82 FR 55401 - 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 223 (November 21, 2017)
Federal Register Volume 82, Issue 223 (November 21, 2017)
| Page Range | 55401-55421 | |
| FR Document | 2017-25063 |
[Federal Register Volume 82, Number 223 (Tuesday, November 21, 2017)] [Notices] [Pages 55401-55421] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-25063] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0220] 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 October 24, 2017 to November 6, 2017. The last biweekly notice was published on November 7, 2017. DATES: Comments must be filed by December 21, 2017. A request for a hearing must be filed by January 22, 2018. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0220. 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: May Ma, Office of Administration, Mail Stop: OWFN-2-A13, 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: Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-5411, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2017-0220, 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-2017-0220. 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-2017-0220, facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of [[Page 55402]] 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 no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC 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 [[Page 55403]] 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 additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP), Van Buren County, Michigan Date of amendment request: August 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17248A389. Description of amendment request: The proposed amendment would revise the PNP Site Emergency Plan (SEP) for the permanently shut down and defueled condition. The proposed PNP SEP changes would revise the shift staffing and Emergency Response Organization (ERO) staffing. 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 to the PNP SEP do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not affect accident initiators or precursors, nor does it alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and augmented ERO to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently shut down and defueled condition. The proposed changes only remove positions that will no longer be credited in the PNP SEP. 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? [[Page 55404]] Response: No. The proposed changes reduce the number of on-shift and augmented ERO positions commensurate with the hazards associated with a permanently shut down and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new accident initiators are created. 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. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the PNP SEP and do not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised PNP SEP will continue to provide the necessary response staff with the proposed changes. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601. NRC Branch Chief: Douglas A. Broaddus. Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Date of amendment request: September 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17244A093. Description of amendment request: The amendments would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR, Section 50.69, ``Risk- informed categorization and treatment of structures, systems and components for nuclear power reactors.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs [structures, systems, and components] subject to NRC [Nuclear Regulatory Commission] special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results. 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: Tamra Domeyer, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: August 30, 2017, as supplemented by letter dated October 24, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17243A014 and ML17297B521, respectively. Description of amendment request: The amendments would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR 50.69, ``Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, with NRC staff edits shown in square brackets: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of [structures, systems, and components] SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment [[Page 55405]] requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions. 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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results. 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 60555. NRC Branch Chief: James G. Danna. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: September 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A069. Description of amendment request: The amendments would revise Technical Specification (TS) requirements related to the direct current (DC) electrical power system. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC Electrical Rewrite--Update to TSTF-360.'' 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 restructures the TS for the direct current (DC) electrical power system. The proposed changes add actions to specifically address battery charger inoperability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Updated Final Safety Analysis Report (UFSAR). Operation in accordance with the proposed TS ensures that the DC electrical power system is capable of performing its function as described in the UFSAR. Therefore, the mitigative functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis, and the probability of previously analyzed accidents will not increase by implementing these changes. The relocation of preventive maintenance surveillances, and certain operating limits and actions, to a newly created licensee- controlled Battery Monitoring and Maintenance Program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, ``Requirements for monitoring the effectiveness of maintenance at nuclear power plants,'' which will ensure the control of maintenance activities associated with the DC electrical power system. The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes. 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 involves restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the UFSAR. Rather, the DC electrical power system is used to supply equipment used to mitigate an accident. 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 established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery maintenance and monitoring program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions. 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 60555. NRC Branch Chief: James G. Danna. Exelon Generation Company, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: October 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A520. [[Page 55406]] Description of amendment request: The amendment would revise the James A. FitzPatrick Nuclear Power Plant Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control'' (ADAMS Accession No. ML16074A448). Specifically, the licensee proposed changes to replace TS requirements related to operations with a potential for draining the reactor vessel (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated, and therefore replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated. The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event. The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed. The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases. 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 change involve a significant reduction in a margin of safety? Response: No. The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin. 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: Donald P. Ferraro, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305, Kennett Square, PA 19348. NRC Branch Chief: James G. Danna. Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3, and 4, Miami-Dade County, Florida Date of amendment request: August 23, 2017, as supplemented by letter dated October 19, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17235B008 and ML17292A789, respectively. Description of amendment request: The amendments would modify the Technical Specifications (TSs) to relocate the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, and Gas Decay Tanks System requirements to licensee-controlled documents and establish a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed amendments also relocate the Standby Feedwater System requirements to licensee-controlled documents and modify related Auxiliary Feedwater (AFW) System requirements. 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. [[Page 55407]] The proposed license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes are administrative in nature and do not alter any plant equipment or the manner in which plant equipment is operated and maintained. All equipment limitations, applicable methodologies and surveillances are maintained by the proposed changes. In addition, the proposed changes to the AFW System requirements enhance plant safety. As such, the proposed changes cannot affect the initiators, the likelihood or the expected outcomes of any analyzed accidents. Therefore, facility operation in accordance with the proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes neither install or remove plant equipment nor alter any plant equipment design, configuration, or method of operation. Hence, no new failure mechanisms are introduced as a result of the proposed changes. 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 license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes neither involve changes to safety analyses assumptions, safety limits, or limiting safety system settings nor adversely impact plant operating margins or the reliability of equipment credited in safety analyses. Therefore, operation of the facility in accordance with the proposed changes will not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William S. Blair, Managing Attorney-- Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, FL 33408-0420. NRC Branch Chief: Undine Shoop. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa Date of amendment request: September 5, 2017. A publicly-available version is in ADAMS under Accession No. ML17248A284. Description of amendment request: The proposed amendment would revise DAEC Technical Specifications 3.5.1, ``ECCS [emergency core cooling system]-Operating.'' The proposed change would decrease the nitrogen supply requirement for the Automatic Depressurization System (ADS) in Surveillance Requirement (SR) 3.5.1.3 from 100 days to 30 days. 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 modifies a SR for verification of the nitrogen supply for the ADS accumulators. Accidents are initiated by the malfunction of plant equipment, or the catastrophic failure of plant structures, systems or components. The performance of this surveillance is not a precursor to any accident previously evaluated and does not change the manner in which the ADS operates. Technical evaluation of the change concluded that a 30-day nitrogen supply is more than adequate to ensure that the reactor is depressurized, so the consequences of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the probability or consequence 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 change does not involve physical alterations to the plant. No new or different type of equipment will be installed, and there are no physical modifications required to existing installed equipment associated with the proposed change. The proposed change does not create any failure mechanism, malfunction or accident initiator not already considered in the design and licensing basis. 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. Although the proposed change will decrease the required supply of nitrogen for the ADS accumulators from 100 days to 30 days, the assessment above has shown that the reactor would be depressurized within 3 days following any postulated accident or event that would create a hostile environment in the drywell. Once initial depressurization is completed, long term core cooling can be assured without ADS. Therefore, the proposed change will not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: August 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17243A469. Description of amendment request: The proposed amendment would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR, part 50.69, ``Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment [[Page 55408]] requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions. 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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. NextEra Energy, Point Beach Nuclear Plant (PBNP), LLC, Docket Nos. 50- 266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: June 23, 2017, as supplemented by letter dated August 21, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17174A458, and ML17233A283, respectively. Description of amendment request: The amendments would revise the Emergency Plan for PBNP to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors,'' which has been endorsed by the NRC. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. No actual facility equipment or accident analyses are affected by the proposed changes. The change revises the NextEra Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, ``Methodology for Development of Emergency Action Levels,'' but does not alter any of the requirements of the Operating License or the Technical Specifications. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed change does not create any new failure modes for existing equipment or any new limiting single failures. Additionally, the proposed change does not involve a change in the methods governing normal plant operation, and all safety functions will continue to perform as previously assumed in the accident analyses. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. 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 amendment involve a significant reduction in a margin of safety? Response: No. The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of safety-related systems and components. The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis. The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of amendment request: September 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17270A076. Description of amendment request: The amendments would relocate the reactor coolant system pressure isolation valve (RCS PIV) table from the technical specifications (TSs) to the technical requirements manual (TRM). The request would also remove references to the table and move all notes and leakage acceptance criteria from the table to the TS surveillance requirements. [[Page 55409]] 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 TS will not alter the way any structure, system, or component (SSC) functions, and will not alter the manner in which the plant is operated. The proposed changes do not alter the design of any SSC. The relocation of the RCS PIV valve lists from the TS to the TRM is an administrative change. Future revisions to the TRM are subject to 10 CFR 50.59. Therefore the probability of an accident previously evaluated is not significantly increased. The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. Therefore, the consequences of an accident previously evaluated are not increased. Therefore, these proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism. The proposed changes are administrative in nature. 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 a margin of safety? Response: No. The proposed changes to the RCS PIV TS are administrative in nature. The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. The proposed changes will not result in changes to system design or setpoints that are intended to ensure timely identification of plant conditions that could be precursors to accidents or potential degradation of accident mitigation systems. The proposed amendment will not result in a design basis or safety limit being exceeded or altered. Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, 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: 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, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina Date of amendment request: October 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17279A715. Description of amendment request: The proposed amendment would increase the Integrated Leak Rate Test (ILRT) Peak Calculated Containment Internal Pressure, P a , listed in Technical Specification (TS) 6.8.4.g, ``Containment Leakage Rate Testing Program,'' to remove the reference to Regulatory Guide (RG) 1.163, ``Performance-Based Containment Leak Test Program,'' dated September 1995 and ANSI/ANS (American National Standards Institute/American Nuclear Society)--56.8-2002, ``Containment System Leakage Testing Requirements,'' and to replace the reference of Nuclear Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based option of 10 CFR part 50, Appendix J,'' with NEI 94-01, Revision 2-A. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, with NRC staff edits in square brackets: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes involve removal of RG 1.163 and ANSl/ANS- 56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the Pa [Peak Calculated Containment Internal Pressure] value for containment leakage testing. The activity does not involve 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. As such, the reactor containment itself and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment 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, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The integrity of the reactor containment is subject to two types of failure mechanisms which can be categorized as (1) activity based and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The updated Pa value reflects the updated mass and energy release and containment response calculations, ensuring a sound technical basis for the local and integrated leakage tests. To mitigate time-based mechanisms, the design and construction requirements of the containment itself combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers], Section XI and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by a Type A test. The change to the Pa value is less than 1 psid [per square inch differential]. Radiological consequences will continue to be evaluated at the Technical Specification allowed leakage, La [allowed leakage] of 0.20 percent by weight of air, which will not be increased despite the increase in Pa . As described in Section 3.5, past leakage testing yielded values well under La . Based on the above, neither the reference changes nor the Pa change involves a significant increase in the 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 involve removal of RG 1.163 and ANSl/ANS- 56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the Pa value for containment leakage testing. The reactor containment and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident. There are not any accident initiators or precursors affected by the revision. The proposed TS change does not involve a physical change to the plant or the manner in which the plant is operated or controlled. Therefore, the proposed TS 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? [[Page 55410]] Response: No. The proposed changes involve removal of RG 1.163 and ANSl/ANS- 56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the Pa value for containment leakage testing. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. Using the same analysis methodology as described in WCAP-10325-P-A [Westinghouse LOCA [loss- of-accident coolant] Mass and Energy Release Model for Containment Design], the updated mass and energy release and containment response analyses corrected input errors identified in the NSALs [Westinghouse Nuclear Safety Advisory Letters] described previously. As shown in Figure 1 [October 6, 2017, submittal], the correction of these errors resulted in a slightly higher predicted peak pressure than that of the current licensing basis but does not pose a significant challenge to the design limit. The specific requirements and conditions of the Primary Containment Leak Rate Testing Program, as defined in the Technical Specifications, exist to ensure that the degree of reactor containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by the Technical Specification is maintained. The containment inspections performed in accordance with ASME, Section XI and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety that is in plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards will continue to be met. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLP, 1111 Pennsylvania Avenue NW., Washington, DC 20004. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: July 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17209A759. Description of amendment request: The amendment request proposes to revise Technical Specification Section 1.1 (TS), Definition of Actuation Logic Test, by adding a new TS Section 1.1 Definition of Actuation Logic Output Test (ALOT), revising existing Surveillance Requirements 3.3.15.1 and 3.3.16.1 and adding new Surveillance Requirements 3.3.15.2 and 3.3.16.2 to implement the new ALOT. This submittal requests approval of the license amendment that is necessary to implement these changes. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(A), licensee has provided its analysis of the issue on no significant hazards consideration determination, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. There are no design changes associated with the proposed amendment. All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable. The [Processor Module Self-Diagnostic (PMS)] will continue to function in a manner consistent with the plant design basis. There will be no changes to the PMS operating limits. The existing ACTUATION LOGIC TEST Surveillance Requirements are revised such that different portions of the PMS logic circuitry are tested on appropriate surveillance test frequencies. The proposed change will not adversely affect accident initiators or precursors or adversely alter the design assumptions, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors. The proposed changes will not alter the ability of structures, systems, and components (SSCs) to perform their specified safety functions to mitigate the consequences of an initiating event within the assumed acceptance limits. 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 any accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR). The applicable radiological dose acceptance criteria will continue to be met. The proposed change revises the frequency of testing certain portions of the PMS logic circuitry, but does not physically alter any safety-related systems. 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 different kind of accident from any accident previously evaluated? Response: No. With respect to any new or different kind of accident, there are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC performs its specified safety function. The proposed change will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed change will not alter any assumptions made in the safety analyses. The proposed change revises the frequency of testing certain portions of the PMS logic circuitry. The proposed change does not involve a physical modification of the plant. 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 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 existing ACTUATION LOGIC TEST Surveillance Requirements are revised such that different portions of the PMS logic circuitry are tested on appropriate surveillance test frequencies. The reliability of the PMS is such that not testing the Component Interface Module (CIM) logic and driver output circuits when the reactor is at power will have a net positive impact on Engineered Safety Feature Actuation System (ESFAS) availability. There will be a reduction in the potential for challenges to the safety systems, coupled with less time that the safety systems are unavailable. There will be no effect on those plant systems necessary to effect the accomplishment of protection functions. No instrument setpoints or system response times are affected. None of the acceptance criteria for any accident analysis will be changed. The proposed change will have no impact on the radiological consequences of a design basis accident. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. [[Page 55411]] NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: August 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17230A365. Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information (text) and involved Tier 2* information (as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information). This amendment request proposes increasing the design pressure of the main steam (MS) isolation valve (MSIV) compartments from 6.0 to 6.5 psi and proposes other changes to the licensing basis regarding descriptions of the MSIV compartments. 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 Nuclear Regulatory Commission (NRC) staff's edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes do not adversely affect the operation of any structures, systems, and components inside or outside the auxiliary building that could initiate or mitigate abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods, tornado missiles, and turbine missiles, or their safety or design analyses, evaluated in the UFSAR. The changes do not adversely affect any design function of the auxiliary building or the structures, systems, and components contained therein. The ability of the affected auxiliary building main steam isolation valve compartments and adjacent rooms, including the main control room, to withstand the pressurization effects from the postulated pipe ruptures is not adversely affected by the increase in design pressure, since the structures, systems, and components therein remain qualified for this service. Therefore, the proposed activity 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 [structure, system, and component (SSC)] such that a new accident initiator or initiating sequence of events is created. 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, and fourth-stage [automatic depressurization system (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 margin of safety for the design of the auxiliary building is maintained through continued use of approved codes and standards as stated in the UFSAR, and adherence to the assumptions used in the analyses of this structure and the events associated with this structure. The auxiliary building continues to be a seismic Category I building with all current structural safety margins maintained. The 3-hour fire rating requirements for the impacted auxiliary building walls are maintained. The equipment housed in the main steam isolation valve compartments continue to be environmentally qualified for their intended service in accordance with the approved codes and standards stated within the UFSAR. Thus, the requested changes will not adversely affect any safety-related equipment, design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change, thus, no margin of safety is reduced. 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: October 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17279A084. Description of amendment request: The amendment request proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document (DCD) Tier 2 information) and involves related changes to plant-specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific DCD Tier 1 material departures. Specifically, the requested amendment proposes to depart from Tier 2 information in UFSAR Subsection 8.3.2.4 describing raceway and cable routing criteria and hazard protection, and involves related changes to plant-specific Tier 1 Table 3.3-6, inspections, tests, analyses, and acceptance criteria information, with corresponding changes to the associated COL Appendix C information. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below with NRC staff edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between raceways. Change 2 reduces the required separation distances between raceways from those documented in [Institute of Electrical and Electronics Engineers (IEEE)] 384-1981. These reduced separation distances are based on specific tests performed on the specified raceway configurations, and the recommendations from those tests contained in the associated report. The NRC staff previously reviewed the descriptions of the ten tests documented in this report, including the ones applicable to the existing UFSAR exceptions, and concluded that they were acceptable, as documented in NUREG-1793, ``Final Safety Evaluation Report Related to Certification of [[Page 55412]] the AP1000 Standard Design,'' (Initial Report) Subsection 8.3.2.2. The reduced separation does not adversely impact the ability to safely shutdown the plant, and maintain it shutdown. The referenced test report has shown a failure of a faulted cable will not propagate to a nearby target cable in way that adversely impacts its function. 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. Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between circuits. Change 2 reduces the required separation distances between circuits from those documented in IEEE 384-1981. This change does not result in a new accident initiator or impact a current accident initiator. 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. Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between circuits. Change 2 reduces the required separation distances between circuits from those documented in IEEE 384-1981. These reduced separation distances are based on specific tests performed on the specified raceway configurations, and the recommendations from those tests contained in the associated report. The NRC staff previously reviewed the descriptions of the ten tests documented in this report, including the ones applicable to the existing UFSAR exceptions, and concluded that they were acceptable, as documented in NUREG-1793, ``Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design,'' (Initial Report) Subsection 8.3.2.2. The reduced separation does not adversely impact the ability to safely shutdown the plant, and maintain it shutdown. The referenced test report has shown a failure of a faulted cable will not propagate to a nearby target cable in a way that adversely impacts its function. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: July 10, 2017. A publicly-available version is in ADAMS under Accession No. ML17191B163. Description of amendment request: The amendments would revise the technical specifications (TSs) by: (1) Adding a Note to the surveillance requirements (SRs) of TS 3.7.7, ``Main Turbine Bypass System,'' to clarify that the SRs are not required to be met when the limiting condition for operation (LCO) does not require the Main Turbine Bypass System to be operable, (2) clarifying that LCO 3.2.3, ``LINEAR HEAT GENERATION RATE (LHGR),'' also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deleting an outdated footnote for LCO 3.2.3. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, ``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. The proposed change does not affect the requirement to meet the LCO, nor does it affect the requirements to perform the SRs when the Main Turbine Bypass System is being used to meet the LCO. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The current safety analysis evaluation is unaffected by this proposed change. The change regarding the outdated footnote has no effect on the actual TS requirements. 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 (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, ``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The change regarding the outdated footnote has no effect on the actual TS requirements. The current safety analysis evaluation is unaffected by these proposed changes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, ``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The applicable safety analyses for TS 3.7.7 is unaffected by this clarification. The change regarding the outdated footnote has no effect on the actual TS requirements. 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 [[Page 55413]] amendment request involves no significant hazards consideration. Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35242. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: September 13, 2017. A publicly-available version is in ADAMS under Accession No. ML17256A626. Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information, and from Technical Specifications as incorporated in Appendix A of the Combined License (COL). Specifically, the proposed changes revise COL Appendix A Technical Specification 3.6.8 to identify the trisodium phosphate (TSP) mass value required in the pH adjustment baskets. The TSP mass value adjusts the pH of the containment water to >7.0 following a postulated accident. 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 activity revises the mass of trisodium phosphate (TSP), which raises the pH of post-accident containment water to 7.0 or greater following a postulated accident. The change to the TSP mass value does not adversely impact the ability to support radionuclide retention with high radioactivity in containment and helps prevent corrosion of containment equipment during long-term floodup conditions. The proposed changes do not adversely impact previously evaluated accidents, because pH control capability is provided to mitigate already postulated accidents. As described in Updated Final Safety Analysis Report (UFSAR) Subsection 15.6.5.3.1.3, the passive core cooling system (PXS) is assumed to provide sufficient TSP to the post-loss-of-coolant accident (LOCA) cooling solution to maintain the pH at greater than or equal to 7.0 following a LOCA. The pH adjustment baskets provide for long-term pH control. Long-term pH control is not adversely impacted as the pH adjustment baskets contain the required amount of TSP to support pH control requirements following a design basis accident (DBA). No safety-related structure, system, component (SSC) or function is adversely affected by this change. The change does not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. The proposed changes do not involve a change to the predicted radiological releases due to postulated accident conditions, thus, 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 activity revises the mass of TSP, which raises the pH of containment to 7.0 or greater following a postulated accident. The proposed activity does not create the possibility of a new or different kind of accident as pH adjustment is used to support proper containment chemistry requirements following an accident. The proposed activity does not adversely affect any safety related equipment, and does not add any new interfaces to safety-related SSCs that adversely affect safety functions. No system or design function or equipment qualification is adversely affected by these changes as the changes do not modify any SSCs that prevent safety functions from being performed. The capability to maintain a maximum containment pH below 9.5 is not adversely impacted by these changes. The changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety related equipment. 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 activity revises the mass of TSP, which raises the pH of containment to 7.0 or greater following a postulated accident. The proposed activity does not affect any other safety-related equipment or fission product barriers. Containment water pH adjustment is not adversely impacted. The requested changes will not adversely affect compliance with any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes as previously evaluated accidents are not impacted. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and based on this review it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazard consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: September 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17272A957. Description of amendment request: The requested amendment proposes to depart from Tier 2* and associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific DCD Tier 2 information). The requested amendment proposes to depart from UFSAR Tier 2* information regarding resolution of human engineering deficiencies (HEDs) contained in Westinghouse Electric Company's report APP-OCS-GEH-320, ``AP1000 Human Factors Engineering Integrated Systems Validation Plan,'' which is incorporated by reference into the VEGP Units 3 and 4 UFSAR. The proposed changes would revise the licensing basis of the combined licenses regarding the process for addressing and re-testing of HEDs identified during the integrated system validation (ISV) as described in Tier 2* document, APPOCS-GEH-320 ``AP1000 Human Factors Engineering Integrated System Validation Plan.'' APPOCS-GEH-320 references APP-OCS-GEH-420, ``Human Factors Engineering Discrepancy Resolution Process,'' which defines the process for tracking, resolution, and closure of HEDs. The proposed changes to APP-OCS-GEH- 320 do not impact APP-OCS-GEH-420. 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 Integrated System Validation (ISV) provides a comprehensive human performance-based assessment of the design of the AP1000 Human-System Interface (HSI) resources, based on their realistic operation [[Page 55414]] within a simulator driven Main Control Room (MCR). The ISV is part of the overall AP1000 Human Factors Engineering (HFE) program. The changes to APP-OCS-GEH-320, which is incorporated by reference into the UFSAR, clarify the resources and methodology used during re- testing performed to verify the effectiveness of Human Engineering Deficiency (HED) resolution. The ISV Plan does not affect the plant itself. Changing APP-OCS-GEH-320 and the UFSAR does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, component (SSC) or function is adversely affected. The changes neither involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate an accident, 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 changes to APP-OCS-GEH-320 and the VEGP 3 and 4 UFSAR affect only the testing and validation of the MCR design and HSI using a plant simulator. Therefore, the changes do not affect the safety- related equipment itself, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected. No system or design function or equipment qualification is adversely affected by the changes. 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 would result in significant fuel cladding failures. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety related equipment. 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 changes to APP-OCS-GEH-320 and the UFSAR affect the testing and validation of the MCR design and HSI using a plant simulator. Therefore, the changes do not affect the assessments or the plant itself. These changes do not affect safety-related equipment or equipment whose failure could initiate an accident, nor does it adversely interface with safety-related equipment or fission product barriers. No safety analysis or design basis acceptance limit/ criterion is challenged or exceeded by the requested change. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and based on this review it appears that the three standards of 10 CFR 50.92 (c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazard consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: September 20, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17265A434. Description of amendment request: The amendments would revise technical specification (TS) requirements related to ``operations with a potential for draining the reactor vessel'' (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires RPV water level to be greater than the top of active irradiated fuel. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' dated December 20, 2016. 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 replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated. The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event. The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed. The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and the Control Room Emergency Outside Air Supply (CREOAS) system. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event. The administrative update to delete expired completion time notes is purely administrative in nature. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the [[Page 55415]] limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases. The administrative update to delete expired completion time notes is purely administrative in nature. 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 change involve a significant reduction in a margin of safety? Response: No. The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin. The administrative update to delete expired completion time notes is purely administrative in nature. 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: Damon D. Obie, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101. NRC Branch Chief: James G. Danna. Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50- 296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County, Alabama Date of amendment request: August 15, 2017. A publicly-available version is in ADAMS under Accession No. ML17228A490. Description of amendment request: The amendments would revise the BFN, Units 1, 2, and 3 Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' by adopting Nuclear Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' as the implementation document for the performance-based Option B of 10 CFR part 50, Appendix J. The proposed changes permanently extend the Type A containment integrated leak rate testing (ILRT) interval from 10 years to 15 years and the Type C local leakage rate testing (LLRT) intervals from 60 months to 75 months. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated? Response: No. The proposed revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current Type A test interval of 10 years would be extended to 15 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident because research documented in NUREG-1493, ``Performance- Based Containment System Leakage Testing Requirements'' [``Performance-Based Containment Leak-Test Program''], September 1995, has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. NUREG-1493 concluded that reducing the Type A testing frequency to one per 20 years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3) shows leakage to be below acceptance criteria, indicating a very leak tight containment. Inspections required by the ASME Code [American Society of Mechanical Engineers Boiler and Press Vessel Code] Section Xl (Subsection IWE) and Maintenance Rule monitoring (10 CFR 50.65, ``Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants'') are performed in order to identify indications of containment degradation that could affect that leak tightness. Types B and C testing required by TSs will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test interval extension will not represent a significant increase in the consequences of an accident. The proposed amendment involves changes to the BFN, Units 1, 2, and 3, 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the units are operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself 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, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment. The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, performance-based leakage testing program. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses. The potential consequences of extending the ILRT interval from 10 years to 15 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem [roentgen equivalent man] per year resulting from design basis accidents was estimated to be very small, and the increase in the LERF [large early release frequency] resulting from the proposed change was determined to be within the guidelines published in NRC RG [Regulatory Guide] 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. TVA has determined that the increase in CCFP [conditional containment failure probability] due to the proposed change would be very small. Based on the above discussions, the proposed changes do not involve an 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. [[Page 55416]] The proposed revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3). The proposed extension to Type A and Type C test intervals does not create the possibility of a new or different type of accident because there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3). The proposed extension to Type A testing will not significantly reduce the margin of safety. NUREG-1493, ``Performance-Based Containment System Leakage Testing Requirements'' [``Performance-Based Containment Leak-Test Program''], September 1995, generic study of the effects of extending containment leakage testing, found that a 20 year extension to Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1% to the individual risk and that the decrease in Type A testing frequency would have a minimal effect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the ASME Code Section Xl (Subsection IWE) and maintenance rule monitoring (10 CFR 50.65, ``Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants'') will further reduce the risk of a containment leakage path going undetected. The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, performance-based leakage testing program, and establishes a 15-year interval for the performance of the primary containment ILRT and a 75-month interval for Type C testing. The amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the 10 CFR part 50, Appendix J Testing Program Plan, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant safety analyses is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, B, and C containment leakage tests will continue to be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A. Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by an ILRT. This ensures that evidence of containment structural degradation is identified in a timely manner. Furthermore, a risk assessment using the current BFN, Units 1, 2, and 3, PRA [probabilistic risk assessment] model concluded that extending the ILRT test interval from 10 years to 15 years results in a very small change to the BFN, Units 1, 2, and 3, risk profile. Accordingly, 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902. NRC Branch Chief: Undine Shoop. Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2 (SQN), Hamilton County, Tennessee Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee Date of amendment request: August 7, 2017. A publicly-available version is in ADAMS under Accession No. ML17219A505. Description of amendment request: The amendment would revise Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio (QPTR),'' and TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' to avoid confusion as to when an incore power distribution measurement for QPTR is required. The amendment would also revise the WBN TSs for consistency with the existing SQN TSs and Westinghouse Standard TSs in NUREG-1431, Revision 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 do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes do not significantly increase the probability of an accident and are consistent with safety analysis assumptions and resultant consequences. Therefore, the changes do not increase 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 result in a change in the manner in which the reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. There are no design changes associated with the change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also [[Page 55417]] maintained. All signals credited as providing primary or secondary protection, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Branch Chief: Undine Shoop. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment requests: December 15, 2016. Brief description of amendments: The amendments modified Technical Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS 3.7.4, ``Steam Generator Power Operated Relief Valves (SG PORVs),'' and TS 3.7.6, ``Condensate Storage System,'' to revise the Completion Times for Limiting Condition for Operation (LCO) of TS LCO 3.4.10 Required Action B.2, TS LCO 3.7.4 Required Action C.2, and TS LCO 3.7.6 Required Action B.2 from 12 to 24 hours. The proposed changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A, Revision 1, ``Provide Consistent Completion Time to Reach MODE 4.'' Date of issuance: October 23, 2017. Effective date: These license amendments are effective as of its date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 294 (Unit 1) and 290 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17254A144; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the renewed licenses and technical specifications. Date of initial notice in Federal Register: April 25, 2017 (82 FR 19099). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 23, 2017. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment requests: December 15, 2016. Brief description of amendments: The amendments modified technical specification (TS) limiting condition for operation (LCO) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' Condition A and Required Action A.1. Condition A was revised to include the situation when one turbine- driven AFW pump is inoperable in MODE 3, immediately following a refueling outage, only applicable if MODE 2 has not been entered following the refueling outage. Required Action A.1 was revised to include the turbine-driven AFW addition to Condition A. The amendments are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-340-A, Revision 3, ``Allow 7 day Completion Time for a turbine- driven AFW pump inoperable.'' Date of issuance: October 23, 2017. Effective date: These license amendments are effective as of its date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 295 (Unit 1) and 291 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17257A297; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the renewed licenses and TSs. Date of initial notice in Federal Register: April 25, 2017 (82 FR 19100). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 23, 2017. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment requests: December 15, 2016. Brief description of amendments: The amendments revised Technical Specification 3.1.2, ``Core Reactivity,'' to revise the Completion Times of Required Actions A.1 and A.2 from 72 hours to 7 days. This proposed change is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the Completion Time when the Core Reactivity Balance is Not Within Limit.'' Date of issuance: October 23, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 296 (Unit 1) and 292 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17261B290; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the Renewed Licenses and Technical Specifications. [[Page 55418]] Date of initial notice in Federal Register: April 11, 2017 (82 FR 17457). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 23, 2017. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment requests: January 11, 2017. Brief description of amendments: The amendments modified Technical Specification (TS) 3.8.1, ``AC 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 was consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction Notes.'' Date of issuance: October 25, 2017. Effective date: These license amendments are effective as of its date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 300 (Unit 1) and 279 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A055; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23620). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 25, 2017. No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the no significant hazards consideration determination nor the license amendment request. 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 requests: January 11, 2017. Brief description of amendments: The amendments modified 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 Specifications Task Force (TSTF) Traveler TSTF-315-A, Revision 0, ``Reduce plant trips due to spurious signals to the NIS [Nuclear Instrumentation System] during physics testing.'' Date of issuance: October 25, 2017. Effective date: These license amendments are effective as of their date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 301 (Unit 1) and 280 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17261B218; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23621). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 25, 2017. No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the proposed no significant hazards consideration determination or to the license amendment request. 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 requests: January 11, 2017. Brief description of amendments: The amendments modify the limiting condition for operation (LCO) Required Action B.2 for Technical Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' LCO Required Action C.2 for TS 3.7.4, ``Steam Generator Power Operated Relief Valves (SG PORVs),'' and LCO Required Action G.1 for TS 3.4.12, ``Low Temperature Overpressure Protection (LTOP) System.'' Specifically, the Completion Times are revised from 12 hours to 24 hours for TS LCO 3.4.10, Required Action B.2, and TS LCO 3.7.4, Required Action C.2; and from 8 hours to 12 hours for TS LCO 3.4.12, Required Action G.1. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A, Revision 1, ``Provide Consistent Completion Time to Reach MODE 4.'' Date of issuance: October 31, 2017. Effective date: These license amendments are effective as of their date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 302 (Unit 1) and 281 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A198; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Renewed Licenses and Technical Specifications. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23622). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017. No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the proposed no significant hazards consideration determination or to the license amendment request. 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 requests: January 11, 2017. Brief description of amendments: The amendments modify Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' Limiting Condition for Operation (LCO) Condition A and Required Action A.1. The proposed changes modify Condition A to expand the condition to include when one turbine driven AFW pump is inoperable in MODE 3. This expanded condition is applicable immediately following a refueling outage and only if MODE 2 has not been entered. Required Action A.1 is revised to state ``affected equipment'' as opposed to ``steam supply'' as a result of the addition of the turbine driven AFW pump to Condition A. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-340-A, Revision 3, ``Allow 7 day Completion Time for a turbine-driven AFW pump inoperable.'' Date of issuance: October 31, 2017. Effective date: These license amendments are effective as of their date of issuance and shall be [[Page 55419]] implemented within 120 days of issuance. Amendment Nos.: 304 (Unit 1) and 283 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17277A313; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23621). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017. No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the proposed no significant hazards consideration determination or to the license amendment request. Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of application for amendments: January 11, 2017. Brief description of amendments: The amendments modify Technical Specification (TS) Limiting Condition for Operation (LCO) 3.9.6, ``Residual Heat Removal (RHR) and Coolant Circulation--Low Water Level,'' to add a note which allows all RHR pumps to be secured for less than or equal to 15 minutes to support the switching of the shutdown cooling loops from one train to another. The changes are consistent with Technical Specifications 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 [Shutdown Cooling] SDC/RHR/[Decay Heat Removal] DHR loop to [be] inoperable to support testing,'' and TSTF-438-A, Revision 0, ``Clarify Exception Notes to be Consistent with the Requirement Being Excepted.'' Date of issuance: October 31, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: Unit 1--303; Unit 2--282. A publicly-available version is in ADAMS under Accession No. ML17271A034; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Renewed Licenses and Technical Specifications. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23623). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017. No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the proposed no significant hazards consideration determination or the license amendment request. Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: December 2, 2016, as supplemented by letters dated April 25, May 22, and October 2, 2017. Brief description of amendment: The amendment revised the Technical Specifications (TSs) to (1) relocate cycle-specific parameters to the Core Operating Limits Report (COLR) consistent with Technical Specification Task Force (TSTF)-339, ``Relocate TS Parameters to COLR;'' (2) delete duplicate reporting requirements in the Administrative Section of TSs consistent with TSTF-5, ``Delete Safety Limit Violation Notification Requirements,'' Revision 1; and (3) delete reference to plant procedure PLP-6, ``Technical Specification Equipment List Program and Core Operating Limits Report,'' in TSs as it pertains to the COLR. Date of issuance: November 6, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 161. A publicly-available version is in ADAMS under Accession No. ML17250A202; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-63: Amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: February 14, 2017 (82 FR 10595). The supplemental letters dated April 25, May 22, and October 2, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the 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 November 6, 2017. No significant hazards consideration comments received: No. Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington Date of amendment request: November 8, 2016, as supplemented by letter dated July 11, 2017. Brief description of amendment: The amendment would, on a one-time basis, extend the completion time from 7 days to 14 days for the Residual Heat Removal Train A subsystem to operable status associated with Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling System]--Operating''; TS 3.6.1.5, ``Residual Heat Removal (RHR) Drywell Spray''; and TS 3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool Cooling.'' This amendment will be used to support preventive maintenance, which replaces the RHR Train A subsystem's pump and motor. Date of issuance: October 30, 2017. Effective date: As of its date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 245. A publicly-available version is in ADAMS under Accession No. ML17290A127; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-21: The amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 14, 2017 (82 FR 10596). The supplemental letter dated July 11, 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 October 30, 2017. No significant hazards consideration comments received: No. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 2, York and Lancaster Counties, Pennsylvania Date of amendment request: May 19, 2017, as supplemented by letter dated August 29, 2017. [[Page 55420]] Brief description of amendment: The amendment revised the Technical Specifications to decrease the number of safety relief valves and safety valves required to be operable when operating at a power level less than or equal to 3,358 megawatts thermal. This change is applicable only to the current Cycle 22 that is scheduled to end in October 2018. Date of issuance: October 25, 2017. Effective date: As of the date of issuance and shall be implemented within 5 days. Amendment No.: 315. A publicly-available version is in ADAMS under Accession No. ML17249A151; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-44: The amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 5, 2017 (82 FR 31094). The supplemental letter dated August 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 October 25, 2017. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of application for amendments: October 27, 2016, as supplemented by the letters dated July 28, 2017, August 30, 2017, and October 19, 2017. Brief description of amendments: The amendments revised the suppression pool swell design analysis. The new analysis utilizes a different computer code and incorporates different analysis assumptions than the current analysis. The changes are necessary because the current design analysis determining the suppression pool swell response to a loss-of-coolant accident was determined to be non-conservative. These changes to the suppression pool swell design analysis do not require any changes to the LSCS Technical Specifications. Changes to the LSCS updated final safety analysis report related to changes to the suppression pool swell design analysis shall be made in accordance with 10 CFR 50.71(e) based on the NRC approval of these changes. Date of issuance: October 30, 2017. Effective date: These license amendments are effective as of the date of its issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 225 for NPF-11 and 211 for NPF-18. A publicly- available version is in ADAMS under Accession No. ML17257A304; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License Nos. NPF-11 and NPF-18: The amendments approved to revise the LSCS updated final safety analysis report related to changes to the suppression pool swell design analysis and the Licenses. Date of initial notice in Federal Register: March 8, 2017 (82 FR 13022). The supplements dated July 28, 2017, August 30, 2017, and October 19, 2017, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 30, 2017. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station (Nine Mile Point), Unit 2, Oswego County, New York Date of amendment request: December 13, 2016, as supplemented by letter dated February 17, 2017. Brief description of amendment: The amendment revised the Nine Mile Point, Unit 2, Technical Specification (TS) safety limit (SL) to increase the low pressure isolation setpoint allowable value, which will result in earlier main steam line isolation. The revised main steam line low pressure isolation capability and the revised SL are intended to ensure that Nine Mile Point, Unit 2, remains within the TS SLs in the event of a pressure regulator failure maximum demand transient. Date of issuance: October 31, 2017. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment No.: 164. A publicly-available version is in ADAMS under Accession No. ML17268A263; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: March 28, 2017 (82 FR 15381). The supplemental letter dated February 17, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 31, 2017. No significant hazards consideration comments received: No. Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: January 23, 2017, as supplemented by letter dated July 3, 2017. Brief description of amendments: The amendments revised the Technical Specifications (TSs) by limiting the MODE of applicability for the Reactor Protection System, Startup, and Operating Rate of Change of Power--High, functional unit trip. Additionally, the amendments added new Limiting Condition for Operation (LCO) 3.0.5 and relatedly modified LCO 3.0.1 and LCO 3.0.2, to provide for placing inoperable equipment under administrative control for the purpose of conducting testing required to demonstrate OPERABILITY. Date of issuance: November 2, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 243 and 194. A publicly-available version is in ADAMS under Accession No. ML17257A015; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: March 28, 2017 (82 FR 15383). The supplemental letter dated July 3, 2017, provided additional information that expanded the scope of the application as originally noticed and changed the NRC staff's original proposed no significant hazards consideration (NSHC) determination as published in the Federal Register. Accordingly, the NRC published a second proposed no significant hazards consideration determination in the Federal Register on September 12, 2017 [[Page 55421]] (82 FR 42849). This notice superseded the original notice in its entirety. It also provided an opportunity to request a hearing by November 13, 2017, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendments. The Commission's related evaluation of the amendments and final NSHC are contained in a Safety Evaluation dated November 2, 2017. No significant hazards consideration comments received: No. Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: December 21, 2016. Brief description of amendments: The amendments modify the Technical Specifications by deleting high-range noble gas effluent monitors' requirements and relocating the requirements to the Turkey Point Offsite Dose Calculation Manual. Date of issuance: October 26, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos: 277 and 272. A publicly-available version is in ADAMS under Accession No. ML17228A563. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: March 14, 2017 (82 FR 13666). The Commission's related evaluation of the amendments is contained in a safety evaluation dated October 26, 2017. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request: August 11, 2017. Brief description of amendments: The amendments request an extension to the time to achieve full compliance with 10 CFR 50.48(c), National Fire Protection Association (NPFA) 805, from November 6, 2017, to the conclusion of the FNP, Unit 1, Spring 2018 Refueling Outage (1R28). The amendments update Attachment S, ``Modification and Implementation Items''; of the previously approved NFPA-805 amendment. Date of issuance: November 1, 2017. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 215 (Unit 1) and 212 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A166; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses. Date of initial notice in Federal Register: August 29, 2017 (82 FR 41059). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 1, 2017. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 14th day of November 2017. For the Nuclear Regulatory Commission. Kathryn M. Brock, Acting Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2017-25063 Filed 11-20-17; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55401
The meeting is open to the public. Commission, Washington, DC 20555– The NRC does not routinely edit
0001. comment submissions to remove
Patrice Little Murray,
For additional direction on obtaining identifying or contact information.
Committee Management Officer. information and submitting comments, If you are requesting or aggregating
[FR Doc. 2017–25135 Filed 11–20–17; 8:45 am] see ‘‘Obtaining Information and comments from other persons for
BILLING CODE 7515–01–P Submitting Comments’’ in the submission to the NRC, then you should
SUPPLEMENTARY INFORMATION section of inform those persons not to include
this document. identifying or contact information that
NUCLEAR REGULATORY FOR FURTHER INFORMATION CONTACT: they do not want to be publicly
COMMISSION Shirley Rohrer, Office of Nuclear disclosed in their comment submission.
Reactor Regulation, U.S. Nuclear Your request should state that the NRC
[NRC–2017–0220]
Regulatory Commission, Washington, does not routinely edit comment
Biweekly Notice; Applications and DC 20555–0001; telephone: 301–415– submissions to remove such information
Amendments to Facility Operating 5411, email: Shirley.Rohrer@nrc.gov. before making the comment
Licenses and Combined Licenses SUPPLEMENTARY INFORMATION:
submissions available to the public or
Involving No Significant Hazards entering the comment into ADAMS.
Considerations I. Obtaining Information and
II. Notice of Consideration of Issuance
Submitting Comments
of Amendments to Facility Operating
AGENCY: Nuclear Regulatory
A. Obtaining Information Licenses and Combined Licenses and
Commission.
Please refer to Docket ID NRC–2017– Proposed No Significant Hazards
ACTION: Biweekly notice. Consideration Determination
0220, facility name, unit number(s),
SUMMARY: Pursuant to Section 189a.(2) plant docket number, application date, The Commission has made a
of the Atomic Energy Act of 1954, as and subject when contacting the NRC proposed determination that the
amended (the Act), the U.S. Nuclear about the availability of information for following amendment requests involve
Regulatory Commission (NRC) is this action. You may obtain publicly- no significant hazards consideration.
publishing this regular biweekly notice. available information related to this Under the Commission’s regulations in
The Act requires the Commission to action by any of the following methods: § 50.92 of title 10 of the Code of Federal
publish notice of any amendments • Federal Rulemaking Web site: Go to Regulations (10 CFR), this means that
issued, or proposed to be issued, and http://www.regulations.gov and search operation of the facility in accordance
grants the Commission the authority to for Docket ID NRC–2017–0220. with the proposed amendment would
issue and make immediately effective • NRC’s Agencywide Documents not (1) involve a significant increase in
any amendment to an operating license Access and Management System the probability or consequences of an
or combined license, as applicable, (ADAMS): You may obtain publicly- accident previously evaluated, or (2)
upon a determination by the available documents online in the create the possibility of a new or
Commission that such amendment ADAMS Public Documents collection at different kind of accident from any
involves no significant hazards http://www.nrc.gov/reading-rm/ accident previously evaluated; or (3)
consideration, notwithstanding the adams.html. To begin the search, select involve a significant reduction in a
pendency before the Commission of a ‘‘ADAMS Public Documents’’ and then margin of safety. The basis for this
request for a hearing from any person. select ‘‘Begin Web-based ADAMS proposed determination for each
This biweekly notice includes all Search.’’ For problems with ADAMS, amendment request is shown below.
notices of amendments issued, or please contact the NRC’s Public The Commission is seeking public
proposed to be issued, from October 24, Document Room (PDR) reference staff at comments on this proposed
2017 to November 6, 2017. The last 1–800–397–4209, 301–415–4737, or by determination. Any comments received
biweekly notice was published on email to pdr.resource@nrc.gov. The within 30 days after the date of
November 7, 2017. ADAMS accession number for each publication of this notice will be
document referenced (if it is available in considered in making any final
DATES: Comments must be filed by
ADAMS) is provided the first time that determination.
December 21, 2017. A request for a Normally, the Commission will not
hearing must be filed by January 22, it is mentioned in this document.
2018. • NRC’s PDR: You may examine and issue the amendment until the
purchase copies of public documents at expiration of 60 days after the date of
ADDRESSES: You may submit comments the NRC’s PDR, Room O1–F21, One publication of this notice. The
by any of the following methods (unless White Flint North, 11555 Rockville Commission may issue the license
this document describes a different Pike, Rockville, Maryland 20852. amendment before expiration of the 60-
method for submitting comments on a day period provided that its final
specific subject): B. Submitting Comments determination is that the amendment
• Federal Rulemaking Web site: Go to Please include Docket ID NRC–2017– involves no significant hazards
http://www.regulations.gov and search 0220, facility name, unit number(s), consideration. In addition, the
for Docket ID NRC–2017–0220. Address plant docket number, application date, Commission may issue the amendment
questions about NRC dockets to Carol and subject in your comment prior to the expiration of the 30-day
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Gallagher; telephone: 301–415–3463; submission. comment period if circumstances
email: Carol.Gallagher@nrc.gov. For The NRC cautions you not to include change during the 30-day comment
technical questions, contact the identifying or contact information that period such that failure to act in a
individual listed in the FOR FURTHER you do not want to be publicly timely way would result, for example in
INFORMATION CONTACT section of this disclosed in your comment submission. derating or shutdown of the facility. If
document. The NRC will post all comment the Commission takes action prior to the
• Mail comments to: May Ma, Office submissions at http:// expiration of either the comment period
of Administration, Mail Stop: OWFN–2– www.regulations.gov as well as enter the or the notice period, it will publish in
A13, U.S. Nuclear Regulatory comment submissions into ADAMS. the Federal Register a notice of
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![55404 Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices
Response: No. CFR, Section 50.69, ‘‘Risk-informed affect any Safety Limits or operating
The proposed changes reduce the number categorization and treatment of parameters used to establish the safety
of on-shift and augmented ERO positions structures, systems and components for margin. The safety margins included in
commensurate with the hazards associated analyses of accidents are not affected by the
nuclear power reactors.’’ proposed change.
with a permanently shut down and defueled
Basis for proposed no significant The regulation requires that there be no
facility. The proposed changes do not involve
installation of new equipment or hazards consideration determination: significant effect on plant risk due to any
modification of existing equipment, so that As required by 10 CFR 50.91(a), the change to the special treatment requirements
no new equipment failure modes are licensee has provided its analysis of the for SSCs and that the SSCs continue to be
introduced. Also, the proposed changes do issue of no significant hazards capable of performing their design basis
not result in a change to the way that the consideration, which is presented functions, as well as to perform any beyond
equipment or facility is operated so that no below: design basis functions consistent with the
new accident initiators are created. categorization process and results.
1. Does the proposed amendment involve Therefore, the proposed change does not
Therefore, the proposed amendment does
a significant increase in the probability or involve a significant reduction in a margin of
not create the possibility of a new or different
consequences of an accident previously safety.
kind of accident from any accident
evaluated?
previously evaluated. The NRC staff has reviewed the
Response: No.
3. Does the proposed amendment involve The proposed change will permit the use licensee’s analysis and, based on this
a significant reduction in a margin of safety? of a risk-informed categorization process to
Response: No.
review, it appears that the three
modify the scope of SSCs [structures, standards of 10 CFR 50.92(c) are
Margin of safety is associated with systems, and components] subject to NRC
confidence in the ability of the fission satisfied. Therefore, the NRC staff
[Nuclear Regulatory Commission] special
product barriers (i.e., fuel cladding, reactor treatment requirements and to implement
proposes to determine that the
coolant system pressure boundary, and alternative treatments per the regulations. amendment request involves no
containment structure) to limit the level of The process used to evaluate SSCs for significant hazards consideration.
radiation dose to the public. The proposed changes to NRC special treatment Attorney for licensee: Tamra Domeyer,
changes are associated with the PNP SEP and requirements and the use of alternative Associate General Counsel, Exelon
do not impact operation of the plant or its requirements ensures the ability of the SSCs Nuclear, 4300 Winfield Road,
response to transients or accidents. The to perform their design function. The Warrenville, IL 60555.
change does not affect the Technical potential change to special treatment NRC Branch Chief: David J. Wrona.
Specifications. The proposed changes do not requirements does not change the design and
involve a change in the method of plant operation of the SSCs. As a result, the Exelon Generation Company, LLC and
operation, and no accident analyses will be proposed change does not significantly affect PSEG Nuclear LLC, Docket Nos. 50–277
affected by the proposed changes. Safety any initiators to accidents previously and 50–278, Peach Bottom Atomic
analysis acceptance criteria are not affected evaluated or the ability to mitigate any Power Station, Units 2 and 3, York and
by the proposed changes. The revised PNP accidents previously evaluated. The
SEP will continue to provide the necessary Lancaster Counties, Pennsylvania
consequences of the accidents previously
response staff with the proposed changes. evaluated are not affected because the Date of amendment request: August
Therefore, the proposed amendment does mitigation functions performed by the SSCs 30, 2017, as supplemented by letter
not involve a significant reduction in a assumed in the safety analysis are not being dated October 24, 2017. Publicly-
margin of safety. modified. The SSCs required to safely shut available versions are in ADAMS under
The NRC staff has reviewed the down the reactor and maintain it in a safe Accession Nos. ML17243A014 and
shutdown condition following an accident
licensee’s analysis and, based on this ML17297B521, respectively.
will continue to perform their design
review, it appears that the three functions.
Description of amendment request:
standards of 10 CFR 50.92(c) are Therefore, the proposed change does not The amendments would modify the
satisfied. Therefore, the NRC staff involve a significant increase in the licensing basis by the addition of a
proposes to determine that the probability or consequences of an accident license condition to allow for the
amendment request involves no previously evaluated. implementation of the provisions of 10
significant hazards consideration. 2. Does the proposed amendment create CFR 50.69, ‘‘Risk-informed
Attorney for licensee: William Dennis, the possibility of a new or different kind of categorization and treatment of
accident from any previously evaluated? structures, systems and components for
Assistant General Counsel, Entergy Response: No.
Nuclear Operations, Inc., 440 Hamilton The proposed change will permit the use
nuclear power reactors.’’
Ave., White Plains, NY 10601. of a risk-informed categorization process to Basis for proposed no significant
NRC Branch Chief: Douglas A. modify the scope of SSCs subject to NRC hazards consideration determination:
Broaddus. special treatment requirements and to As required by 10 CFR 50.91(a), the
implement alternative treatments per the licensee has provided its analysis of the
Exelon Generation Company (EGC), regulations. The proposed change does not issue of no significant hazards
LLC, Docket Nos. STN 50–456 and STN change the functional requirements, consideration, which is presented
50–457, Braidwood Station, Units 1 and configuration, or method of operation of any below, with NRC staff edits shown in
2, Will County, Illinois and Docket Nos. SSC. Under the proposed change, no square brackets:
STN 50–454 and STN 50–455, Byron additional plant equipment will be installed.
Station, Unit Nos. 1 and 2, Ogle County, Therefore, the proposed change does not 1. Does the proposed change involve a
create the possibility of a new or different significant increase in the probability or
Illinois consequences of an accident previously
kind of accident from any accident
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Date of amendment request: previously evaluated. evaluated?
September 1, 2017. A publicly-available 3. Does the proposed amendment involve Response: No.
version is in ADAMS under Accession a significant reduction in a margin of safety? The proposed change will permit the use
No. ML17244A093. Response: No. of a risk-informed categorization process to
The proposed change will permit the use modify the scope of [structures, systems, and
Description of amendment request: of a risk-informed categorization process to components] SSCs subject to NRC special
The amendments would modify the modify the scope of SSCs subject to NRC treatment requirements and to implement
licensing basis by the addition of a special treatment requirements and to alternative treatments per the regulations.
license condition to allow for the implement alternative treatments per the The process used to evaluate SSCs for
implementation of the provisions of 10 regulations. The proposed change does not changes to NRC special treatment
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55407
The proposed license amendments modify standards of 10 CFR 50.92(c) are mechanism, malfunction or accident initiator
the Turkey Point TS by relocating the satisfied. Therefore, the NRC staff not already considered in the design and
Explosive Gas Monitoring Instrumentation, proposes to determine that the licensing basis.
Explosive Gas Mixture, Gas Decay Tanks and Therefore, the proposed change does not
amendment request involves no
Standby Feedwater System requirements to create the possibility of a new or different
licensee controlled documents, by relatedly
significant hazards consideration. kind of accident from any accident
modifying the AFW System requirements Attorney for licensee: William S. previously evaluated.
and by establishing a Gas Decay Tank Blair, Managing Attorney—Nuclear, 3. Does the proposed change involve a
Explosive Gas and Radioactivity Monitoring Florida Power & Light Company, 700 significant reduction in a margin of safety?
Program. The proposed changes are Universe Blvd., MS LAW/JB, Juno Response: No.
administrative in nature and do not alter any Beach, FL 33408–0420. Although the proposed change will
plant equipment or the manner in which NRC Branch Chief: Undine Shoop. decrease the required supply of nitrogen for
plant equipment is operated and maintained. the ADS accumulators from 100 days to 30
All equipment limitations, applicable NextEra Energy Duane Arnold, LLC, days, the assessment above has shown that
methodologies and surveillances are Docket No. 50–331, Duane Arnold the reactor would be depressurized within 3
maintained by the proposed changes. In Energy Center (DAEC), Linn County, days following any postulated accident or
addition, the proposed changes to the AFW Iowa event that would create a hostile
System requirements enhance plant safety. environment in the drywell. Once initial
Date of amendment request: depressurization is completed, long term core
As such, the proposed changes cannot affect
the initiators, the likelihood or the expected September 5, 2017. A publicly-available cooling can be assured without ADS.
outcomes of any analyzed accidents. version is in ADAMS under Accession Therefore, the proposed change will not
Therefore, facility operation in accordance No. ML17248A284. involve a significant reduction in a margin of
with the proposed changes would not Description of amendment request: safety.
involve a significant increase in the The proposed amendment would revise
The NRC staff has reviewed the
probability or consequences of an accident DAEC Technical Specifications 3.5.1,
previously evaluated. licensee’s analysis and, based on this
‘‘ECCS [emergency core cooling system]-
2. Does the proposed amendment create review, it appears that the three
Operating.’’ The proposed change
the possibility of a new or different kind of standards of 10 CFR 50.92(c) are
would decrease the nitrogen supply
accident from any accident previously satisfied. Therefore, the NRC staff
evaluated?
requirement for the Automatic
proposes to determine that the
Response: No. Depressurization System (ADS) in
amendment request involves no
The proposed license amendments modify Surveillance Requirement (SR) 3.5.1.3
significant hazards consideration.
the Turkey Point TS by relocating the from 100 days to 30 days.
Basis for proposed no significant Attorney for licensee: William Blair,
Explosive Gas Monitoring Instrumentation,
Explosive Gas Mixture, Gas Decay Tanks and hazards consideration determination: P.O. Box 14000, Juno Beach, FL 33408–
Standby Feedwater System requirements to As required by 10 CFR 50.91(a), the 0420.
licensee controlled documents, by relatedly licensee has provided its analysis of the NRC Branch Chief: David J. Wrona.
modifying the AFW System requirements issue of no significant hazards NextEra Energy Duane Arnold, LLC,
and by establishing a Gas Decay Tank consideration, which is presented
Explosive Gas and Radioactivity Monitoring
Docket No. 50–331, Duane Arnold
Program. The proposed changes neither
below: Energy Center, Linn County, Iowa
install or remove plant equipment nor alter 1. Does the proposed change involve a Date of amendment request: August
any plant equipment design, configuration, significant increase in the probability or 31, 2017. A publicly-available version is
or method of operation. Hence, no new consequences of an accident previously
evaluated?
in ADAMS under Accession No.
failure mechanisms are introduced as a result
of the proposed changes. Response: No. ML17243A469.
Therefore, the proposed changes do not The proposed change modifies a SR for Description of amendment request:
create the possibility of a new or different verification of the nitrogen supply for the The proposed amendment would
kind of accident from any previously ADS accumulators. Accidents are initiated by modify the licensing basis by the
evaluated. the malfunction of plant equipment, or the addition of a license condition to allow
3. Does the proposed amendment involve catastrophic failure of plant structures, for the implementation of the provisions
a significant reduction in a margin of safety? systems or components. The performance of of 10 CFR, part 50.69, ‘‘Risk-Informed
Response: No. this surveillance is not a precursor to any
Categorization and Treatment of
The proposed license amendments modify accident previously evaluated and does not
change the manner in which the ADS Structures, Systems, and Components
the Turkey Point TS by relocating the
Explosive Gas Monitoring Instrumentation, operates. Technical evaluation of the change (SSCs) for Nuclear Power Reactors.’’
Explosive Gas Mixture, Gas Decay Tanks and concluded that a 30-day nitrogen supply is Basis for proposed no significant
Standby Feedwater System requirements to more than adequate to ensure that the reactor hazards consideration determination:
licensee controlled documents, by relatedly is depressurized, so the consequences of an As required by 10 CFR 50.91(a), the
modifying the AFW System requirements accident remain unchanged. licensee has provided its analysis of the
and by establishing a Gas Decay Tank Therefore, the proposed change does not issue of no significant hazards
Explosive Gas and Radioactivity Monitoring involve a significant increase in the
consideration, which is presented
Program. The proposed changes neither probability or consequence of a previously
evaluated accident. below:
involve changes to safety analyses
assumptions, safety limits, or limiting safety 2. Does the proposed change create the 1. Does the proposed change involve a
system settings nor adversely impact plant possibility of a new or different kind of significant increase in the probability or
accident from any accident previously
asabaliauskas on DSKBBXCHB2PROD with NOTICES
operating margins or the reliability of consequences of an accident previously
equipment credited in safety analyses. evaluated? evaluated?
Therefore, operation of the facility in Response: No. Response: No.
accordance with the proposed changes will The proposed change does not involve The proposed change will permit the use
not involve a significant reduction in the physical alterations to the plant. No new or of a risk-informed categorization process to
margin of safety. different type of equipment will be installed, modify the scope of SSCs subject to NRC
and there are no physical modifications special treatment requirements and to
The NRC staff has reviewed the required to existing installed equipment implement alternative treatments per the
licensee’s analysis and, based on this associated with the proposed change. The regulations. The process used to evaluate
review, it appears that the three proposed change does not create any failure SSCs for changes to NRC special treatment
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55409
Basis for proposed no significant review, it appears that the three demonstrate the integrity of the reactor
hazards consideration determination: standards of 10 CFR 50.92(c) are containment exist to ensure the plant’s
As required by 10 CFR 50.91(a), the satisfied. Therefore, the NRC staff ability to mitigate the consequences of an
accident, and do not involve the prevention
licensee has provided its analysis of the proposes to determine that the
or identification of any precursors of an
issue of no significant hazards amendment request involves no accident.
consideration, which is presented significant hazards consideration. Therefore, the proposed changes do not
below: Attorney for licensee: Jeffrie J. Keenan, involve a significant increase in the
1. Does the proposed change involve a PSEG Nuclear LLC—N21, P.O. Box 236, probability or consequences of an accident
significant increase in the probability or Hancocks Bridge, NJ 08038. previously evaluated.
consequences of an accident previously NRC Branch Chief: James G. Danna. The integrity of the reactor containment is
evaluated? subject to two types of failure mechanisms
Response: No. South Carolina Electric & Gas Company, which can be categorized as (1) activity based
The proposed changes to the TS will not South Carolina Public Service and (2) time based. Activity based failure
alter the way any structure, system, or Authority, Docket No. 50–395, Virgil C. mechanisms are defined as degradation due
component (SSC) functions, and will not Summer Nuclear Station, Unit No. 1, to system and/or component modifications or
alter the manner in which the plant is Fairfield County, South Carolina maintenance. Local leak rate test
operated. The proposed changes do not alter requirements and administrative controls
the design of any SSC. The relocation of the Date of amendment request: October such as configuration management and
RCS PIV valve lists from the TS to the TRM 6, 2017. A publicly-available version is procedural requirements for system
is an administrative change. Future revisions in ADAMS under Accession No. restoration ensure that containment integrity
to the TRM are subject to 10 CFR 50.59. ML17279A715. is not degraded by plant modifications or
Therefore the probability of an accident maintenance activities. The updated Pa value
Description of amendment request:
previously evaluated is not significantly reflects the updated mass and energy release
The proposed amendment would and containment response calculations,
increased.
The proposed changes do not alter the RCS increase the Integrated Leak Rate Test ensuring a sound technical basis for the local
PIV leakage limits contained in the TS nor do (ILRT) Peak Calculated Containment and integrated leakage tests.
they alter the frequency for testing of the RCS Internal Pressure, Pa, listed in Technical To mitigate time-based mechanisms, the
PIV. Therefore, the consequences of an Specification (TS) 6.8.4.g, ‘‘Containment design and construction requirements of the
accident previously evaluated are not Leakage Rate Testing Program,’’ to containment itself combined with the
increased. remove the reference to Regulatory containment inspections performed in
Therefore, these proposed changes do not Guide (RG) 1.163, ‘‘Performance-Based accordance with ASME [American Society of
represent a significant increase in the Mechanical Engineers], Section XI and the
probability or consequences of an accident Containment Leak Test Program,’’ dated
Maintenance Rule serve to provide a high
previously evaluated. September 1995 and ANSI/ANS degree of assurance that the containment will
2. Does the proposed change create the (American National Standards Institute/ not degrade in a manner that is detectable
possibility of a new or different kind of American Nuclear Society)—56.8–2002, only by a Type A test. The change to the Pa
accident from any accident previously ‘‘Containment System Leakage Testing value is less than 1 psid [per square inch
evaluated? Requirements,’’ and to replace the differential]. Radiological consequences will
Response: No. reference of Nuclear Energy Institute continue to be evaluated at the Technical
The proposed changes do not involve a (NEI) 94–01, Revision 3–A, ‘‘Industry Specification allowed leakage, La [allowed
modification to the physical configuration of leakage] of 0.20 percent by weight of air,
the plant or changes in the methods Guideline for Implementing
Performance-Based option of 10 CFR which will not be increased despite the
governing normal plant operation. The increase in Pa. As described in Section 3.5,
proposed changes will not impose any new part 50, Appendix J,’’ with NEI 94–01, past leakage testing yielded values well
or different requirement or introduce a new Revision 2–A. under La. Based on the above, neither the
accident initiator, accident precursor, or Basis for proposed no significant reference changes nor the Pa change involves
malfunction mechanism. The proposed hazards consideration determination: a significant increase in the consequences of
changes are administrative in nature. As required by 10 CFR 50.91(a), the an accident previously evaluated.
Therefore, the proposed changes do not licensee has provided its analysis of the 2. Does the proposed change create the
create the possibility of a new or different possibility of a new or different kind of
kind of accident from any accident issue of no significant hazards
consideration, which is presented accident from any accident previously
previously evaluated. evaluated?
3. Do the proposed changes involve a below, with NRC staff edits in square
Response: No.
significant reduction in a margin of safety? brackets: The proposed changes involve removal of
Response: No. 1. Does the proposed change involve a RG 1.163 and ANSl/ANS–56.8–2002
The proposed changes to the RCS PIV TS significant increase in the probability or references, replacement of NEI 94–01,
are administrative in nature. The proposed consequences of an accident previously Revision 3–A with NEI 94–01, Revision 2–A,
changes do not alter the RCS PIV leakage evaluated? and an increase in the Pa value for
limits contained in the TS nor do they alter Response: No. containment leakage testing. The reactor
the frequency for testing of the RCS PIV. The
The proposed changes involve removal of containment and the testing requirements
proposed changes will not result in changes
RG 1.163 and ANSl/ANS–56.8–2002 invoked to periodically demonstrate the
to system design or setpoints that are
references, replacement of NEI 94–01, integrity of the reactor containment exist to
intended to ensure timely identification of
Revision 3–A with NEI 94–01, Revision 2–A, ensure the plant’s ability to mitigate the
plant conditions that could be precursors to
and an increase in the Pa [Peak Calculated consequences of an accident. There are not
accidents or potential degradation of accident
Containment Internal Pressure] value for any accident initiators or precursors affected
mitigation systems.
asabaliauskas on DSKBBXCHB2PROD with NOTICES
containment leakage testing. The activity by the revision. The proposed TS change
The proposed amendment will not result
does not involve a physical change to the does not involve a physical change to the
in a design basis or safety limit being
plant or a change in the manner in which the plant or the manner in which the plant is
exceeded or altered. Therefore, since the
proposed changes do not impact the response plant is operated or controlled. The operated or controlled.
of the plant to a design basis accident, the containment is designed to provide an Therefore, the proposed TS change does
proposed changes do not involve a essentially leak tight barrier against the not create the possibility of a new or different
significant reduction in a margin of safety. uncontrolled release of radioactivity to the kind of accident from any accident
environment for postulated accidents. As previously evaluated.
The NRC staff has reviewed the such, the reactor containment itself and the 3. Does the proposed change involve a
licensee’s analysis and, based on this testing requirements invoked to periodically significant reduction in a margin of safety?
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![55410 Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices
Response: No. Description of amendment request: probability or consequences of an accident
The proposed changes involve removal of The amendment request proposes to previously evaluated.
RG 1.163 and ANSl/ANS–56.8–2002 revise Technical Specification Section 2. Does the proposed amendment create
references, replacement of NEI 94–01, 1.1 (TS), Definition of Actuation Logic the possibility of a different kind of accident
Revision 3–A with NEI 94–01, Revision 2–A, from any accident previously evaluated?
and an increase in the Pa value for
Test, by adding a new TS Section 1.1 Response: No.
containment leakage testing. The proposed Definition of Actuation Logic Output With respect to any new or different kind
TS change does not involve a physical Test (ALOT), revising existing of accident, there are no proposed design
change to the plant or a change in the manner Surveillance Requirements 3.3.15.1 and changes nor are there any changes in the
in which the plant is operated or controlled. 3.3.16.1 and adding new Surveillance method by which any safety-related plant
Using the same analysis methodology as Requirements 3.3.15.2 and 3.3.16.2 to SSC performs its specified safety function.
described in WCAP–10325–P–A implement the new ALOT. This The proposed change will not affect the
[Westinghouse LOCA [loss-of-accident submittal requests approval of the normal method of plant operation or change
coolant] Mass and Energy Release Model for license amendment that is necessary to any operating parameters. No equipment
Containment Design], the updated mass and performance requirements will be affected.
energy release and containment response
implement these changes. The proposed change will not alter any
analyses corrected input errors identified in Basis for proposed no significant assumptions made in the safety analyses.
the NSALs [Westinghouse Nuclear Safety hazards consideration determination: The proposed change revises the frequency
Advisory Letters] described previously. As As required by 10 CFR 50.91(A), of testing certain portions of the PMS logic
shown in Figure 1 [October 6, 2017, licensee has provided its analysis of the circuitry. The proposed change does not
submittal], the correction of these errors issue on no significant hazards involve a physical modification of the plant.
resulted in a slightly higher predicted peak consideration determination, which is No new accident scenarios, transient
pressure than that of the current licensing presented below: precursors, failure mechanisms, or limiting
basis but does not pose a significant single failures will be introduced as a result
challenge to the design limit. 1. Does the proposed amendment involve of this amendment. There will be no adverse
The specific requirements and conditions a significant increase in the probability or effect or challenges imposed on any safety-
of the Primary Containment Leak Rate consequences of an accident previously related system as a result of this amendment.
Testing Program, as defined in the Technical evaluated? Therefore, the proposed amendment does
Specifications, exist to ensure that the degree Response: No. not create the possibility of a new or different
of reactor containment structural integrity There are no design changes associated kind of accident from any accident
and leak-tightness that is considered in the with the proposed amendment. All design, previously evaluated.
plant safety analysis is maintained. The material, and construction standards that 3. Does the proposed amendment involve
overall containment leak rate limit specified were applicable prior to this amendment a significant reduction in a margin of safety?
by the Technical Specification is maintained. request will continue to be applicable. Response: No.
The containment inspections performed in The [Processor Module Self-Diagnostic The existing ACTUATION LOGIC TEST
accordance with ASME, Section XI and the (PMS)] will continue to function in a manner Surveillance Requirements are revised such
Maintenance Rule serve to provide a high consistent with the plant design basis. There that different portions of the PMS logic
degree of assurance that the containment will will be no changes to the PMS operating circuitry are tested on appropriate
not degrade in a manner that is detectable limits. The existing ACTUATION LOGIC surveillance test frequencies. The reliability
only by Type A testing. The combination of TEST Surveillance Requirements are revised of the PMS is such that not testing the
these factors ensures that the margin of safety such that different portions of the PMS logic Component Interface Module (CIM) logic and
that is in plant safety analysis is maintained. circuitry are tested on appropriate driver output circuits when the reactor is at
The design, operation, testing methods and surveillance test frequencies. power will have a net positive impact on
acceptance criteria for Type A, B, and C The proposed change will not adversely Engineered Safety Feature Actuation System
containment leakage tests specified in affect accident initiators or precursors or (ESFAS) availability. There will be a
applicable codes and standards will continue adversely alter the design assumptions, reduction in the potential for challenges to
to be met. conditions, and configuration of the facility, the safety systems, coupled with less time
Therefore, the proposed TS change does or the manner in which the plant is operated that the safety systems are unavailable.
not involve a significant reduction in a and maintained, with respect to such There will be no effect on those plant
margin of safety. initiators or precursors. systems necessary to effect the
The proposed changes will not alter the accomplishment of protection functions.
The NRC staff has reviewed the ability of structures, systems, and No instrument setpoints or system
licensee’s analysis and, based on this components (SSCs) to perform their specified response times are affected. None of the
review, it appears that the three safety functions to mitigate the consequences acceptance criteria for any accident analysis
standards of 10 CFR 50.92(c) are of an initiating event within the assumed will be changed.
satisfied. Therefore, the NRC staff acceptance limits. The proposed change will have no impact
proposes to determine that the Accident analysis acceptance criteria will on the radiological consequences of a design
amendment request involves no continue to be met with the proposed basis accident.
changes. The proposed changes will not Therefore, the proposed amendment does
significant hazards consideration. affect the source term, containment isolation,
Attorney for licensee: Kathryn M. not involve a significant increase in the
or radiological release assumptions used in probability or consequences of an accident
Sutton, Morgan, Lewis & Bockius LLP, evaluating the radiological consequences of previously evaluated.
1111 Pennsylvania Avenue NW., any accident previously evaluated. The
Washington, DC 20004. proposed changes will not alter any The NRC staff has reviewed the
NRC Branch Chief: Michael T. assumptions or change any mitigation actions licensee’s analysis and, based on this
Markley. in the radiological consequence evaluations review, it appears that the three
asabaliauskas on DSKBBXCHB2PROD with NOTICES
in the Updated Final Safety Analysis Report standards of 10 CFR 50.92(c) are
Southern Nuclear Operating Company, (UFSAR). satisfied. Therefore, the NRC staff
Docket Nos. 52–025 and 52–026, Vogtle The applicable radiological dose proposes to determine that the
Electric Generating Plant, Units 3 and 4, acceptance criteria will continue to be met. amendment request involves no
Burke County, Georgia The proposed change revises the frequency
of testing certain portions of the PMS logic
significant hazards consideration.
Date of amendment request: July 28, circuitry, but does not physically alter any Attorney for licensee: M. Stanford
2017. A publicly-available version is in safety-related systems. Blanton, Balch & Bingham LLP, 1710
ADAMS under Accession No. Therefore, the proposed amendment does Sixth Avenue North, Birmingham, AL
ML17209A759. not involve a significant increase in the 35203–2015.
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55411
NRC Branch Chief: Jennifer Dixon- might initiate a new or different kind of NRC Branch Chief: Jennifer Dixon-
Herrity. accident, or alter any [structure, system, and Herrity.
component (SSC)] such that a new accident
Southern Nuclear Operating Company, initiator or initiating sequence of events is Southern Nuclear Operating Company,
Docket Nos. 52–025 and 52–026, Vogtle created. The proposed changes do not Docket Nos. 52–025 and 52–026, Vogtle
Electric Generating Plant, Units 3 and 4, adversely affect the physical design and Electric Generating Plant, Units 3 and 4,
Burke County, Georgia operation of the [in-containment refueling Burke County, Georgia
water storage tank (IRWST)] injection, drain,
Date of amendment request: August containment recirculation, and fourth-stage Date of amendment request: October
18, 2017. A publicly-available version is [automatic depressurization system (ADS)] 6, 2017. A publicly-available version is
in ADAMS under Accession No. valves, including as-installed inspections, in ADAMS under Accession No.
ML17230A365. and maintenance requirements, as described ML17279A084.
Description of amendment request: in the UFSAR. Therefore, the operation of the
Description of amendment request:
The requested amendment proposes to IRWST injection, drain, containment
recirculation, and fourth-stage ADS valves is The amendment request proposes to
depart from approved AP1000 Design depart from Tier 2 information in the
Control Document (DCD) Tier 2 not adversely affected. These proposed
changes do not adversely affect any other Updated Final Safety Analysis Report
information (text) and involved Tier 2* (UFSAR) (which includes the plant-
SSC design functions or methods of
information (as incorporated into the operation in a manner that results in a new specific Design Control Document
Updated Final Safety Analysis Report failure mode, malfunction, or sequence of (DCD) Tier 2 information) and involves
(UFSAR) as plant-specific DCD events that affect safety-related or nonsafety- related changes to plant-specific Tier 1
information). related equipment. Therefore, this activity information, with corresponding
This amendment request proposes does not allow for a new fission product changes to the associated combined
increasing the design pressure of the release path, result in a new fission product
license (COL) Appendix C information.
main steam (MS) isolation valve (MSIV) barrier failure mode, or create a new
sequence of events that result in significant Pursuant to the provisions of 10 CFR
compartments from 6.0 to 6.5 psi and 52.63(b)(1), an exemption from elements
proposes other changes to the licensing fuel cladding failures.
Therefore, the proposed amendment does of the design as certified in the 10 CFR
basis regarding descriptions of the MSIV not create the possibility of a new or different part 52, Appendix D, design
compartments. kind of accident from any accident certification rule is also requested for
Basis for proposed no significant previously evaluated. the plant-specific DCD Tier 1 material
hazards consideration determination: 3. Does the proposed amendment involve departures. Specifically, the requested
As required by 10 CFR 50.91(a), the a significant reduction in a margin of safety? amendment proposes to depart from
licensee has provided its analysis of the Response: No.
Tier 2 information in UFSAR
issue of no significant hazards The margin of safety for the design of the
auxiliary building is maintained through Subsection 8.3.2.4 describing raceway
consideration, which is presented below
continued use of approved codes and and cable routing criteria and hazard
with Nuclear Regulatory Commission
standards as stated in the UFSAR, and protection, and involves related changes
(NRC) staff’s edits in square brackets:
adherence to the assumptions used in the to plant-specific Tier 1 Table 3.3–6,
1. Does the proposed amendment involve analyses of this structure and the events inspections, tests, analyses, and
a significant increase in the probability or associated with this structure. The auxiliary acceptance criteria information, with
consequences of an accident previously building continues to be a seismic Category
evaluated?
corresponding changes to the associated
I building with all current structural safety COL Appendix C information.
Response: No. margins maintained. The 3-hour fire rating
The proposed changes do not adversely requirements for the impacted auxiliary Basis for proposed no significant
affect the operation of any structures, building walls are maintained. The hazards consideration determination:
systems, and components inside or outside equipment housed in the main steam As required by 10 CFR 50.91(a), the
the auxiliary building that could initiate or isolation valve compartments continue to be licensee has provided its analysis of the
mitigate abnormal events, e.g., accidents, environmentally qualified for their intended issue of no significant hazards
anticipated operational occurrences, service in accordance with the approved consideration, which is presented below
earthquakes, floods, tornado missiles, and codes and standards stated within the
turbine missiles, or their safety or design
with NRC staff edits in square brackets:
UFSAR. Thus, the requested changes will not
analyses, evaluated in the UFSAR. The adversely affect any safety-related 1. Does the proposed amendment involve
changes do not adversely affect any design equipment, design code, function, design a significant increase in the probability or
function of the auxiliary building or the analysis, safety analysis input or result, or consequences of an accident previously
structures, systems, and components design/safety margin. No safety analysis or evaluated?
contained therein. The ability of the affected design basis acceptance limit/criterion is Response: No.
auxiliary building main steam isolation valve challenged or exceeded by the requested Changes 1, 3 and 4 are clarifications only
compartments and adjacent rooms, including change, thus, no margin of safety is reduced. and do not represent a change to the
the main control room, to withstand the Therefore, the proposed change does not minimum required separation distance
pressurization effects from the postulated involve a significant reduction in a margin of between raceways. Change 2 reduces the
pipe ruptures is not adversely affected by the safety. required separation distances between
increase in design pressure, since the raceways from those documented in
structures, systems, and components therein The NRC staff has reviewed the [Institute of Electrical and Electronics
remain qualified for this service. licensee’s analysis and, based on this Engineers (IEEE)] 384–1981. These reduced
Therefore, the proposed activity does not review, it appears that the three separation distances are based on specific
asabaliauskas on DSKBBXCHB2PROD with NOTICES
involve a significant increase in the standards of 10 CFR 50.92(c) are tests performed on the specified raceway
probability or consequences of an accident satisfied. Therefore, the NRC staff configurations, and the recommendations
previously evaluated. proposes to determine that the from those tests contained in the associated
2. Does the proposed amendment create report. The NRC staff previously reviewed
amendment request involves no
the possibility of a new or different kind of the descriptions of the ten tests documented
accident from any accident previously significant hazards consideration. in this report, including the ones applicable
evaluated? Attorney for licensee: M. Stanford to the existing UFSAR exceptions, and
Response: No. Blanton, Balch & Bingham LLP, 1710 concluded that they were acceptable, as
The proposed changes do not affect the Sixth Avenue North, Birmingham, AL documented in NUREG–1793, ‘‘Final Safety
operation of any systems or equipment that 35203–2015. Evaluation Report Related to Certification of
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55415
limiting drain time or to be in service Tennessee Valley Authority (TVA), the ASME Code [American Society of
depending on the limiting drain time. Should Docket Nos. 50–259, 50–260, and 50– Mechanical Engineers Boiler and Press
those systems be unable to be placed into 296, Browns Ferry Nuclear Plant (BFN), Vessel Code] Section Xl (Subsection IWE)
service, the consequences are no different and Maintenance Rule monitoring (10 CFR
Units 1, 2, and 3, Limestone County,
than if those systems were unable to perform 50.65, ‘‘Requirements for Monitoring the
Alabama Effectiveness of Maintenance at Nuclear
their function under the current TS
requirements. The event of concern under the Date of amendment request: August Power Plants’’) are performed in order to
15, 2017. A publicly-available version is identify indications of containment
current requirements and the proposed
in ADAMS under Accession No. degradation that could affect that leak
changes are an unexpected draining event.
tightness. Types B and C testing required by
The proposed changes do not create new ML17228A490. TSs will identify any containment opening
failure mechanisms, malfunctions, or Description of amendment request: such as valves that would otherwise be
accident initiators that would cause a The amendments would revise the BFN, detected by the Type A tests. These factors
draining event or a new or different kind of Units 1, 2, and 3 Technical show that a Type A test interval extension
accident not previously evaluated or Specification (TS) 5.5.12, ‘‘Primary will not represent a significant increase in
included in the design and licensing bases. Containment Leakage Rate Testing the consequences of an accident.
The administrative update to delete Program,’’ by adopting Nuclear Energy The proposed amendment involves
expired completion time notes is purely Institute (NEI) 94–01, Revision 3–A, changes to the BFN, Units 1, 2, and 3, 10 CFR
administrative in nature. 50 Appendix J Testing Program Plan. The
‘‘Industry Guideline for Implementing proposed amendment does not involve a
Therefore, the proposed changes do not
Performance-Based Option of 10 CFR physical change to the plant or a change in
create the possibility of a new or different
part 50, Appendix J,’’ as the the manner in which the units are operated
kind of accident from any previously
evaluated.
implementation document for the or controlled. The primary containment
3. Does the proposed change involve a performance-based Option B of 10 CFR function is to provide an essentially leak
part 50, Appendix J. The proposed tight barrier against the uncontrolled release
significant reduction in a margin of safety?
changes permanently extend the Type A of radioactivity to the environment for
Response: No.
containment integrated leak rate testing postulated accidents. As such, the
The proposed changes replace existing TS containment itself and the testing
requirements related to OPDRVs with new (ILRT) interval from 10 years to 15 years
requirements to periodically demonstrate the
requirements on RPV WIC. The current and the Type C local leakage rate testing integrity of the containment exist to ensure
requirements do not have a stated safety basis (LLRT) intervals from 60 months to 75 the plant’s ability to mitigate the
and no margin of safety is established in the months. consequences of an accident, and do not
licensing basis. The safety basis for the new Basis for proposed no significant involve any accident precursors or initiators.
requirements is to protect Safety Limit hazards consideration determination: Therefore, the probability of occurrence of an
2.1.1.3. New requirements are added to As required by 10 CFR 50.91(a), the accident previously evaluated is not
determine the limiting time in which the licensee has provided its analysis of the significantly increased by the proposed
RPV water inventory could drain to the top issue of no significant hazards amendment.
of the fuel in the reactor vessel should an The proposed amendment adopts the NRC-
consideration, which is presented
unexpected draining event occur. Plant accepted guidelines of NEI 94–01, Revision
below: 3–A, for development of the BFN, Units 1, 2,
configurations that could result in lowering
the RPV water level to the TAF within one 1. Does the proposed amendment involve and 3, performance-based leakage testing
hour are now prohibited. New escalating a significant increase in the probability or program. Implementation of these guidelines
compensatory measures based on the limiting consequence of an accident previously continues to provide adequate assurance that
drain time replace the current controls. The evaluated? during design basis accidents, the primary
proposed TS establish a safety margin by Response: No. containment and its components will limit
The proposed revision to TS 5.5.12 leakage rates to less than the values assumed
providing defense-in-depth to ensure that the
changes the testing period to a permanent 15- in the plant safety analyses. The potential
Safety Limit is protected and to protect the
year interval for Type A testing (10 CFR part consequences of extending the ILRT interval
public health and safety. While some less
50, Appendix J, Option B, ILRT) and a 75- from 10 years to 15 years have been
restrictive requirements are proposed for month interval for Type C testing (10 CFR evaluated by analyzing the resulting changes
plant configurations with long calculated part 50, Appendix J, Option B, LLRT). The in risk. The increase in risk in terms of
drain times, the overall effect of the change current Type A test interval of 10 years person-rem [roentgen equivalent man] per
is to improve plant safety and to add safety would be extended to 15 years from the last year resulting from design basis accidents
margin. Type A test. The proposed extension to Type was estimated to be very small, and the
The administrative update to delete A testing does not involve a significant increase in the LERF [large early release
expired completion time notes is purely increase in the consequences of an accident frequency] resulting from the proposed
administrative in nature. because research documented in NUREG– change was determined to be within the
Therefore, the proposed changes do not 1493, ‘‘Performance-Based Containment guidelines published in NRC RG [Regulatory
involve a significant reduction in a margin of System Leakage Testing Requirements’’ Guide] 1.174. Additionally, the proposed
safety. [‘‘Performance-Based Containment Leak-Test change maintains defense-in-depth by
Program’’], September 1995, has found that, preserving a reasonable balance among
The NRC staff has reviewed the generically, very few potential containment prevention of core damage, prevention of
licensee’s analysis and, based on this leakage paths are not identified by Type B containment failure, and consequence
review, it appears that the three and C tests. NUREG–1493 concluded that mitigation. TVA has determined that the
standards of 10 CFR 50.92(c) are reducing the Type A testing frequency to one increase in CCFP [conditional containment
satisfied. Therefore, the NRC staff per 20 years was found to lead to an failure probability] due to the proposed
change would be very small.
asabaliauskas on DSKBBXCHB2PROD with NOTICES
imperceptible increase in risk. A high degree
proposes to determine that the of assurance is provided through testing and Based on the above discussions, the
amendment request involves no inspection that the containment will not proposed changes do not involve an increase
significant hazards consideration. degrade in a manner detectable only by Type in the probability or consequences of an
Attorney for licensee: Damon D. Obie, A testing. The last Type A test (performed accident previously evaluated.
Associate General Counsel, Talen November 19, 2010 for BFN, Unit 1, June 3, 2. Does the proposed amendment create
Energy Supply, LLC, 835 Hamilton St., 2009 for BFN, Unit 2 and May 12, 2012 for the possibility of a new or different kind of
BFN, Unit 3) shows leakage to be below accident from any accident previously
Suite 150, Allentown, PA 18101. acceptance criteria, indicating a very leak evaluated?
NRC Branch Chief: James G. Danna. tight containment. Inspections required by Response: No.
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![55416 Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices
The proposed revision to TS 5.5.12 degree of primary containment structural consideration, which is presented
changes the testing period to a permanent 15- integrity and leak-tightness that is considered below:
year interval for Type A testing (10 CFR part in the plant safety analyses is maintained.
50, Appendix J, Option B, ILRT) and a 75- The overall containment leakage rate limit 1. Does the proposed amendment involve
month interval for Type C testing (10 CFR specified by the TS is maintained, and the a significant increase in the probability or
part 50, Appendix J, Option B, LLRT). The Type A, B, and C containment leakage tests consequences of an accident previously
current test interval of 10 years, based on will continue to be performed at the evaluated?
past performance, would be extended to 15 frequencies established in accordance with Response: No.
years from the last Type A test (performed The proposed changes do not adversely
the NRC-accepted guidelines of NEI 94–01,
November 19, 2010 for BFN, Unit 1, June 3, affect accident initiators or precursors nor
Revision 3–A.
2009 for BFN, Unit 2 and May 12, 2012 for alter the design assumptions, conditions, or
Containment inspections performed in
BFN, Unit 3). The proposed extension to configuration of the facility or the manner in
accordance with other plant programs serve
Type A and Type C test intervals does not which the plant is operated and maintained.
to provide a high degree of assurance that the
create the possibility of a new or different The proposed changes do not alter or prevent
containment will not degrade in a manner
type of accident because there are no the ability of structures, systems, and
that is detectable only by an ILRT. This
physical changes being made to the plant and components (SSCs) from performing their
ensures that evidence of containment
there are no changes to the operation of the intended function to mitigate the
structural degradation is identified in a consequences of an initiating event within
plant that could introduce a new failure timely manner. Furthermore, a risk
mode creating an accident or affecting the the assumed acceptance limits. The proposed
assessment using the current BFN, Units 1, changes do not affect the source term,
mitigation of an accident. 2, and 3, PRA [probabilistic risk assessment]
Therefore, the proposed changes do not containment isolation, or radiological release
model concluded that extending the ILRT assumptions used in evaluating the
create the possibility of a new or different test interval from 10 years to 15 years results
kind of accident from any accident radiological consequences of an accident
in a very small change to the BFN, Units 1, previously evaluated. Further, the proposed
previously evaluated. 2, and 3, risk profile.
3. Does the proposed amendment involve changes do not increase the types or amounts
Accordingly, the proposed changes do not of radioactive effluent that may be released
a significant reduction in a margin of safety? involve a significant reduction in a margin of
Response: No. offsite, nor significantly increase individual
safety. or cumulative occupational/public radiation
The proposed revision to TS 5.5.12
The NRC staff has reviewed the exposures. The proposed changes do not
changes the testing period to a permanent 15-
licensee’s analysis and, based on this significantly increase the probability of an
year interval for Type A testing (10 CFR part
accident and are consistent with safety
50, Appendix J, Option B, ILRT) and a 75- review, it appears that the three analysis assumptions and resultant
month interval for Type C testing (10 CFR standards of 10 CFR 50.92(c) are consequences.
part 50, Appendix J, Option B, LLRT). The satisfied. Therefore, the NRC staff Therefore, the changes do not increase the
current test interval of 10 years, based on proposes to determine that the probability or consequences of an accident
past performance, would be extended to 15
amendment request involves no previously evaluated.
years from the last Type A test (performed 2. Does the proposed amendment create
November 19, 2010 for BFN, Unit 1, June 3, significant hazards consideration.
Attorney for licensee: General the possibility of a new or different kind of
2009 for BFN, Unit 2 and May 12, 2012 for accident from any accident previously
BFN, Unit 3). The proposed extension to Counsel, Tennessee Valley Authority,
evaluated?
Type A testing will not significantly reduce 400 West Summit Hill Dr., WT 6A, Response: No.
the margin of safety. NUREG–1493, Knoxville, TN 37902. The proposed changes do not result in a
‘‘Performance-Based Containment System NRC Branch Chief: Undine Shoop. change in the manner in which the reactor
Leakage Testing Requirements’’ trip system (RTS) and engineered safety
[‘‘Performance-Based Containment Leak-Test Tennessee Valley Authority, Docket feature actuation system (ESFAS) provide
Program’’], September 1995, generic study of Nos. 50–327 and 50–328, Sequoyah plant protection. The RTS and ESFAS will
the effects of extending containment leakage Nuclear Plant, Units 1 and 2 (SQN), continue to have the same setpoints after the
testing, found that a 20 year extension to Hamilton County, Tennessee proposed changes are implemented. There
Type A leakage testing resulted in an are no design changes associated with the
imperceptible increase in risk to the public. Tennessee Valley Authority, Docket change. The changes do not involve a
NUREG–1493 found that, generically, the Nos. 50–390 and 50–391, Watts Bar physical alteration of the plant (i.e., no new
design containment leakage rate contributes Nuclear Plant, Units 1 and 2 (WBN), or different type of equipment will be
about 0.1% to the individual risk and that the Rhea County, Tennessee installed) or a change in the methods
decrease in Type A testing frequency would governing normal plant operation. In
have a minimal effect on this risk since 95% Date of amendment request: August 7,
addition, the changes do not impose any new
of the potential leakage paths are detected by 2017. A publicly-available version is in or different requirements. The changes do not
Type C testing. Regular inspections required ADAMS under Accession No. alter assumptions made in the safety
by the ASME Code Section Xl (Subsection ML17219A505. analysis. The proposed changes are
IWE) and maintenance rule monitoring (10 Description of amendment request: consistent with the safety analysis
CFR 50.65, ‘‘Requirements for Monitoring the The amendment would revise Technical assumptions and current plant operating
Effectiveness of Maintenance at Nuclear Specification (TS) 3.2.4, ‘‘Quadrant practice.
Power Plants’’) will further reduce the risk of Therefore, the changes do not create the
a containment leakage path going undetected.
Power Tilt Ratio (QPTR),’’ and TS 3.3.1,
possibility of a new or different kind of
The proposed amendment adopts the NRC- ‘‘Reactor Trip System (RTS) accident from any accident previously
accepted guidelines of NEI 94–01, Revision Instrumentation,’’ to avoid confusion as evaluated.
3–A, for development of the BFN, Units 1, 2, to when an incore power distribution 3. Does the proposed amendment involve
and 3, performance-based leakage testing measurement for QPTR is required. The a significant reduction in a margin of safety?
program, and establishes a 15-year interval amendment would also revise the WBN
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Response: No.
for the performance of the primary TSs for consistency with the existing The proposed changes do not alter the
containment ILRT and a 75-month interval SQN TSs and Westinghouse Standard manner in which safety limits, limiting safety
for Type C testing. The amendment does not TSs in NUREG–1431, Revision 4. system settings or limiting conditions for
alter the manner in which safety limits, operation are determined. The safety analysis
limiting safety system setpoints, or limiting Basis for proposed no significant acceptance criteria are not impacted by these
conditions for operation are determined. The hazards consideration determination: changes. Redundant RTS and ESFAS trains
specific requirements and conditions of the As required by 10 CFR 50.91(a), the are maintained, and diversity with regard to
10 CFR part 50, Appendix J Testing Program licensee has provided its analysis of the the signals that provide reactor trip and
Plan, as defined in the TS, ensure that the issue of no significant hazards engineered safety features actuation is also
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![55418 Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices
Date of initial notice in Federal Limiting Condition for Operation (LCO) Action G.1. The changes are consistent
Register: April 11, 2017 (82 FR 17457). section of TS 3.3.1, ‘‘Reactor Trip with Technical Specifications Task
The Commission’s related evaluation System (RTS) Instrumentation,’’ to be Force (TSTF) Traveler TSTF–352–A,
of the amendment is contained in a reduced from ‘‘4’’ to ‘‘3’’ to allow one Revision 1, ‘‘Provide Consistent
Safety Evaluation dated October 23, nuclear instrumentation channel to be Completion Time to Reach MODE 4.’’
2017. used as an input to the reactivity Date of issuance: October 31, 2017.
No significant hazards consideration computer for physics testing without Effective date: These license
comments received: No. placing the nuclear instrumentation amendments are effective as of their
channel in a tripped condition. This date of issuance and shall be
Duke Energy Carolinas, LLC, Docket implemented within 120 days of
Nos. 50–369 and 50–370, McGuire proposed change is consistent with
Technical Specifications Task Force issuance.
Nuclear Station, Units 1 and 2, Amendment Nos.: 302 (Unit 1) and
Mecklenburg County, North Carolina (TSTF) Traveler TSTF–315–A, Revision
0, ‘‘Reduce plant trips due to spurious 281 (Unit 2). A publicly-available
Date of amendment requests: January signals to the NIS [Nuclear version is in ADAMS under Accession
11, 2017. Instrumentation System] during physics No. ML17269A198; documents related
Brief description of amendments: The testing.’’ to these amendments are listed in the
amendments modified Technical Date of issuance: October 25, 2017. Safety Evaluation enclosed with the
Specification (TS) 3.8.1, ‘‘AC Sources— Effective date: These license amendments.
Operating,’’ to allow greater flexibility amendments are effective as of their Renewed Facility Operating License
in performing Surveillance date of issuance and shall be Nos. NPF–9 and NPF–17: Amendments
Requirements (SRs) by modifying Mode implemented within 120 days of revised the Renewed Licenses and
restriction notes in TS SRs 3.8.1.8, issuance. Technical Specifications.
3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. Amendment Nos.: 301 (Unit 1) and Date of initial notice in Federal
This proposed change was consistent 280 (Unit 2). A publicly-available Register: May 23, 2017 (82 FR 23622).
with Technical Specifications Task version is in ADAMS under Accession The Commission’s related evaluation
Force (TSTF) Traveler TSTF–283–A, No. ML17261B218; documents related of the amendments is contained in a
Revision 3, ‘‘Modify Section 3.8 Mode to these amendments are listed in the Safety Evaluation dated October 31,
Restriction Notes.’’ Safety Evaluation enclosed with the 2017.
Date of issuance: October 25, 2017. amendments. No significant hazards consideration
Effective date: These license Renewed Facility Operating License comments received: Yes. One comment
amendments are effective as of its date Nos. NPF–9 and NPF–17: Amendments from a member of the public was
of issuance and shall be implemented revised the renewed facility operating received, however it was not related to
within 120 days of issuance. licenses and technical specifications. the proposed no significant hazards
Amendment Nos.: 300 (Unit 1) and Date of initial notice in Federal consideration determination or to the
279 (Unit 2). A publicly-available Register: May 23, 2017 (82 FR 23621). license amendment request.
version is in ADAMS under Accession The Commission’s related evaluation
Duke Energy Carolinas, LLC, Docket
No. ML17269A055; documents related of the amendments is contained in a
Nos. 50–369 and 50–370, McGuire
to these amendments are listed in the Safety Evaluation dated October 25,
Nuclear Station, Units 1 and 2,
Safety Evaluation enclosed with the 2017.
No significant hazards consideration Mecklenburg County, North Carolina
amendments.
Renewed Facility Operating License comments received: Yes. One comment Date of amendment requests: January
Nos. NPF–9 and NPF–17: Amendments from a member of the public was 11, 2017.
revised the renewed facility operating received, however it was not related to Brief description of amendments: The
licenses and technical specifications. the proposed no significant hazards amendments modify Technical
Date of initial notice in Federal consideration determination or to the Specification (TS) 3.7.5, ‘‘Auxiliary
Register: May 23, 2017 (82 FR 23620). license amendment request. Feedwater (AFW) System,’’ Limiting
The Commission’s related evaluation Condition for Operation (LCO)
Duke Energy Carolinas, LLC, Docket Condition A and Required Action A.1.
of the amendments is contained in a
Nos. 50–369 and 50–370, McGuire The proposed changes modify
Safety Evaluation dated October 25,
Nuclear Station, Units 1 and 2, Condition A to expand the condition to
2017.
No significant hazards consideration Mecklenburg County, North Carolina include when one turbine driven AFW
comments received: Yes. One comment Date of amendment requests: January pump is inoperable in MODE 3. This
from a member of the public was 11, 2017. expanded condition is applicable
received, however it was not related to Brief description of amendments: The immediately following a refueling
the no significant hazards consideration amendments modify the limiting outage and only if MODE 2 has not been
determination nor the license condition for operation (LCO) Required entered. Required Action A.1 is revised
amendment request. Action B.2 for Technical Specification to state ‘‘affected equipment’’ as
(TS) 3.4.10, ‘‘Pressurizer Safety Valves,’’ opposed to ‘‘steam supply’’ as a result
Duke Energy Carolinas, LLC, Docket LCO Required Action C.2 for TS 3.7.4, of the addition of the turbine driven
Nos. 50–369 and 50–370, McGuire ‘‘Steam Generator Power Operated AFW pump to Condition A. The
Nuclear Station, Units 1 and 2,
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Relief Valves (SG PORVs),’’ and LCO changes are consistent with Technical
Mecklenburg County, North Carolina Required Action G.1 for TS 3.4.12, ‘‘Low Specifications Task Force (TSTF)
Date of amendment requests: January Temperature Overpressure Protection Traveler TSTF–340–A, Revision 3,
11, 2017. (LTOP) System.’’ Specifically, the ‘‘Allow 7 day Completion Time for a
Brief description of amendments: The Completion Times are revised from 12 turbine-driven AFW pump inoperable.’’
amendments modified Technical hours to 24 hours for TS LCO 3.4.10, Date of issuance: October 31, 2017.
Specification (TS) 3.1.8, ‘‘PHYSICS Required Action B.2, and TS LCO 3.7.4, Effective date: These license
TESTS Exceptions,’’ to allow the Required Action C.2; and from 8 hours amendments are effective as of their
numbers of channels required by the to 12 hours for TS LCO 3.4.12, Required date of issuance and shall be
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55419
implemented within 120 days of revised the Renewed Licenses and Safety Evaluation dated November 6,
issuance. Technical Specifications. 2017.
Amendment Nos.: 304 (Unit 1) and Date of initial notice in Federal No significant hazards consideration
283 (Unit 2). A publicly-available Register: May 23, 2017 (82 FR 23623). comments received: No.
version is in ADAMS under Accession The Commission’s related evaluation
No. ML17277A313; documents related Energy Northwest, Docket No. 50–397,
of the amendments is contained in a
to these amendments are listed in the Columbia Generating Station, Benton
Safety Evaluation dated October 31,
Safety Evaluation enclosed with the County, Washington
2017.
amendments. No significant hazards consideration Date of amendment request:
Renewed Facility Operating License comments received: Yes. One comment November 8, 2016, as supplemented by
Nos. NPF–9 and NPF–17: Amendments from a member of the public was letter dated July 11, 2017.
revised the renewed facility operating received, however it was not related to Brief description of amendment: The
licenses and technical specifications. the proposed no significant hazards amendment would, on a one-time basis,
Date of initial notice in Federal consideration determination or the extend the completion time from 7 days
Register: May 23, 2017 (82 FR 23621). license amendment request. to 14 days for the Residual Heat
The Commission’s related evaluation Removal Train A subsystem to operable
of the amendments is contained in a Duke Energy Progress, LLC, Docket No. status associated with Technical
Safety Evaluation dated October 31, 50–400, Shearon Harris Nuclear Power Specification (TS) 3.5.1, ‘‘ECCS
2017. Plant, Unit 1, Wake and Chatham [Emergency Core Cooling System]—
No significant hazards consideration Counties, North Carolina Operating’’; TS 3.6.1.5, ‘‘Residual Heat
comments received: Yes. One comment Date of amendment request: Removal (RHR) Drywell Spray’’; and TS
from a member of the public was December 2, 2016, as supplemented by 3.6.2.3, ‘‘Residual Heat Removal (RHR)
received, however it was not related to letters dated April 25, May 22, and Suppression Pool Cooling.’’ This
the proposed no significant hazards October 2, 2017. amendment will be used to support
consideration determination or to the Brief description of amendment: The preventive maintenance, which replaces
license amendment request. amendment revised the Technical the RHR Train A subsystem’s pump and
Duke Energy Carolinas, LLC, Docket Specifications (TSs) to (1) relocate motor.
Nos. 50–369 and 50–370, McGuire cycle-specific parameters to the Core Date of issuance: October 30, 2017.
Nuclear Station, Units 1 and 2, Operating Limits Report (COLR) Effective date: As of its date of
Mecklenburg County, North Carolina consistent with Technical Specification issuance and shall be implemented
Task Force (TSTF)-339, ‘‘Relocate TS within 60 days from the date of
Date of application for amendments:
Parameters to COLR;’’ (2) delete issuance.
January 11, 2017.
Brief description of amendments: The duplicate reporting requirements in the Amendment No.: 245. A publicly-
amendments modify Technical Administrative Section of TSs available version is in ADAMS under
Specification (TS) Limiting Condition consistent with TSTF–5, ‘‘Delete Safety Accession No. ML17290A127;
for Operation (LCO) 3.9.6, ‘‘Residual Limit Violation Notification documents related to this amendment
Heat Removal (RHR) and Coolant Requirements,’’ Revision 1; and (3) are listed in the Safety Evaluation
Circulation—Low Water Level,’’ to add delete reference to plant procedure enclosed with the amendment.
a note which allows all RHR pumps to PLP–6, ‘‘Technical Specification Renewed Facility Operating License
be secured for less than or equal to 15 Equipment List Program and Core No. NPF–21: The amendment revised
minutes to support the switching of the Operating Limits Report,’’ in TSs as it the Renewed Facility Operating License
shutdown cooling loops from one train pertains to the COLR. and Technical Specifications.
to another. The changes are consistent Date of issuance: November 6, 2017. Date of initial notice in Federal
with Technical Specifications Task Effective date: As of the date of Register: February 14, 2017 (82 FR
Force (TSTF) Travelers TSTF–349–A, issuance and shall be implemented 10596). The supplemental letter dated
Revision 1, ‘‘Add Note to LCO 3.9.5 within 90 days of issuance. July 11, 2017, provided additional
Allowing Shutdown Cooling Loops Amendment No.: 161. A publicly- information that clarified the
Removal from Operation,’’ TSTF–361– available version is in ADAMS under application, did not expand the scope of
A, Revision 2, ‘‘Allow standby Accession No. ML17250A202; the application as originally noticed,
[Shutdown Cooling] SDC/RHR/[Decay documents related to this amendment and did not change the NRC staff’s
Heat Removal] DHR loop to [be] are listed in the Safety Evaluation original proposed no significant hazards
inoperable to support testing,’’ and enclosed with the amendment. consideration determination as
TSTF–438–A, Revision 0, ‘‘Clarify Renewed Facility Operating License published in the Federal Register.
Exception Notes to be Consistent with No. NPF–63: Amendment revised the The Commission’s related evaluation
the Requirement Being Excepted.’’ Facility Operating License and TSs. of the amendment is contained in a
Date of issuance: October 31, 2017. Date of initial notice in Federal Safety Evaluation dated October 30,
Effective date: As of the date of Register: February 14, 2017 (82 FR 2017.
issuance and shall be implemented 10595). The supplemental letters dated No significant hazards consideration
within 120 days of issuance. April 25, May 22, and October 2, 2017, comments received: No.
provided additional information that
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Amendment Nos.: Unit 1—303; Unit
clarified the application, did not expand Exelon Generation Company, LLC and
2—282. A publicly-available version is
the scope of the application as originally PSEG Nuclear LLC, Docket No. 50–277,
in ADAMS under Accession No.
noticed, and did not change the staff’s Peach Bottom Atomic Power Station,
ML17271A034; documents related to
original proposed no significant hazards Unit 2, York and Lancaster Counties,
these amendments are listed in the
consideration determination as Pennsylvania
Safety Evaluation enclosed with the
amendments. published in the Federal Register. Date of amendment request: May 19,
Renewed Facility Operating License The Commission’s related evaluation 2017, as supplemented by letter dated
Nos. NPF–9 and NPF–17: Amendments of the amendment is contained in a August 29, 2017.
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![Federal Register / Vol. 82, No. 223 / Tuesday, November 21, 2017 / Notices 55421
(82 FR 42849). This notice superseded Date of issuance: November 1, 2017. and to make additional corrections to
the original notice in its entirety. It also Effective date: As of the date of the form.
provided an opportunity to request a issuance and shall be implemented DATES: Comments are encouraged and
hearing by November 13, 2017, but within 30 days of issuance. will be accepted until January 22, 2018.
indicated that if the Commission makes Amendment Nos.: 215 (Unit 1) and ADDRESSES: You may send or deliver
a final NSHC determination, any such 212 (Unit 2). A publicly-available comments to Kimberly A. Holden,
hearing would take place after issuance version is in ADAMS under Accession Deputy Associate Director for Talent
of the amendments. No. ML17269A166; documents related Acquisition and Workforce Shaping,
The Commission’s related evaluation to these amendments are listed in the Employee Services, U.S. Office of
of the amendments and final NSHC are Safety Evaluation enclosed with the Personnel Management, Room 6351D,
contained in a Safety Evaluation dated amendments. 1900 E Street NW., Washington, DC
November 2, 2017. Renewed Facility Operating License 20415–9700; email at employ@opm.gov;
No significant hazards consideration Nos. NPF–2 and NPF–8: The or fax at (202) 606–2329; and to OMB
comments received: No. amendments revised the Renewed Designee, OPM Desk Officer, Office of
Florida Power & Light Company, Docket Facility Operating Licenses. Management and Budget, Office of
Nos. 50–250 and 50–251, Turkey Point Date of initial notice in Federal Information and Regulatory Affairs,
Nuclear Generating Unit Nos. 3 and 4, Register: August 29, 2017 (82 FR New Executive Office Building NW.,
Miami-Dade County, Florida 41059). Room 10235, Washington, DC 20503.
The Commission’s related evaluation FOR FURTHER INFORMATION CONTACT:
Date of amendment request:
of the amendments is contained in a Roseanna Ciarlante by telephone at
December 21, 2016.
Brief description of amendments: The Safety Evaluation dated November 1, (267) 932–8640; by fax at (202) 606–
amendments modify the Technical 2017. 4430; by TTY at (202) 418–3134; or by
Specifications by deleting high-range No significant hazards consideration email at Roseanna.Ciarlante@opm.gov.
noble gas effluent monitors’ comments received: No. SUPPLEMENTARY INFORMATION: The Office
requirements and relocating the Dated at Rockville, Maryland, this 14th day of Management and Budget is
requirements to the Turkey Point Offsite of November 2017. particularly interested in comments
Dose Calculation Manual. For the Nuclear Regulatory Commission. that:
Date of issuance: October 26, 2017. Kathryn M. Brock, 1. Evaluate whether the proposed
Effective date: As of the date of Acting Director, Division of Operating Reactor collection of information is necessary
issuance and shall be implemented Licensing, Office of Nuclear Reactor for the proper performance of the
within 90 days of issuance. Regulation. functions of the agency, including
Amendment Nos: 277 and 272. A [FR Doc. 2017–25063 Filed 11–20–17; 8:45 am] whether the information will have
publicly-available version is in ADAMS BILLING CODE 7590–01–P practical utility;
under Accession No. ML17228A563. 2. Evaluate the accuracy of the
Documents related to these amendments agency’s estimate of the burden of the
are listed in the Safety Evaluation proposed collection of information,
enclosed with the amendments. OFFICE OF PERSONNEL including the validity of the
Renewed Facility Operating License MANAGEMENT methodology and assumptions used;
Nos. DPR–31 and DPR–41: Amendments 3. Enhance the quality, utility, and
revised the Renewed Facility Operating Submission for Review: Reinstatement
clarity of the information to be
Licenses and TSs. of a Previously Approved Information
collected; and
Date of initial notice in Federal Collection With Revision, U.S. Office of 4. Minimize the burden of the
Register: March 14, 2017 (82 FR Personnel Management (OPM) collection of information on those who
13666). Standard Form (SF) 15, Application for are to respond, including through the
The Commission’s related evaluation 10-Point Veteran Preference, OMB No. use of appropriate automated,
of the amendments is contained in a 3206–0001 electronic, mechanical, or other
safety evaluation dated October 26, technological collection techniques or
2017. AGENCY: U.S. Office of Personnel
Management. other forms of information technology,
No significant hazards consideration e.g., permitting electronic submissions
comments received: No. ACTION: 60-Day notice and request for
comments. of responses.
Southern Nuclear Operating Company, The SF 15, Application for 10-Point
Docket Nos. 50–348 and 50–364, Joseph SUMMARY: This notice announces the Veteran Preference, is used by veterans
M. Farley Nuclear Plant, Units 1 and 2, Office of Personnel Management’s as both a request for preference and a
Houston County, Alabama (OPM) plan to submit to the Office of guide to determine the appropriate
Date of amendment request: August Management and Budget (OMB) a documentation to submit to support
11, 2017. request for reinstatement of a revised their claims of 10-point veterans’
Brief description of amendments: The information collection for the Standard preference when applying for Federal
amendments request an extension to the Form (SF) 15, Application for 10-Point employment. The SF 15, and the
time to achieve full compliance with 10 Veteran Preference. The SF–15 is used accompanying documentation, is used
asabaliauskas on DSKBBXCHB2PROD with NOTICES
CFR 50.48(c), National Fire Protection by agencies, OPM examining offices, by agencies, OPM examining offices,
Association (NPFA) 805, from and agency appointing officials to and agency appointing officials to
November 6, 2017, to the conclusion of adjudicate individuals’ claims for adjudicate individuals’ claims for
the FNP, Unit 1, Spring 2018 Refueling veterans’ preference in accordance with veterans’ preference in accordance with
Outage (1R28). The amendments update the Veterans’ Preference Act of 1944. the Veterans’ Preference Act of 1944.
Attachment S, ‘‘Modification and OPM’s revisions are necessary to update The proposed revisions to the SF 15 are
Implementation Items’’; of the language as a result of the enactment of necessary to update language as a result
previously approved NFPA–805 the Gold Star Fathers Act of 2015, of the enactment of the Gold Star
amendment. derived veterans’ preference for parents, Fathers Act of 2015 (Pub. L. 114–62),
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Document Created: 2017-11-21 00:43:03
Document Modified: 2017-11-21 00:43:03
| 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 December 21, 2017. A request for a hearing must be filed by January 22, 2018. | |
| Contact | Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-5411, email: [email protected] | |
| FR Citation | 82 FR 55401 |
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