consideration. kind of accident from any accident
sradovich on DSK3GMQ082PROD with NOTICES
1. Does the proposed amendment involve Attorney for licensee: Kate B. Nolan, previously evaluated.
a significant increase in the probability or Deputy General Counsel, Duke Energy 3. Does the proposed amendment involve
consequences of an accident previously Carolinas, LLC, 550 South Tryon a significant reduction in the margin of
evaluated? Street—DEC45A, Charlotte, NC 28202– safety?
Response: No. 1802. Response: No.
The proposed changes revise the CNS TS The proposed changes provide additional
to ensure that the appropriate actions are NRC Branch Chief: Michael T. time to investigate and to implement
taken to establish containment closure in the Markley. appropriate operating restrictions when
VerDate Sep<11>2014 21:15 May 22, 2017 Jkt 241001 PO 00000 Frm 00098 Fmt 4703 Sfmt 4703 E:\FR\FM\23MYN1.SGM 23MYN1](https://thefederalregister.org/doc/82_FR_23713/page/4.svg)
![Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices 23619
measured core reactivity is not within the changes are consistent with Technical Therefore, the proposed changes do not
specified limit of the predicted values. The Specification Task Force (TSTF) involve a significant reduction in a margin of
additional time will not have a significant Traveler TSTF–197–A, Revision 2, safety.
effect on plant safety due to the
conservatisms used in designing the reactor
‘‘Require Containment Closure When The NRC staff has reviewed the
core and performing the safety analyses, and Shutdown Cooling Requirements Are licensee’s analysis and, based on this
the low probability of an accident or Not Met.’’ review, it appears that the three
transient which would approach the core Basis for proposed no significant standards of 10 CFR 50.92(c) are
design limits during the additional time. hazards consideration determination: satisfied. Therefore, the NRC staff
Therefore, the proposed changes do not proposes to determine that the
involve a significant reduction in a margin of As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the amendment request involves no
safety.
issue of no significant hazards significant hazards consideration.
The NRC staff has reviewed the consideration, which is presented Attorney for licensee: Kate B. Nolan,
licensee’s analysis and, based on this below: Deputy General Counsel, Duke Energy
review, it appears that the three Carolinas, LLC, 550 South Tryon
standards of 10 CFR 50.92(c) are 1. Does the proposed amendment involve Street—DEC45A, Charlotte, NC 28202–
satisfied. Therefore, the NRC staff a significant increase in the probability or
consequences of an accident previously
1802.
proposes to determine that the NRC Branch Chief: Michael T.
evaluated?
amendment request involves no Response: No. Markley.
significant hazards consideration. The proposed changes revise the MNS TS
Attorney for licensee: Kate B. Nolan, Duke Energy Carolinas, LLC, Docket
to ensure that the appropriate actions are
Deputy General Counsel, Duke Energy Nos. 50–369 and 50–370, McGuire
taken to establish containment closure in the
Corporation, LLC, 550 South Tryon event that Residual Heat Removal
Nuclear Station (MNS), Units 1 and 2,
Street—DEC45A, Charlotte, NC 28202– requirements are not met during refueling Mecklenburg County, North Carolina
1802. operations. Containment closure would be Date of amendment request: January
NRC Branch Chief: Michael T. appropriate for mitigation of a loss of 11, 2017. A publicly-available version is
Markley. shutdown cooling accident, but it does not in ADAMS under Accession No.
affect the initiation of the accident. The
Duke Energy Carolinas, LLC, Docket ML17025A069.
containment purge system isolation valves
Nos. 50–369 and 50–370, McGuire will be capable of being closed automatically
Description of amendment request:
Nuclear Station (MNS), Units 1 and 2, on a high containment radiation signal, such The amendments would modify
Mecklenburg County, North Carolina that there will be no significant increase in Technical Specification (TS) 3.6.3,
the radiological consequences of a loss of ‘‘Containment Isolation Valves,’’ to add
Date of amendment request: January a Note to TS Limiting Condition for
shutdown cooling.
11, 2017. A publicly-available version is Therefore, the proposed changes do not Operation (LCO) 3.6.3 Required Actions
in ADAMS under Accession No. involve a significant increase in the A.2, C.2 and E.2 to allow isolation
ML17025A069. probability or consequences of any accident devices that are locked, sealed or
Description of amendment request: previously evaluated.
The amendments would modify otherwise secured to be verified by use
2. Does the proposed amendment create of administrative means. This proposed
Technical Specification (TS) 3.9.5, the possibility of a new or different kind of
‘‘Residual Heat Removal (RHR) and change is consistent with Technical
accident from any accident previously
Coolant Circulation—High Water evaluated? Specification Task Force (TSTF)
Level,’’ and TS 3.9.6, ‘‘Residual Heat Response: No. Traveler TSTF–269–A, Revision 2,
Removal (RHR) and Coolant The proposed changes do not involve a ‘‘Allow Administrative Means of
Circulation—Low Water Level.’’ physical alteration to the plant (i.e., no new Position Verification for Locked or
Condition A of TS 3.9.5 applies when or different type of equipment will be Sealed Valves.’’
RHR requirements are not met, and
installed) or a change to the methods Basis for proposed no significant
governing normal plant operation. The hazards consideration determination:
includes four required actions. Required containment purge system isolation valves
Action A.4 requires, within 4 hours, the As required by 10 CFR 50.91(a), the
will remain capable of being closed licensee has provided its analysis of the
closure of all containment penetrations automatically on a high containment
providing direct access from issue of no significant hazards
radiation signal.
containment atmosphere to outside Therefore, the proposed changes do not consideration, which is presented
atmosphere. The proposed changes create the possibility of a new or different below:
revise Required Action A.4 and add new kind of accident from any accident 1. Does the proposed amendment involve
Required Actions A.5, A.6.1, and A.6.2 previously evaluated. a significant increase in the probability or
to clarify that the intent of the required 3. Does the proposed amendment involve consequences of an accident previously
a significant reduction in the margin of evaluated?
actions is to establish containment
safety? Response: No.
closure. Each of these required actions Response: No. The proposed changes modify MNS TS
will have a completion time of 4 hours. Currently the Technical Specifications are 3.6.3, ‘‘Containment Isolation Valves’’. This
Condition B of TS 3.9.6 applies when no vague and overly restrictive concerning the TS currently includes actions that require
RHR loop is in operation, and includes requirement for containment closure when penetrations to be isolated and periodically
three required actions. Required Action shutdown cooling is lost. The proposed verified to be isolated. A Note is proposed to
B.3 requires the closure of all changes eliminate unclear requirements and be added to TS 3.6.3 Required Actions A.2,
containment penetrations providing provide a clear way to establish containment C.2, and E.2, to allow isolation devices that
sradovich on DSK3GMQ082PROD with NOTICES
direct access from containment closure that meets the [TS] Bases description, are locked, sealed, or otherwise secured to be
which is to prevent radioactive gas from verified by use of administrative means. The
atmosphere to outside atmosphere. The
being released from the containment during proposed changes do not affect any plant
proposed changes are the same as the a loss of shutdown cooling incident. The equipment, test methods, or plant operation,
proposed changes to TS 3.9.5, consisting containment purge system isolation valves and is not an initiator of any analyzed
of a revision to Required Action B.3 and will remain capable of being closed accident sequence. The inoperable
the addition of new Required Actions automatically on a high containment containment penetrations will continue to be
B.4, B.5.1, and B.5.2. These proposed radiation signal. isolated, and hence perform their isolation
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![23620 Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices
function. Operation in accordance with the (TSTF) Traveler TSTF–283–A, Revision the time that a DG is unavailable because
proposed TSs will ensure that all analyzed 3, ‘‘Modify Section 3.8 Mode Restriction testing to reestablish operability can be
accidents will continue to be mitigated as Notes.’’ performed without a plant shutdown. The
previously analyzed. Basis for proposed no significant proposed changes also require an assessment
Therefore, the proposed changes do not to verify that plant safety will be maintained
hazards consideration determination:
involve a significant increase in the or enhanced by performance of the
probability or consequences of any accident As required by 10 CFR 50.91(a), the Surveillance in the current prohibited
previously evaluated. licensee has provided its analysis of the Modes.
2. Does the proposed amendment create issue of no significant hazards Therefore, the proposed changes do not
the possibility of a new or different kind of consideration, which is presented involve a significant reduction in a margin of
accident from any accident previously below: safety.
evaluated?
Response: No.
1. Does the proposed amendment involve The NRC staff has reviewed the
a significant increase in the probability or licensee’s analysis and, based on this
The proposed changes do not involve a consequences of an accident previously
physical alteration to the plant (i.e., no new review, it appears that the three
evaluated? standards of 10 CFR 50.92(c) are
or different type of equipment will be Response: No.
installed) or a change to the methods The proposed changes modify Mode satisfied. Therefore, the NRC staff
governing normal plant operation. The restriction Notes in TS SRs 3.8.1.8, 3.8.1.11, proposes to determine that the
changes do not alter the assumptions made 3.8.1.16, 3.8.1.17, and 3.8.1.19 to allow amendment request involves no
in the safety analysis. performance of the Surveillance in whole or significant hazards consideration.
Therefore, the proposed changes do not in part to reestablish Diesel Generator (DG) Attorney for licensee: Kate B. Nolan,
create the possibility of a new or different Operability, and to allow the crediting of Deputy General Counsel, Duke Energy
kind of accident from any accident unplanned events that satisfy the
previously evaluated. Carolinas, LLC, 550 South Tryon
Surveillance(s) [Requirements]. The Street—DEC45A, Charlotte, NC 28202–
3. Does the proposed amendment involve emergency diesel generators and their
a significant reduction in the margin of associated emergency loads are accident
1802.
safety? mitigating features, and are not an initiator of NRC Branch Chief: Michael T.
Response: No. any accident previously evaluated. As a Markley.
The proposed changes will not affect the result, the probability of any accident
operation of plant equipment or the function Duke Energy Carolinas, LLC, Docket
previously evaluated is not significantly
of any equipment assumed in the accident increased. To manage any increase in risk,
Nos. 50–369 and 50–370, McGuire
analysis. Affected containment penetrations the proposed changes require an assessment Nuclear Station, Units 1 and 2,
will continue to be isolated as required by to verify that plant safety will be maintained Mecklenburg County, North Carolina
the existing TS. or enhanced by performance of the
Therefore, the proposed changes do not
Date of amendment request: January
Surveillance in the current prohibited 11, 2017. A publicly-available version is
involve a significant reduction in a margin of Modes. The radiological consequences of an
safety. accident previously evaluated during the in ADAMS under Accession No.
period that the DG is being tested to ML17025A069.
The NRC staff has reviewed the Description of amendment request:
licensee’s analysis and, based on this reestablish operability are no different from
the radiological consequences of an accident The amendments would modify
review, it appears that the three previously evaluated while the DG is Technical Specification (TS) 3.4.12,
standards of 10 CFR 50.92(c) are inoperable. As a result, the consequences of ‘‘Low Temperature Overpressure
satisfied. Therefore, the NRC staff any accident previously evaluated are not Protection (LTOP) System,’’ to increase
proposes to determine that the increased. the time allowed for swapping charging
amendment request involves no Therefore, the proposed changes do not
pumps to 1 hour. Additionally, an
significant hazards consideration. involve a significant increase in the
probability or consequences of any accident existing note in the Applicability
Attorney for licensee: Kate B. Nolan,
previously evaluated. section of TS 3.4.12 is being reworded
Deputy General Counsel, Duke Energy
2. Does the proposed amendment create and relocated to the Limiting Condition
Carolinas, LLC, 550 South Tryon
the possibility of a new or different kind of for Operation section of TS 3.4.12 as
Street—DEC45A, Charlotte, NC 28202– accident from any accident previously Note 2. These proposed changes are
1802. evaluated?
NRC Branch Chief: Michael T. consistent with Technical Specification
Response: No. Task Force (TSTF) Traveler TSTF–285–
Markley. The proposed changes do not involve a
physical alteration to the plant (i.e., no new
A, Revision 1, ‘‘Charging Pump Swap
Duke Energy Carolinas, LLC, Docket or different type of equipment will be LTOP Allowance.’’
Nos. 50–369 and 50–370, McGuire installed) or a change to the methods Basis for proposed no significant
Nuclear Station, Units 1 and 2, governing normal plant operation. The hazards consideration determination:
Mecklenburg County, North Carolina changes do not alter the assumptions made As required by 10 CFR 50.91(a), the
Date of amendment request: January in the safety analysis. licensee has provided its analysis of the
Therefore, the proposed changes do not issue of no significant hazards
11, 2017. A publicly-available version is create the possibility of a new or different
in ADAMS under Accession No. consideration, which is presented
kind of accident from any accident
ML17025A069. below:
previously evaluated.
Description of amendment request: 3. Does the proposed amendment involve 1. Does the proposed amendment involve
The amendments would modify a significant reduction in the margin of a significant increase in the probability or
Technical Specification (TS) 3.8.1, ‘‘AC safety? consequences of an accident previously
[Alternating Current] Sources— Response: No. evaluated?
sradovich on DSK3GMQ082PROD with NOTICES
Operating,’’ to allow greater flexibility The purpose of Surveillances is to verify Response: No.
in performing Surveillance that equipment is capable of performing its The proposed changes increase the time
assumed safety function. The proposed allowed for swapping charging pumps from
Requirements (SRs) by modifying Mode changes will only allow the performance of 15 minutes to one hour, and make several
restriction notes in TS SRs 3.8.1.8, the Surveillances to reestablish operability, other associated administrative changes and
3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. and the proposed changes may not be used clarifications to the TS. These changes do not
This proposed change is consistent with to remove a DG from service. In addition, the affect event initiators or precursors. Thus, the
Technical Specification Task Force proposed changes will potentially shorten proposed changes do not involve a
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![Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices 23623
amendment request involves no 1. Does the proposed amendment involve Reactor Coolant System, to provide mixing of
significant hazards consideration. a significant increase in the probability or borated coolant, and to prevent boron
Attorney for licensee: Kate B. Nolan, consequences of an accident previously stratification. Removal of system components
evaluated? from service as described above, and with
Deputy General Counsel, Duke Energy Response: No. limitations in place to maintain the ability of
Carolinas, LLC, 550 South Tryon The proposed changes add two notes to the RHR System to perform its safety
Street—DEC45A, Charlotte, NC 28202– MNS TS LCO 3.9.6. Note 1 would allow function, does not significantly impact the
1802. securing the operating train of Residual Heat margin of safety. Operators will continue to
NRC Branch Chief: Michael T. Removal (RHR) for up to 15 minutes to have adequate time to respond to any off-
Markley. support switching operating trains, subject to normal events. Removing the system from
certain restrictions. Note 2 to would allow service, for a limited period of time, with
Duke Energy Carolinas, LLC, Docket one RHR loop to be inoperable for up to 2 other operational restrictions, limits the
Nos. 50–369 and 50–370, McGuire hours for surveillance testing provided the consequences to those already assumed in
Nuclear Station (MNS), Units 1 and 2, other RHR loop is Operable and in operation. the Updated Final Safety Analysis Report
Mecklenburg County, North Carolina These provisions are operational allowances. (UFSAR).
Neither operational allowance is an initiator Therefore, the proposed changes do not
Date of amendment request: January to any accident previously evaluated. In involve a significant reduction in the margin
11, 2017. A publicly-available version is addition, the proposed changes will not of safety.
in ADAMS under Accession No. affect the source term, containment isolation,
or radiological release assumptions used in The NRC staff has reviewed the
ML17025A069.
evaluating the radiological consequences of licensee’s analysis and, based on this
Description of amendment request: review, it appears that the three
any accident previously evaluated.
The amendments would modify Therefore, the proposed changes do not standards of 10 CFR 50.92(c) are
Technical Specification (TS) 3.9.6, involve a significant increase in the satisfied. Therefore, the NRC staff
‘‘Residual Heat Removal (RHR) and probability or consequences of any accident proposes to determine that the
Coolant Circulation—Low Water Level,’’ previously evaluated. amendment request involves no
to add Note 1 to the Limiting Condition 2. Does the proposed amendment create
significant hazards consideration.
for Operation (LCO) Section of TS 3.9.6 the possibility of a new or different kind of
accident from any accident previously
Attorney for licensee: Kate B. Nolan,
to allow the securing of the operating Deputy General Counsel, Duke Energy
train of RHR for up to 15 minutes to evaluated?
Response: No. Carolinas, LLC, 550 South Tryon
support switching operating trains. The Street—DEC45A, Charlotte, NC 28202–
The proposed changes do not involve a
allowance is restricted to three physical alteration to the plant (i.e., no new 1802.
conditions: (a) The core outlet or different type of equipment will be NRC Branch Chief: Michael T.
temperature is maintained greater than installed) or a change to the methods Markley.
10 degrees Fahrenheit below saturation governing normal plant operation. The
temperature; (b) no operations are changes do not alter the assumptions made Entergy Nuclear Operations, Inc.,
permitted that would cause an in the safety analysis. Docket No. 50–255, Palisades Nuclear
introduction of coolant into the Reactor Therefore, the proposed changes do not Plant (PNP), Van Buren County,
Coolant System (RCS) with boron create the possibility of a new or different Michigan
kind of accident from any accident
concentration less than that required to Date of amendment request: March
previously evaluated.
meet the minimum required boron 3. Does the proposed amendment involve 30, 2017. A publicly-available version is
concentration of LCO 3.9.1; and (c) no a significant reduction in the margin of in ADAMS under Accession No.
draining operations to further reduce safety? ML17089A380.
RCS water volume are permitted. Response: No. Description of amendment request:
Additionally, the amendments would An operational allowance is proposed The proposed amendment would revise
modify the LCO Section of TS 3.9.6 to which would allow securing the operating the PNP Cyber Security Plan (CSP)
add Note 2 which would allow one train of RHR for up to 15 minutes to support Milestone 8 full implementation date
required RHR loop to be inoperable for switching operating trains, subject to certain from December 15, 2017, to May 31,
restrictions. Considering these restrictions,
up to 2 hours for surveillance testing, 2020. This amendment request is in
combined with the short time frame allowed
provided that the other RHR loop is to swap operating RHR trains, and the ability support of PNP’s transition, starting on
operable and in operation. These to start an operating RHR train, if needed, the October 1, 2018, from an operating
proposed changes are consistent with occurrence of an event that would require power plant to a decommissioned plant.
Technical Specification Task Force immediate operation of an RHR train is Basis for proposed no significant
(TSTF) Travelers TSTF–349–A, extremely remote. hazards consideration determination:
Revision 1, ‘‘Add Note to LCO 3.9.5 An operational allowance is also proposed As required by 10 CFR 50.91(a), the
Allowing Shutdown Cooling Loops which would allow one RHR loop to be licensee has provided its analysis of the
Removal from Operation,’’ TSTF–361– inoperable for up to 2 hours for surveillance issue of no significant hazards
testing provided the other RHR loop is
A, Revision 2, ‘‘Allow Standby SDC consideration, which is presented
operable and in operation. A similar
[Shutdown Cooling]/RHR/DHR [Decay allowance currently appears in MNS TS below:
Heat Removal] Loop to be Inoperable to 3.4.7, ‘‘Reactor Coolant System (RCS) 1. Does the proposed change involve a
Support Testing,’’ and TSTF–438–A, Loops—MODE 5, Loops Filled,’’ and MNS TS significant increase in the probability or
Revision 0, ‘‘Clarify Exception Notes to 3.4.8, ‘‘RCS Loops—MODE 5, Loops Not consequences of an accident previously
be Consistent with the Requirement Filled,’’ and the conditions under which the evaluated?
Being Excepted.’’ operational allowance would be applied in Response: No.
sradovich on DSK3GMQ082PROD with NOTICES
Basis for proposed no significant TS 3.9.6 are not significantly different from The proposed changes to the CSP
those specifications. This operational implementation schedule is administrative in
hazards consideration determination:
allowance provides the flexibility to perform nature. This change does not alter accident
As required by 10 CFR 50.91(a), the surveillance testing, while ensuring that analysis assumptions, add any initiators, or
licensee has provided its analysis of the there is reasonable time for operators to affect the function of plant systems or the
issue of no significant hazards respond to and mitigate any expected manner in which systems are operated,
consideration, which is presented failures. The purpose of the RHR System is maintained, modified, tested, or inspected.
below: to remove decay and sensible heat from the The proposed change does not require any
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![Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices 23625
Basis for proposed no significant Attorney for licensee: William B. such, the containment will continue to
hazards consideration determination: Glew, Jr., Associate General Counsel— perform its design function as a barrier to
As required by 10 CFR 50.91(a), the Entergy Services, Inc., 440 Hamilton fission product releases. In addition, the
containment and the testing requirements
licensee has provided its analysis of the Avenue, White Plains, New York 10601.
invoked to periodically demonstrate the
issue of no significant hazards NRC Branch Chief: Robert J. integrity of the containment and the
consideration, which is presented Pascarelli. assessment of the [leak-tightness] of the
below: Entergy Operations, Inc., System Energy drywell exist to ensure the plant’s ability to
1. Do the proposed changes ‘‘involve a mitigate the consequences of an accident,
Resources, Inc., South Mississippi
significant increase in the probability or and do not involve the prevention or
Electric Power Association, and Entergy identification of any precursors of an
consequences of an accident previously
Mississippi, Inc., Docket No. 50–416, accident. Therefore, this proposed extension
evaluated’’?
Response: No. Grand Gulf Nuclear Station, Unit 1 does not involve a significant increase in the
The proposed amendments simply change (GGNS), Claiborne County, Mississippi probability of an accident previously
the name of a licensee. The name change is evaluated.
Date of amendment request: Therefore, the proposed change does not
purely administrative. None of the functions December 29, 2016. A publicly-available
or responsibility of any of the Grand Gulf result in a significant increase in the
version is in ADAMS under Accession probability or consequences of an accident
licensees will change as a result of the
amendments. The proposed amendments do No. ML16364A338. previously evaluated.
not alter the design, function, or operation of Description of amendment request: 2. Does the proposed change create the
any plant equipment. As such, the accident The proposed amendment would revise possibility of a new or different kind of
and transient analyses contained in the the Technical Specifications (TSs) for accident from any accident previously
facility updated final safety analysis report GGNS. The amendment would allow for evaluated?
will not be affected. a one cycle extension to the 10-year Response: No.
Therefore, the proposed changes do not The proposed amendment to the Technical
frequency of the GGNS containment Specifications (TS) involves the extension of
involve a significant increase in the integrated leakage rate test (ILRT) or
probability or consequences of an accident the Grand Gulf Nuclear Station, Unit 1
previously evaluated.
Type A test and the drywell bypass leak (GGNS) Type A integrated leakage rate test
2. Do the proposed changes ‘‘create the rate test (DWBT). These tests are and the drywell bypass leakage rate test
possibility of a new or different kind of required by TS 5.5.12, ‘‘10 CFR part 50, intervals to 11.5 years. The containment and
accident from any accident previously Appendix J [Primary Reactor the testing requirements to periodically
evaluated’’? Containment Leakage Testing for Water- demonstrate the integrity of the containment
Response: No. Cooled Power Reactors], Testing exist to ensure the plant’s ability to mitigate
The proposed amendments simply change Program,’’ and TS Surveillance the consequences of an accident do not
the name of a licensee. The proposed name involve any accident precursors or initiators.
Requirement 3.6.5.1.1, respectively. The The proposed change does not involve a
change is purely administrative. None of the
functions or responsibility of any of the
proposed change would permit the physical change to the plant (i.e., no new or
Grand Gulf licensees will change as a result existing ILRT and DWBT frequency to different type of equipment will be installed)
of the amendments. The proposed be extended from 10 years to 11.5 years. or a change to the manner in which the plant
amendments do not alter the design, Basis for proposed no significant is operated or controlled.
function, or operation of any plant hazards consideration determination: Therefore, the proposed change does not
equipment. As such, the accident and As required by 10 CFR 50.91(a), the create the possibility of a new or different
transient analyses contained in the facility licensee has provided its analysis of the kind of accident from any previously
updated final safety analysis report will not issue of no significant hazards evaluated.
be affected. 3. Does the proposed change involve a
consideration, which is presented
Therefore, the proposed changes do not significant reduction in the margin of safety?
create the possibility of a new or different below, with NRC edits in [brackets]: Response: No.
kind of accident from any accident 1. Does the proposed amendment involve The proposed amendment to the Technical
previously evaluated. a significant increase in the probability or Specifications (TS) involves the extension of
3. Do the proposed changes ‘‘involve a consequences of an accident previously the Grand Gulf Nuclear Station, Unit 1
significant reduction in the margin of evaluated? (GGNS) Type A integrated leakage rate test
safety’’? Response: No. and the drywell bypass leakage rate test
Response: No. The proposed amendment to the Technical intervals to 11.5 years. This amendment does
The proposed amendments simply change Specifications (TS) involves the extension of not alter the manner in which safety limits,
the name of a licensee. The name change is the Grand Gulf Nuclear Station, Unit 1 limiting safety system set points, or limiting
purely administrative. None of the functions (GGNS) Type A integrated leakage rate test conditions for operation are determined. The
or responsibility of any of the Grand Gulf and the drywell bypass leakage rate test specific requirements and conditions of the
licensees will change as a result of the intervals to 11.5 years. TS 10 CFR part 50, Appendix J, Testing
amendments. The proposed amendments do The proposed extension does not involve Program for containment leak rate testing
not alter the design, function, or operation of either a physical change to the plant or a exist to ensure that the degree of containment
any plant equipment. As such, the accident change in the manner in which the plant is structural integrity and leak-tightness that is
and transient analyses contained in the operated or controlled. The containment is considered in the plant safety analysis is
facility updated final safety analysis report designed to provide an essentially leak tight maintained. The overall containment leak
will not be affected. barrier against the uncontrolled release of rate limit specified by TS is maintained.
Therefore, the proposed changes do not radioactivity to the environment for The proposed change involves the
involve a significant reduction in the margin postulated accidents. Type B and C testing extension of the interval for only the Type A
of safety. ensures that individual containment isolation containment leakage rate test and the drywell
The NRC staff has reviewed the valves are essentially leak tight. In addition, bypass leakage rate test for GGNS. The
sradovich on DSK3GMQ082PROD with NOTICES
licensee’s analysis and, based on this aggregate Type B and C leakage rates support proposed surveillance interval extension is
the leakage tightness of primary containment bounded by the 15-year Type A test interval
review, it appears that the three
by minimizing potential leakage paths. The currently authorized within NEI 94–01,
standards of 10 CFR 50.92(c) are assessment of the [leak-tightness] of the Revision 3–A. The design, operation, testing
satisfied. Therefore, the NRC staff drywell will continue to be performed at methods, and acceptance criteria for Types
proposes to determine that the least once each operating cycle. The A, B, and C containment leakage tests
amendment request involves no proposed amendment will not change the specified in applicable codes and standards
significant hazards consideration. leakage rate acceptance requirements. As would continue to be met with the
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![Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices 23627
Therefore, the proposed changes do not modify TS Table 3.3.2.1–1, ‘‘Control limits since it is required by TS that only
involve a significant increase in the Rod Block Instrumentation,’’ Footnotes NRC-approved methods be used to determine
probability or consequences of an accident (a) through (e), and would relocate cycle the limits. The proposed change continues to
previously evaluated. require operation within the core thermal
specific MCPR values previously limits as obtained from NRC-approved reload
2. Does the proposed amendment create
the possibility of a new or different kind of specified in TS Table 3.3.2.1–1, design methodologies and the actions to be
accident from any accident previously Footnotes (a) through (e) to TS taken if a limit is exceeded remain
evaluated? 5.6.5(a)(4) by reference to the DAEC unchanged, again, in accordance with
Response: No. COLR. existing TS.
The proposed changes to incorporate Basis for proposed no significant Therefore, the proposed change has no
requirements for the SGSV actuator trains in hazards consideration determination: impact to the margin of safety.
TS 3.7.2 do not involve any design or As required by 10 CFR 50.91(a), the The NRC staff has reviewed the
physical changes to the facility, including the licensee has provided its analysis of the
SGSVs and actuator trains themselves. No
licensee’s analysis and, based on this
physical alteration of the plant is involved,
issue of no significant hazards review, it appears that the three
as no new or different type of equipment is consideration, which is presented standards of 10 CFR 50.92(c) are
to be installed. The proposed changes do not below: satisfied. Therefore, the NRC staff
alter any assumptions made in the safety 1. Does the proposed change involve a proposes to determine that the
analyses, nor do they involve any changes to significant increase in the probability or amendment request involves no
plant procedures for ensuring that the plant consequences of an accident previously significant hazards consideration.
is operated within analyzed limits. As such, evaluated? Attorney for licensee: William Blair,
no new failure modes or mechanisms that Response: No. P.O. Box 14000, Juno Beach, FL 33408–
could cause a new or different kind of The proposed change is an administrative 0420.
accident from any previously evaluated are change that does not affect any plant systems,
being introduced.
NRC Branch Chief: David J. Wrona.
structures, or components designed for the
Therefore, the proposed changes do not prevention or mitigation of previously NextEra Energy Seabrook, LLC, Docket
create the possibility of a new or different evaluated accidents. No new equipment is No. 50–443, Seabrook Station, Unit No.
kind of accident from any previously added nor is installed equipment being 1, Rockingham County, New Hampshire
evaluated. changed or operated in a different manner.
3. Does the proposed amendment involve Relocation of the Control Rod Block Florida Power & Light Company, et al.,
a significant reduction in a margin of safety? Instrumentation MCPR values to the COLR Docket Nos. 50–335 and 50–389, St.
Response: No. has no influence or impact on, nor does it Lucie Plant, Unit Nos. 1 and 2, St. Lucie
The proposed changes to incorporate contribute in any way to the probability or County, Florida
requirements for the SGSV actuator trains do consequences of transients or accidents. The
not alter the manner in which safety limits COLR will continue to be controlled by the Date of amendment request: March
or limiting safety system settings are NextEra programs and procedures that 30, 2017. A publicly-available version is
determined. No changes to instrument/ comply with TS 5.6.5. Transient analyses in ADAMS under Accession No.
system actuation setpoints are involved. The addressed in the Final Safety Analysis Report ML17093A688.
safety analysis acceptance criteria are not will continue to be performed in the same Description of amendment request:
affected by this change and the proposed manner with respect to changes in the cycle- The amendments would revise technical
changes will not permit plant operation in a dependent parameters obtained from the use specification requirements to operate
configuration outside the design basis. of NRC-approved reload design ventilation systems with charcoal filters
Therefore, the proposed changes do not methodologies, which ensures that the
involve a significant reduction in a margin of transient evaluation of new reloads are
from 10 hours to 15 minutes in
safety. bounded by previously accepted analyses. accordance with Technical
Therefore, the proposed TS change does Specifications Task Force (TSTF)
The NRC staff has reviewed the Traveler TSTF–522, Revision 0, ‘‘Revise
not involve an increase in the probability or
licensee’s analysis and, based on this consequences of a previously evaluated Ventilation System Surveillance
review, it appears that the three accident. Requirements to Operate for 10 hours
standards of 10 CFR 50.92(c) are 2. Does the proposed change create the per Month.’’
satisfied. Therefore, the NRC staff possibility of a new or different kind of Basis for proposed no significant
proposes to determine that the accident from any accident previously hazards consideration determination:
amendment request involves no evaluated? As required by 10 CFR 50.91(a), the
significant hazards consideration. Response: No.
The proposed administrative change does
licensee has provided its analysis of the
Attorney for licensee: Robert B. issue of no significant hazards
Haemer, Senior Nuclear Counsel, One not involve any changes to the operation,
testing, or maintenance of any safety-related, consideration, which is presented
Cook Place, Bridgman, MI 49106. below:
or otherwise important to safety systems. All
NRC Branch Chief: David J. Wrona. systems important to safety will continue to 1. Does the proposed change involve a
NextEra Energy Duane Arnold, LLC, be operated and maintained within their significant increase in the probability or
Docket No. 50–331, Duane Arnold design bases. Relocation of the Control Rod consequences of an accident previously
Energy Center (DAEC), Linn County, Block Instrumentation MCPR values to the evaluated?
COLR has no influence or impact on new or Response: No.
Iowa different kind of accidents. The proposed change replaces an existing
Date of amendment request: March Therefore, the proposed change does not Surveillance Requirement to operate the
24, 2017. A publicly-available version is create the possibility of a new or different CREMAFS [Control Room Emergency
in ADAMS under Accession No. kind of accident from any accident Makeup Air and Filtration System],
previously evaluated. FSBEACS [Fuel Storage Building Emergency
sradovich on DSK3GMQ082PROD with NOTICES
ML17086A442.
Description of amendment request: 3. Does the proposed change involve a Air Cleaning System], and SBVS [Shield
significant reduction in a margin of safety? Building Ventilation System] equipped with
The proposed change would relocate
Response: No. electric heaters for at least a continuous 10-
cycle specific minimum critical power The margin of safety is not affected by the hour period in accordance with the SFCP
ratio (MCPR) values to the DAEC core relocation of cycle-specific Control Rod [Surveillance Frequency Control Program]
operating limits report (COLR). The Block Instrumentation MCPR values from the with a requirement to operate the systems for
proposed amendment would revise the TS to the COLR. Appropriate measures exist 15 continuous minutes with heaters
DAEC technical specifications (TS) to to control the values of these cycle-specific operating.
VerDate Sep<11>2014 21:15 May 22, 2017 Jkt 241001 PO 00000 Frm 00107 Fmt 4703 Sfmt 4703 E:\FR\FM\23MYN1.SGM 23MYN1](https://thefederalregister.org/doc/82_FR_23713/page/13.svg)
![23632 Federal Register / Vol. 82, No. 98 / Tuesday, May 23, 2017 / Notices
methodology, as established in NRC clarified the application, did not expand SUMMARY: The U.S. Nuclear Regulatory
Regulatory Guide 1.183, ‘‘Alternative the scope of the application as originally Commission (NRC) is issuing a partial
Radiological Source Terms for noticed, and did not change the NRC exemption in response to an October 13,
Evaluating Design Basis Accidents at staff’s original proposed no significant 2016, request from Dominion Energy
Nuclear Power Reactors,’’ July 2000 hazards consideration determination as Kewaunee, Inc. (the licensee or DEK).
(ADAMS Accession No. ML003716792), published in the Federal Register. The issuance of the exemption would
is used to calculate the offsite and The Commission’s related evaluation grant DEK a partial exemption from
control room radiological consequences of the amendments is contained in a regulations that require the retention of
of postulated accidents for DCPP, Units Safety Evaluation dated April 27, 2017. records for certain systems, structures,
1 and 2. The amendments revised No significant hazards consideration and components associated with the
Technical Specification (TS) 1.1, comments received: No. Kewaunee Power Station (KPS) until the
‘‘Definitions,’’ for the definition of Dose termination of the KPS operating
Equivalent I–131; TS 3.4.16, ‘‘RCS Tennessee Valley Authority, Docket No.
license.
[Reactor Coolant System] Specific 50–390, Watts Bar Nuclear Plant, Unit 1,
Rhea County, Tennessee DATES: The exemption was issued on
Activity,’’ to revise the noble gas May 10, 2017.
activity limit; TS 3.6.3, ‘‘Containment Date of amendment request: June 7,
ADDRESSES: Please refer to Docket ID
Isolation Valves,’’ to require the 48-inch 2016.
containment purge supply and exhaust NRC–2017–0121 when contacting the
Brief description of amendment: The
valves to be sealed closed during Modes NRC about the availability of
amendment modified the Technical
1, 2, 3, and 4; TS 5.5.11, ‘‘Ventilation information regarding this document.
Specifications to allow the use of
Filter Testing Program (VFTP),’’ to You may obtain publicly-available
Component Cooling System (CCS) pump
change the allowable methyl iodide information related to this document
2B–B to support Train 1B operability
penetration testing criteria for the using any of the following methods:
when the normally aligned CCS pump • Federal Rulemaking Web site: Go to
auxiliary building system charcoal filter; C–S is removed from service.
TS 5.5.19, ‘‘Control Room Habitability http://www.regulations.gov and search
Date of issuance: April 27, 2017. for Docket ID NRC–2017–0121. Address
Program,’’ to replace ‘‘whole body or its Effective date: As of the date of
equivalent to any part of the body,’’ questions about NRC dockets to Carol
issuance and shall be implemented Gallagher; telephone: 301–415–3463;
with ‘‘Total Effective Dose Equivalent,’’ within 60 days of issuance.
which is the dose criteria specified in 10 email: Carol.Gallagher@nrc.gov. For
Amendment No.: 113. A publicly- technical questions, contact the
CFR 50.67, and Appendix D, available version is in ADAMS under
‘‘Additional Conditions,’’ for Facility individual listed in the FOR FURTHER
Accession No. ML17081A263; INFORMATION CONTACT section of this
Operating License Nos. DPR–80 and documents related to this amendment
DPR–82 for DCPP, Units 1 and 2, to add document.
are listed in the Safety Evaluation • NRC’s Agencywide Documents
additional license conditions. enclosed with the amendment.
Date of issuance: April 27, 2017. Access and Management System
Facility Operating License No. NPF– (ADAMS): You may obtain publicly-
Effective date: As of its date of 90: Amendment revised the Facility
issuance and shall be implemented available documents online in the
Operating License and Technical ADAMS Public Documents collection at
within 365 days from the date of Specifications.
issuance. http://www.nrc.gov/reading-rm/
Date of initial notice in Federal adams.html. To begin the search, select
Amendment Nos.: Unit 1–230; Unit
Register: September 13, 2016 (81 FR ‘‘ADAMS Public Documents’’ and then
2–232. A publicly-available version is in
62932). select ‘‘Begin Web-based ADAMS
ADAMS under Accession No.
The Commission’s related evaluation Search.’’ For problems with ADAMS,
ML17012A246; documents related to
of the amendment is contained in a please contact the NRC’s Public
these amendments are listed in the
Safety Evaluation dated April 27, 2017. Document Room (PDR) reference staff at
Safety Evaluation enclosed with the
No significant hazards consideration 1–800–397–4209, 301–415–4737, or by
amendments.
comments received: No. email to pdr.resource@nrc.gov. The
Facility Operating License Nos. DPR–
80 and DPR–82: The amendments Dated at Rockville, Maryland, this 11th day ADAMS accession number for each
revised the Facility Operating Licenses of May 2017. document referenced (if it is available in
and Technical Specifications. For the Nuclear Regulatory Commission. ADAMS) is provided the first time that
Date of initial notice in Federal Kathryn M. Brock, it is mentioned in this document.
Register: The license amendment Deputy Director, Division of Operating • NRC’s PDR: You may examine and
request was originally noticed in the Reactor Licensing, Office of Nuclear Reactor purchase copies of public documents at
Federal Register on October 13, 2015 Regulation. the NRC’s PDR, Room O1–F21, One
(80 FR 61486). As a result of the [FR Doc. 2017–10570 Filed 5–22–17; 8:45 am] White Flint North, 11555 Rockville
supplemental letters dated October 22, BILLING CODE 7590–01–P Pike, Rockville, Maryland 20852.
November 2, November 6, and FOR FURTHER INFORMATION CONTACT: Ted
December 17, 2015; and February 1, H. Carter, Office of Nuclear Material
February 10, April 21, June 9, and NUCLEAR REGULATORY Safety and Safeguards; U.S. Nuclear
September 15, 2016, the notice was COMMISSION Regulatory Commission, Washington,
reissued in its entirety to include the DC 20555–0001; telephone: 301–415–
[Docket Nos. 50–305; NRC–2017–0121]
sradovich on DSK3GMQ082PROD with NOTICES
revised scope, description of the 5543; email: Ted.Carter@nrc.gov.
amendment request, and proposed no SUPPLEMENTARY INFORMATION:
significant hazards consideration Dominion Energy Kewaunee, Inc.;
determination on November 8, 2016 (81 Kewaunee Power Station I. Background
FR 78664). AGENCY: Nuclear Regulatory By letter dated May 14, 2013, DEK
The supplemental letters dated Commission. submitted a certification of permanent
October 6 and December 27, 2016, removal of fuel from the KPS reactor
ACTION: Exemption; issuance.
provided additional information that vessel (ADAMS Accession No.
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Document Created: 2017-05-23 02:14:52
Document Modified: 2017-05-23 02:14:52
| 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 June 22, 2017. A request for a hearing must be filed by July 24, 2017. | |
| 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 | 82 FR 23615 |
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