81 FR 70175 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 81, Issue 196 (October 11, 2016)
Federal Register Volume 81, Issue 196 (October 11, 2016)
| Page Range | 70175-70190 | |
| FR Document | 2016-24321 |
[Federal Register Volume 81, Number 196 (Tuesday, October 11, 2016)] [Notices] [Pages 70175-70190] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2016-24321] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2016-0207] 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 September 13, 2016 to September 26, 2016. The last biweekly notice was published on September 27, 2016. DATES: Comments must be filed by November 10, 2016. A request for a hearing must be filed by December 12, 2016. 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-2016-0207. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, 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: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected]. I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0207, 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-2016-0207. 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-2016-0207, facility name, unit number(s), 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 posts all comment submissions at http://www.regulations.gov as well as entering 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 submissions 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 [[Page 70176]] margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition 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, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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/. If a petition is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (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. The petition must also set forth the specific contentions which the petitioner seeks to have litigated at 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 shall 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 those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall 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 these requirements 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, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with the NRC's regulations, policies, and procedures. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day 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). 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 decide 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 held 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 any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by December 12, 2016. 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. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c). If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in 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 [[Page 70177]] written statement of 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, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene (hereinafter ``petition''), and documents filed by interested governmental entities participating 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. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [email protected], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition (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. System requirements for accessing the E-Submittal server are available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Electronic Filing Help Desk will not be able to offer assistance in using unlisted software. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a petition. Submissions should 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 documents are 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 documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically using the NRC's adjudicatory E- Filing system may seek assistance by contacting the NRC 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 7 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, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document 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 http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a petition will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. 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. The Commission will issue a notice or order granting or denying a hearing request or intervention petition, designating the issues for any hearing that will be held and designating the Presiding Officer. A notice granting a hearing will be published in the Federal Register and served on the parties to the hearing. 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. [[Page 70178]] Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, New Hill, North Carolina Date of amendment request: May 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16151A001. Description of amendment request: The amendment would revise the Shearon Harris Nuclear Power Plant, Unit 1, technical specifications (TSs) to institute a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, the specifics of which will be contained in a licensee-controlled document. It also relocates to this program the current TS surveillance requirements (SRs) for evaluating diesel fuel oil, along with the SRs for the draining, sediment removal, and cleaning of each main fuel oil storage tank at least once every 10 years. In addition, an exception is proposed to Regulatory Guide (RG) 1.137, Revision 1, ``Fuel Oil Systems for Standby Diesel Generators,'' for the allowance of performing sampling of new fuel oil offsite prior to its addition to the fuel oil storage tanks. 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 amendment institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program. The specifics of this program will be contained in a licensee-controlled document. The current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks will be relocated to this program. The American Society for Testing and Materials (ASTM) standard references pertaining to new and stored fuel oil will be relocated to the aforementioned program; however, requirements to perform testing in accordance with applicable ASTM standards are retained in the TS. Requirements to perform surveillances of both new and stored diesel fuel oil are also retained in the TS. Evaluations of future changes to the licensee-controlled document will be conducted pursuant to the requirements of 10 CFR 50.59. A more rigorous testing of water and sediment content is added to the ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. These changes will not affect nor degrade the ability of the emergency diesel generators (DGs) to perform their specified safety functions as the diesel fuel oil continues to be properly evaluated. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems or components from performing their intended function to mitigate the consequences on an initiating event with 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 and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, of which the current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks are relocated, including pertinent ASTM standard references. A more rigorous testing of water and sediment content is added to the ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. These changes do not alter the way any structure, system, or component functions and does not modify the manner in which the plant is operated. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. The proposed amendment institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, the specifics of which will be contained in a licensee-controlled document. The current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks will be relocated to this program, along with the pertinent ASTM standard references. Changes to the licensee- controlled document are performed in accordance with the provisions of 10 CFR 50.59, thereby providing an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there is no significant reduction in a margin of safety. A more rigorous testing of water and sediment content is added to the ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. The margin of safety provided by the DGs is unaffected by the proposed changes since there continue to be TS requirements to ensure fuel oil is of the appropriate quality and reliability for emergency DG use. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies, while maintaining sufficient controls to preserve the current margins of safety. Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of ``no significant hazards consideration'' is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Business Services, 550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202. NRC Acting Branch Chief: Jeanne A. Dion. Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16242A332. Description of amendment request: The amendment would revise technical specification (TS) 5.5.6, Primary Containment Leak Rate Testing Program. These revisions would extend the Type A Primary Containment Integrated Leak Rate Test interval to 15 years and extend the Type C Local Leak Rate Test testing interval up 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 [[Page 70179]] consequences of an accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the JAF [James A. FitzPatrick Nuclear Power Plant] Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked 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 the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0087 person rem/year. EPRI [Electric Power Research Institute] Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of <= 1.0 person-rem per year, or <= 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of the risk assessment for this amendment meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The JAF Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that 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 design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers] Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the JAF Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.6 involves the extension of the JAF Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for JAF. The proposed surveillance interval extension is bounded by the 15- year ILRT Interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the 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 would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals. The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No.1, DeWitt County, Illinois Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A300. Description of amendment request: The proposed changes supports changes to the organization, staffing, and training requirements contained in [[Page 70180]] Section 5.0 of the technical specifications (TSs) after the license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor pressure vessel. 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 would not take effect until CPS has permanently ceased operation and entered a permanently defueled condition. The proposed changes would revise the CPS TS by deleting or modifying certain portions of the TS administrative controls described in Section 5.0 of the TS that are no longer applicable to a permanently shutdown and defueled facility. The proposed changes do not involve any physical changes to plant structures, systems, and components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting control settings, limiting conditions for operation, surveillance requirements, or design features. The deletion and modification of provisions of the facility administrative controls do not affect the design of SSCs necessary for safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the Spent Fuel Pool (SFP). The proposed changes are administrative in nature and do not affect any accidents applicable to the safe management of spent irradiated fuel or the permanently shutdown and defueled condition of the reactor. In a permanently defueled condition, the only credible accidents are the Fuel Handling Accident (FHA), Postulated Radioactive Releases Due to Liquid Radwaste Tank Failures, and Cask Drop Accident. Other accidents such as Loss of Coolant Accident, Loss of Feedwater, and Reactivity and Power Distribution Anomalies will no longer be applicable to a permanently defueled reactor plant. The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a permanently defueled condition will be the only operation allowed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents. The proposed changes in the administrative controls do not affect the ability to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations. The proposed changes narrow the focus of nuclear safety concerns to those associated with safely maintaining spent nuclear fuel. These changes remove the implication that CPS can return to operation once the final certification required by 10 CFR 50.82(a)(1)(ii) is submitted to the NRC. Any event involving safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the SFP would evolve slowly enough that no immediate response would be required to protect the health and safety of the public or station personnel. Adequate communications capability is provided to allow facility personnel to safely manage storage and handling of irradiated fuel. As a result, no changes to radiological release parameters are involved. There is no effect on the type or amount of radiation released, and there is no effect on predicted offsite doses in the event of an accident. Therefore, the proposed changes do not involve a significant increase in the probability or consequence of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to delete and/or modify certain TS administrative controls have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The proposed changes do not result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled and CPS will no longer be authorized to operate the reactor. The proposed changes will continue to require proper control and monitoring of safety significant parameters and activities. The proposed changes do not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers in support of maintaining the plant in a permanently shutdown and defueled condition (e.g., fuel cladding and SFP cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident. The proposed changes do not alter the protection system design or create new failure modes. The proposed changes do not involve a physical alteration of the plant, and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes involve deleting and/or modifying certain TS administrative controls once the CPS facility has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for CPS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). As a result, the occurrence of certain design basis postulated accidents are no longer considered credible when the reactor is permanently defueled. The only remaining credible accidents are the FHA, the Postulated Radioactive Releases Due to Liquid Radwaste Tank Failures, and the Cask Drop Accident. The FHA is the limiting Chapter 15 dose event for CPS in its decommissioned state. The proposed changes do not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA. The proposed changes are limited to those portions of the TS administrative controls that are not related to the safe storage and maintenance of spent irradiated fuel. These proposed changes do not directly involve any physical equipment limits or parameters. The requirements that are proposed to be revised and/or deleted from the CPS TS are not credited in the existing accident analysis for the remaining applicable postulated accidents; therefore, they do not contribute to the margin of safety associated with the accident analysis. Certain postulated DBAs [design-basis accidents] involving the reactor are no longer possible because the reactor will be permanently shut down and defueled and CPS will no longer be authorized to operate the reactor. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear,. 4300 Winfield Road, Warrenville, IL 60555. Acting NRC Branch Chief: G. Edward Miller. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida [[Page 70181]] Florida Power and 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: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16214A276. Description of amendment request: The amendments would revise the Technical Specifications (TS) consistent with Technical Specifications Task Force Traveler 545, Revision 3, ``TS Inservice Testing [IST] Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555). 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 eliminates Technical Specifications (TS) Section 5.5.6 and 5.5.7, ``Inservice Testing Program,'' for Duane Arnold and Point Beach, respectively, and eliminates TS Section 6.8.4.i, ``Inservice Testing Program'' for St. Lucie Units 1 and 2. The proposed change eliminates the requirements regarding [IST] from TS 4.0.5 in the Seabrook and Turkey Point TS. Most requirements in the [IST] Program are removed, as they are duplicative of requirements in the ASME OM [American Society of Mechanical Engineers Operation and Maintenance] Code, as clarified by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining requirements related to the IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, ``Inservice Testing Program,'' is added to the TS, which references the requirements of 10 CFR 50.55a(f). Performance of [IST] is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of [IST] performed. In most cases, the frequency of [IST] is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408- 0420. Acting NRC Branch Chief: Jeanne A. Dion. Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A030. Description of amendment request: The proposed amendment would eliminate technical specification (TS), Section 5.5.5, ``Inservice Testing [IST] Program,'' to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test Frequency.'' A new defined term, ``Inservice Testing Program,'' is added to TS Section 1.1, ``Definitions.'' The proposed change to the TS is consistent with TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarify SR [surveillance requirement] Usage Rule Application to Section 5.5 Testing.'' TS SRs that currently refer to the IST Program from Section 5.5.6 would be revised to refer to the new defined term, ``Inservice Testing Program.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS Chapter 5, ``Administrative Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating the ``Inservice Testing Program'' specification. Most requirements in the IST Program are removed as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining requirements in the Section 5.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to the regulations. A new defined term, ``Inservice Testing Program,'' is added to the TS, which references the requirements of 10 CFR 50.55a(f). Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 [[Page 70182]] are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered. 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 a margin of safety? Response: No. The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS SR 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified frequency, and will instead require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401. NRC Branch Chief: David J. Wrona. South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: August 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16225A437. Description of amendment request: The amendment request proposes changes to plant-specific Tier 2 information incorporated into the Updated Final Safety Analysis Report (UFSAR), and involves changes to combined license Appendix C (and corresponding plant-specific Tier 1 information). The proposed changes are to information identifying the frontal face area and screen surface area for the In-Containment Refueling Water Storage Tank (IRWST) screens, the location and dimensions of the protective plate located above the containment recirculation (CR) screens, and increasing the maximum Normal Residual Heat Removal System (RNS) flowrate through the IRWST and CR screens. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 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 location and dimensions of the protective plate continues to provide sufficient space surrounding the containment recirculation screens for debris to settle before reaching the screens as confirmed by an evaluation demonstrating that the protective plate continues to fulfill its design function of preventing debris from reaching the screens. In addition, the increase to the minimum IRWST screen size reinforces the ability of the screens to perform their design function with the increased RNS maximum flowrate proposed. The proposed changes do not adversely affect any accident initiating component, and thus the probabilities of the accidents previously evaluated are not affected. The affected equipment does not adversely affect the ability of equipment to contain radioactive material. Because the proposed change does not affect a release path or increase the expected dose rates, the potential radiological releases in the UFSAR accident analyses are unaffected. 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 to change the location and dimensions of the protective plate above the containment recirculation screens, to change the minimum IRWST screen size, and to increase the maximum RNS flowrate through the IRWST and CR screens does not alter the method in which safety functions are accomplished. The analyses demonstrate that the screens are able to perform accident, and no new failure modes are introduced by the proposed change. 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 change to the design does not change any of the codes or standards to which the IRWST screens, containment recirculation screens, and containment recirculation screen protective plate are designed as documented in the UFSAR. The containment recirculation screen protective plate continues to prevent debris from reaching the CR screens, and the IRWST and CR screens maintain their ability to block debris while at the proposed increase in RNS maximum flowrate. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by 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 [[Page 70183]] satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC, 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South Carolina Date of amendment request: September 8, 2016. A publicly-available version is in ADAMS under Accession No. ML16252A200. Description of amendment request: The amendment request proposes changes to the Fire Pump Head and Diesel Fuel Day Tank. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). 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 increase in head pressure by the proposed change to the fire protection system (FPS) motor-driven and diesel-driven fire pumps maintains compliance with National Fire Protection Association (NFPA) Standard NFPA-14, Standard for the Installation of Standpipe, Private Hydrants, and Hose Systems, 2000 Edition, requirements by providing adequate pressure in the standpipe and automatic sprinkler system to maintain the ability to fight and/or contain a postulated fire. The proposed change to the diesel-driven fire pump fuel day tank volume maintains the availability of the diesel-driven fire pump for service upon failure of the electric motor-driven fire pump or a loss of offsite power by providing a fuel day tank that is reserved exclusively for the diesel-driven pump and meets the minimum capacity requirements of NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, 1999 Edition. These changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and component's (SSC's) accident initiator or initiating sequence of events. These changes have no adverse impact on the support, design, or operation of mechanical and fluid systems. The response of systems to postulated accident conditions is not adversely affected by the proposed changes. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. Consequently, the plant response to previously evaluated accidents is not impacted, nor does the proposed change create any new accident precursors. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes to the fire pump performance specifications and fire pump fuel day tank volume do not affect any safety-related equipment, nor do they add any new interface to safety-related SSCs. No system or design function or equipment qualification is affected by this change. The changes do not introduce 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 proposed changes maintain compliance with the applicable Codes and Standards, thereby maintaining the margin of safety associated with these SSCs. The proposed changes do not alter any applicable design codes, code compliance, design function, or safety analysis. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change, thus the margin of safety is not reduced. Because no safety analysis or design basis acceptance limit/ criterion is challenged or exceeded by these changes, no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. 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: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16243A463. Description of amendment request: The amendment would remove the administrative controls associated with the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.5.4, ``Refueling Water Storage Tank.'' 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 change removes an administrative note added by Amendment No. 192. The administrative control applied by Amendment No. 192 was issued to prevent or reduce the risk for drainage of the Reactor Water Storage Tank (RWST) when aligned to the non-safety, non-seismic purification system. The station has implemented a modification that qualifies the interconnection of the RWST to the purification system. The installed design prevents the RWST being drained below the current Technical Specifications minimum volume requirement due to a failure in the non-safety purification system. The RWST will continue to perform its safety function and the overall system performance has not been affected [by] this proposed amendment. Assumptions previously made in evaluating the consequences of the accident are not altered, and the consequences of the accident are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation. Therefore, the proposed change has insignificant impact on the probability and consequences of an accident previously evaluated. A combination of design and administrative controls ensure that the Purification Loop maintains RWST boron concentration and water volume requirements whenever the contents of the RWST are processed through the system. The RWST is operated under System Operating Procedure for the Spent Fuel Cooling System and is protected by maintaining the isolation valve for the lower return line locked closed in modes 1 through 4. 2. Does the proposed change create the possibility of a new or different kind of [[Page 70184]] accident from any accident previously evaluated? Response: No. The proposed change does not introduce a new or different accident previously evaluated. The station implemented a qualified design that prevents the RWST from being drained below the current TS 3.5.4.a minimum volume requirement. The proposed change does not alter the design requirements of the RWST or any Structure, System or Component or its function during accident conditions. The changes do not alter assumptions made in the safety analysis and the current TS LCO are maintained. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation. The proposed change removes a note added by Amendment No. 192 that applied an administrative control to manage the risk of a postulated RWST drainage scenario by the purification 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 change involve a significant reduction in a margin of safety? Response: No. The proposed change removes a note added by Amendment No. 192. The proposed change does not alter the safety limits, limiting safety system settings or limiting conditions for operation of the RWST. The modification preserved the current licensing and design bases of the RWST, therefore the margin of safety for the RWST are not affected. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLP, 1111 Pennsylvania Avenue NW., Washington, DC 20004. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Inc. (SNC); Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant (HNP), Unit No. 2, Appling County, Georgia Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16245A257. Description of amendment request: The amendment would revise the values for the reactor core Safety Limit 2.1.1.2 for Minimum Critical Power Ratios for both single and dual recirculation loop operation. 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 brackets: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The Safety Limit Minimum Critical Power Ratio (SLMCPR) ensures that 99.9% of the fuel rods in the core will not be susceptible to boiling transition during normal operation or the most limiting postulated design-basis transient event. The new SLMCPR values preserve the existing margin to the onset of transition boiling; therefore, the probability of fuel damage is not increased as a result of this proposed change. The determination of the revised HNP Unit 2 SLMCPRs has been performed using NRC-approved methods of evaluation. These plant-specific calculations are performed each operating cycle and may require changes for future cycles. The revised SLMCPR values do not change the method of operating the plant; therefore, they have no effect on the probability of an accident, initiating event, or transient: 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 result only from a specific analysis for the HNP Unit 2 core reload design. These changes do not involve any new or different methods for operating the facility. No new initiating events or transients result from these changes. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The new SLMCPRs have been calculated using NRC-approved methods of evaluation with plant and cycle-specific input values for the fuel and core design for the upcoming cycle of operation. The SLMCPR values ensure that 99.9% of the fuel rods in the core will not be susceptible to boiling transition during normal operation or the most limiting postulated design-basis transient event. The operating MCPR limit is set appropriately above the safety limit value to ensure adequate margin when the cycle-specific transients are evaluated. Accordingly, the margin of safety is maintained with the revised values. 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35242. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: July 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16211A436. Description of amendment request: The amendment request proposes to add to License Condition 2.D.(1) of the VEGP Units 3 and 4 combined licenses an Interim Amendment Request process for changes during construction when emergent conditions are present. 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'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 amendment would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs [combined licenses] to allow construction to continue, at SNC's [Southern Nuclear Operating Company] own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue a determination that construction could continue pending SNC's initiation of the COL-ISG-025 PAR [preliminary amendment request]/LAR [license amendment request] process. The requirement to include a Nuclear Construction Safety Assessment ensures that the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. If the continued construction would result a material decrease in safety, then continued construction would not be authorized. [[Page 70185]] The proposed amendment does not modify the design, construction, or operation of any plant structures, systems, or components (SSCs), nor does it change any procedures or method of control for any SSCs. Because the proposed amendment does not change the design, construction, or operation of any SSCs, it does not adversely affect any design function as described in the Updated Final Safety Analysis Report. The proposed amendment does not affect the probability of an accident previously evaluated. Similarly, because the proposed amendment does not alter the design or operation of the nuclear plant or any plant SSCs, the proposed amendment does not represent a change to the radiological effects of an accident, and therefore, does not involve an increase in the consequences of an accident previously evaluated. 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 amendment would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow construction to continue, at SNC's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would NRC issue a determination that construction could continue pending SNC's initiation of the COL-ISG-025 PAR/LAR process. The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment only adds a new screening process and does not change the design, construction, or operation of the nuclear plant or any plant operations. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow construction to continue, at SNC's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue determination that construction could continue pending SNC's initiation of the COL-ISG-025 PAR/LAR process. The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment does not alter any design function or safety analysis. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed amendment, thus the margin of safety is not reduced. The only impact of this activity is the addition of an Interim Amendment Request process. 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. Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia Date of amendment request: September 9, 2016. A publicly-available version is in ADAMS under Accession No. ML16253A412. Description of amendment request: The amendment request proposes changes to update the Protection and Safety Monitoring System (PMS) design, specifically the description of the roles of the Qualified Data Processing System (QDPS) and the safety displays. The proposed changes add Main Control Room (MCR) safety-related display divisions A and D to plant-specific Tier 1 (and associated COL Appendix C) and the Updated Final Safety Analysis Report (UFSAR), and correct the name of the QDPS in the UFSAR by referring to the QDPS as a system, rather than a subsystem. Because, this proposed change requires a departure from Tier 1 information in Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). 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 to the roles of the qualified data processing system (QDPS) and safety-related displays, as well as the change to add Division A and Division D of the main control room (MCR) safety-related displays to the listing of PMS equipment, as identified in Combined License (COL) Appendix C (and plant-specific Tier 1) Table 2.5.2-1 and Updated Final Safety Analysis Report (UFSAR) Table 3.11-1 and 3l.6-2 do not alter any accident initiating component/system failure or event, thus the probabilities of the accidents previously evaluated are not affected. The proposed changes do not adversely affect safety-related equipment or a radioactive material barrier, and this activity dos not involve the containment of radioactive material. The radioactive material source terms and release paths used in the safety analysis are unchanged, thus the radiological releases in the UFSAR accident analysis 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 change to the roles of the QDPS and safety-related displays, as well as the change to add Division A and Division D of the MCR safety-related displays to the listing of PMS equipment, as identified in COL Appendix C (and plant-specific Tier 1) Table 2.5.2-1 and UFSAR Table 3.11-1 and 3l.6-2 does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not alter the design or capability of any sensors which provide input to the QDPS. The functionality of the QDPS to process the input obtained from sensors into data to be sent to the safety displays is not affected by the proposed changes. The proposed changes do not affect any functions performed by the safety displays, nor do the proposed changes affect the capability of the safety displays to display the data received from the QDPS. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. [[Page 70186]] 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. There is no safety-related structure, system or component (SSC) or function adversely affected by the proposed change to the roles of the QDPS and safety-related displays, nor by the change to add Division A and Division D of the MCR safety-related displays to the listing of Protection and Safety Monitoring System (PMS) equipment. The proposed changes do not alter the mechanisms by which system components are actuated or controlled. Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia Date of amendment request: September 9, 2016. A publicly-available version is in ADAMS under Accession No. ML16253A204. Description of amendment request: The amendment request proposes changes to revise plant-specific Tier 1, plant-specific Tier 2, and combined license (COL) Appendix C information concerning the details of the Class 1E direct current and uninterruptible power supply system (IDS), specifically adding seven Class 1E fuse panels to the IDS design. These proposed changes provide electrical isolation between the non-Class 1E IDS battery monitors and their respective Class 1E battery banks. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below with NRC staff edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to revise plant-specific Tier 1, COL Appendix C, and [Updated Final Safety Analysis Report (UFSAR)] information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect the design functions of the IDS, including the Class 1E battery banks and the battery monitors. These proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits as described in the current licensing basis do not have an adverse effect on any of the design functions of any plant systems. The proposed changes do not adversely affect any plant electrical system and do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect any plant electrical system and do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. There is no safety-related [structure, system, and component (SSC)] or function adversely affected by the proposed change to add IDS fuse isolation panels to non-Class 1E IDS battery monitors and Class 1E IDS circuits. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes and no margin or safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: September 13, 2016. A publicly-available version is in ADAMS under Accession No. ML16257A711. Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* information. The proposed departure consists of changes to Tier 2* information in the UFSAR to change the provided minimum reinforcement area in the column line 7.3 wall from elevation 82'-6'' to elevation 100'-0''. 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. As indicated in the UFSAR Subsection 3H.5.1.2, the wall at column line 7.3 is a shear wall that connects the shield building and the nuclear island exterior wall at column line I. Deviations were identified in [[Page 70187]] the constructed wall from the design requirements. The wall was repaired in accordance with American Concrete Institute (ACI) 349- 01. This change impacts UFSAR Table 3H.5-5. For the south face of the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel for wall section 11 for the vertical reinforcement from the wall segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12 in\2\/ft to 3.08 in\2\/ft. The change of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building for postulated loads. The criteria and requirements of ACI 349-01 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349- 01 and therefore maintains the margin of safety. This change does not involve any accident initiating components or events, thus leaving the probabilities of an accident unaltered. The reduced margin does not adversely affect any safety-related structures or equipment nor does the reduced margin reduce the effectiveness of a radioactive material barrier. Thus, the proposed change would not affect any safety-related accident mitigating function served by the containment internal structures. 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 reduction of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building. As demonstrated by the continued conformance to the applicable codes and standards governing the design of the structures, the wall withstands the same effects as previously evaluated. There is no change to the design function of the wall, and no new failure mechanisms are identified as the same types of accidents are presented to the wall before and after the change. 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 change of the provided versus required vertical reinforcing steel, identified in UFSAR Table 3H.5-5, is not a significant reduction in the margin of safety. For the south face of the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel for wall section 11 for the vertical reinforcement from the wall segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12 in\2\/ft to 3.08 in\2\/ft. The change of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building for postulated loads. The criteria and requirements of ACI 349-01 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349- 01 and therefore maintains the margin of safety. The reduction in margin does not alter any design function, design analysis, or safety analysis input or result, and sufficient margin exists to justify departure from the Tier 2 * requirements for the wall. As such, because the system continues to respond to design basis accidents in the same manner as before without any changes to the expected response of the structure, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes. Accordingly, no significant safety margin is reduced by the 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 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. Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama Date of amendment request: August 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16225A663. Description of amendment request: The amendments would modify the Technical Specifications (TSs) for Units 1, 2, and 3 by revising TS 4.3.1.2, ``Fuel Storage Criticality,'' to preclude the placement of fuel in the new fuel storage vaults. This TS change would remove the existing TS 4.3.1.2 criticality criteria wording in its entirety, and replaces it with language that specifically restricts the placement of fuel in the new fuel storage vaults. 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 amendment does not change the fuel handling processes, the fuel handling equipment, or require alteration of the plant fuel storage systems. The amendment places a restriction on use of the new fuel storage vaults, requiring that new fuel be placed only in the spent fuel pool racks. Because no changes to fuel handling equipment, fuel storage systems, or fuel handling processes are involved, the proposed amendment does not increase the probability or consequences of a fuel handling accident. Therefore, the proposed change does not increase the probability or consequences of a previously evaluated accident. 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 modification to the Technical Specifications does not require changes to the plant hardware or alter the operating characteristics of any plant system. As a result, no new failure modes are being introduced. Therefore, the change does not introduce a new or different kind of accident from those previously evaluated. 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 change to TS 4.3.1.2 ensures that the criticality margins of safety for fuel storage are maintained, by excluding the new fuel storage vault as an approved fuel storage location. The change restricts the storage of new fuel to the spent fuel pool racks, which are fully analyzed from a criticality standpoint. The change does not physically alter the fuel storage systems, or modify fuel storage requirements in such a way as to degrade the margins of criticality safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Sherry A. Quirk, General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902. NRC Acting Branch Chief: Jeanne A. Dion. Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia Date of amendment request: May 10, 2016. A publicly-available version is in ADAMS under Accession No. ML16134A069. Description of amendment request: The amendments would extend the Surry Power Station, Unit Nos. 1 and 2, Technical Specification 3.2, ``Chemical and Volume Control System,'' paragraph E requirements for primary grade water [[Page 70188]] (PG) lockout from being applicable in Refueling Shutdown and Cold Shutdown to being applicable in Refueling Shutdown, Cold Shutdown, Intermediate Shutdown, and Hot Shutdown (except during the approach to critical and within 1 hour following reactor shutdown from reactor critical or power operation). 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 conservatively imposes additional operational controls on the highest capacity flow path of PG to the Reactor Coolant System (RCS). These controls are currently credited in the boron dilution analysis in Refueling Shutdown and Cold Shutdown modes. The proposed change extends these controls into Intermediate and Hot Shutdown modes. As such, the change will provide defense against rapid reactivity insertions due to boron dilution events and reduce the probability of boron dilution events. The proposed change will have no impact on normal operating plant releases and will not increase the predicted radiological consequences of accidents postulated in the UFSAR [Updated Final Safety Analysis Report]. The proposed change makes no physical modifications and does not change plant design. Therefore, neither the probability of occurrence nor the consequences of any accident previously evaluated is significantly increased. 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 is an extension of existing operational controls on PG flow to the RCS to include additional operating modes. The change precludes high flow rate boron dilutions in Intermediate and Hot Shutdown modes similar to the current TS requirement in Refueling and Cold Shutdown modes. It does not affect the operation of the emergency boration function of the Chemical and Volume Control System (CVCS). Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change provides defense against rapid reactivity insertions to potential boron dilution events in shutdown operating modes and reduces the probability of boron dilution events. As such, it increases the margin of safety for the boron dilution event. 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219. NRC Branch Chief: Michael T. Markley. 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 request: January 18, 2016, as supplemented by letter dated June 20, 2016. Brief description of amendments: The amendments revised Technical Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,'' to allow (1) an increase in the existing Type A Integrated Leakage Rate Testing Program test interval from 10 years to 15 years, in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J,'' and the conditions and limitations specified in NEI 94-01, Revision 2-A; (2) adoption of an extension of the containment isolation valve leakage testing (Type C) frequency from the 60 months currently permitted by 10 CFR part 50, appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A; (3) adoption of the use of American National Standards Institute/American Nuclear Society (ANSI/ANS)-56.8-2002, ``Containment System Leakage Testing Requirements''; and (4) adoption of a more conservative grace interval of 9 months for Type A, Type B, and Type C leakage tests, in accordance with NEI 94-01, Revision 3-A. The amendments also made the following administrative changes: (1) Deletion of the information regarding the performance of containment visual inspections as required by Regulatory Position C.3, as the containment inspections are addressed in TS Surveillance Requirement 3.6.1.1, and (2) deletion of the information regarding the performance of the next Catawba Nuclear Station, Unit 1, Type A test no later than November 13, 2015, and the next Catawba Nuclear Station, Unit 2, Type A test no later than February 6, 2008, as both Type A tests have already occurred. Date of issuance: September 12, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 286 (Unit 1) and 282 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16229A113; 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 [[Page 70189]] revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: March 15, 2016 (81 FR 13839). The supplemental letter dated June 20, 2016, 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 amendments is contained in a Safety Evaluation dated September 12, 2016. 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 request: February 18, 2016, as supplemented by letter dated June 30, 2016. Brief description of amendment: The amendments modified Technical Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,'' for a one-time extension to the 10-year frequency of the integrated leakage rate test (ILRT) or Type A test. This revision extends the period from 10 years to 10.5 years between successive tests, changing the performance of the next ILRT from fall 2017 to spring 2019 for Unit 1 and from spring 2017 to fall 2018 for Unit 2. Date of issuance: September 26, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 290 (Unit 1) and 269 (Unit 2). A publicly available version is in ADAMS under Accession No. ML16236A053; 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 TSs. Date of initial notice in Federal Register: May 10, 2016 (81 FR 28894). The supplemental letter dated June 30, 2016, 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 amendments is contained in a Safety Evaluation dated September 26, 2016. No significant hazards consideration comments received: No. Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina Duke Energy Progress, Inc., Docket No. 50-261; H. B. Robinson Steam Electric Plant Unit No. 2 (RNP), Darlington County, South Carolina Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, (HNP), Wake and Chatham Counties, North Carolina Date of amendment request: February 1, 2016. Description of amendment request: The amendments revised the licensee's name from Duke Energy Progress, Inc. to Duke Energy Progress, LLC. Date of issuance: September 13, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 271 and 299 (BSEP); 152 (HNP); 246 (RNP). A publicly-available version is in ADAMS under Accession No. ML16217A118; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-71, DPR-62 (BSEP), NPF- 63 (HNP), and NFP-23 (RNP): Amendments revised the Renewed Facility Operating Licenses. Date of initial notice in Federal Register: April 12, 2016 (81 FR 21596). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 13, 2016. No significant hazards consideration comments received: No. Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: October 29, 2015, as supplemented by letters dated, February 16, 2016, August 8 and 26, 2016, and September 8 and 16, 2016. Brief description of amendment: The amendment revised Technical Specifications to allow the `A' Emergency Service Water (ESW) pump to be inoperable for 14 days to allow for the replacement of the `A' Train ESW pump. The amendment is applicable on a one-time basis. Date of issuance: September 16, 2016. Effective date: As of the date of issuance and shall be implemented by October 29, 2016. Amendment No.: 153. A publicly-available version is in ADAMS under Accession No. ML16253A059; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment. Renewed Facility Operating License No. NPF-63: Amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: January 5, 2016 (81 FR 260). The supplemental letters dated February 16, 2016, August 8 and 26, 2016, and September 8 and 16, 2016, 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 an SE dated September 16, 2016. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York Date of application for amendment: March 18, 2016. Brief description of amendment: The amendments revised the technical specifications (TSs) on a change to the method of calculating core reactivity for the purpose of performing the Reactivity Anomalies surveillance. Date of issuance: September 15, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: Unit 1-224 and Unit 2-158. A publicly-available version is in ADAMS under Accession No. ML16188A029; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License No. DPR-63 and NPF-69: The amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: May 10, 2016 (81 FR 28897). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 15, 2016. No significant hazards consideration comments received: No. [[Page 70190]] FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant (PNPP), Unit No. 1, Lake County, Ohio Date of amendment request: October 29, 2015, as supplemented by letter dated April 22, 2016. Brief description of amendment: The amendment revised the PNPP emergency action level (EAL) scheme to one based on the Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors.'' Date of issuance: September 14, 2016. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment No.: 173. A publicly-available version is in ADAMS under Accession No. ML16158A331; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-58: The amendment revised the Facility Operating License to authorize revision to the PNPP emergency plan. Date of initial notice in Federal Register: December 22, 2015 (80 FR 79620). The supplemental letter dated April 22, 2016, 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 September 14, 2016. 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 19, 2016, as supplemented by a letter dated May 6, 2016. Brief description of amendments: The amendments revised the Operating Licenses' licensing basis to allow elimination of the end-of- cycle moderator temperature coefficient (MTC) surveillance test as supported by NRC-Approved Topical Report CE NPSD-91 1-A and Amendment 1-A, ``Analysis of Moderator Temperature Coefficients in Support of a Change in the Technical Specification End of Cycle Negative MTC Limit,'' and St. Lucie specific supporting information. The amendments also add NRC-approved Westinghouse PARAGON Topical Report WCAP-16045-P- A, Revision 0, ``Qualification of the Two-Dimensional Transport Code PARAGON,'' to the Technical Specification list of Core Operating Limits Report methodologies. Date of issuance: September 19, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 235 and 185. A publicly-available version is in ADAMS under Accession No. ML16183A138; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: March 29, 2016 (81 FR 17506). The supplemental letter dated May 6, 2016, 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 an SE dated September 19, 2016. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 28th day of September 2016. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2016-24321 Filed 10-7-16; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices 70175
works, participants are requested not to Commission that such amendment adams.html. To begin the search, select
include copyrighted materials in their involves no significant hazards ‘‘ADAMS Public Documents’’ and then
submission, except for limited excerpts consideration, notwithstanding the select ‘‘Begin Web-based ADAMS
that serve the purpose of the pendency before the Commission of a Search.’’ For problems with ADAMS,
adjudicatory filings and would request for a hearing from any person. please contact the NRC’s Public
constitute a Fair Use application. This biweekly notice includes all Document Room (PDR) reference staff at
If a person other than the licensee notices of amendments issued, or 1–800–397–4209, 301–415–4737, or by
requests a hearing, that person shall set proposed to be issued, from September email to pdr.resource@nrc.gov. The
forth with particularity the manner in 13, 2016 to September 26, 2016. The last ADAMS accession number for each
which his interest is adversely affected biweekly notice was published on document referenced (if it is available in
by this Order and shall address the September 27, 2016. ADAMS) is provided the first time that
criteria set forth in 10 CFR 2.309(d) and DATES: Comments must be filed by it is mentioned in this document.
(f). November 10, 2016. A request for a • NRC’s PDR: You may examine and
If a hearing is requested by a person hearing must be filed by December 12, purchase copies of public documents at
whose interest is adversely affected, the 2016. the NRC’s PDR, Room O1–F21, One
Commission will issue a separate Order White Flint North, 11555 Rockville
ADDRESSES: You may submit comments
designating the time and place of any Pike, Rockville, Maryland 20852.
by any of the following methods (unless
hearings, as appropriate. If a hearing is
this document describes a different B. Submitting Comments
held, the issue to be considered at such
method for submitting comments on a Please include Docket ID NRC–2016–
hearing shall be whether this Order
specific subject): 0207, facility name, unit number(s),
should be sustained. • Federal Rulemaking Web site: Go to
In the absence of any request for application date, and subject in your
http://www.regulations.gov and search
hearing, or written approval of an comment submission.
for Docket ID NRC–2016–0207. Address The NRC cautions you not to include
extension of time in which to request a
questions about NRC dockets to Carol identifying or contact information that
hearing, the provisions specified in
Gallagher; telephone: 301–415–3463; you do not want to be publicly
Section V above shall be final 30 days
email: Carol.Gallagher@nrc.gov. For disclosed in your comment submission.
after issuance of this Order without
technical questions, contact the The NRC posts all comment
further order or proceedings. If an
individual listed in the FOR FURTHER submissions at http://
extension of time for requesting a
INFORMATION CONTACT section of this www.regulations.gov as well as entering
hearing has been approved, the
document. the comment submissions into ADAMS.
provisions specified in Section V shall • Mail comments to: Cindy Bladey,
be final when the extension expires if a The NRC does not routinely edit
Office of Administration, Mail Stop: comment submissions to remove
hearing request has not been received. OWFN–12–H08, U.S. Nuclear identifying or contact information.
Dated at Rockville, Maryland, this 3rd day of Regulatory Commission, Washington,
October 2016.
If you are requesting or aggregating
DC 20555–0001. comments from other persons for
For the Nuclear Regulatory Commission. For additional direction on obtaining submission to the NRC, then you should
Patricia K. Holahan, information and submitting comments,
Director, Office of Enforcement inform those persons not to include
see ‘‘Obtaining Information and identifying or contact information that
[FR Doc. 2016–24463 Filed 10–7–16; 8:45 am] Submitting Comments’’ in the they do not want to be publicly
BILLING CODE 7590–01–P SUPPLEMENTARY INFORMATION section of
disclosed in their comment submission.
this document. Your request should state that the NRC
FOR FURTHER INFORMATION CONTACT: Kay does not routinely edit comment
NUCLEAR REGULATORY Goldstein, Office of Nuclear Reactor
COMMISSION submissions to remove such information
Regulation, U.S. Nuclear Regulatory before making the comment
[NRC–2016–0207] Commission, Washington DC 20555– submissions available to the public or
0001; telephone: 301–415–1506, email: entering the comment submissions into
Biweekly Notice; Applications and Kay.Goldstein@nrc.gov. ADAMS.
Amendments to Facility Operating II. Notice of Consideration of Issuance
Licenses and Combined Licenses I. Obtaining Information and
Submitting Comments of Amendments to Facility Operating
Involving No Significant Hazards Licenses and Combined Licenses and
Considerations A. Obtaining Information Proposed No Significant Hazards
AGENCY: Nuclear Regulatory Please refer to Docket ID NRC–2016– Consideration Determination
Commission. 0207, facility name, unit number(s), The Commission has made a
ACTION: Biweekly notice. plant docket number, application date, proposed determination that the
and subject when contacting the NRC following amendment requests involve
SUMMARY: Pursuant to Section 189a. (2) about the availability of information for no significant hazards consideration.
of the Atomic Energy Act of 1954, as this action. You may obtain publicly- Under the Commission’s regulations in
amended (the Act), the U.S. Nuclear available information related to this § 50.92 of title 10 of the Code of Federal
Regulatory Commission (NRC) is action by any of the following methods: Regulations (10 CFR), this means that
publishing this regular biweekly notice. • Federal rulemaking Web site: Go to operation of the facility in accordance
asabaliauskas on DSK3SPTVN1PROD with NOTICES
The Act requires the Commission to http://www.regulations.gov and search with the proposed amendment would
publish notice of any amendments for Docket ID NRC–2016–0207. not (1) involve a significant increase in
issued, or proposed to be issued, and • NRC’s Agencywide Documents the probability or consequences of an
grants the Commission the authority to Access and Management System accident previously evaluated, or (2)
issue and make immediately effective (ADAMS): You may obtain publicly- create the possibility of a new or
any amendment to an operating license available documents online in the different kind of accident from any
or combined license, as applicable, ADAMS Public Documents collection at accident previously evaluated; or (3)
upon a determination by the http://www.nrc.gov/reading-rm/ involve a significant reduction in a
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![Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices 70179
consequences of an accident previously inspections performed in accordance with containment leak rate tests and Type C tests
evaluated? ASME [American Society of Mechanical for JAF. The proposed surveillance interval
Response: No. Engineers] Section Xl, the Maintenance Rule, extension is bounded by the 15-year ILRT
The proposed amendment to the TS and TS requirements serve to provide a high Interval and the 75-month Type C test
involves the extension of the JAF [James A. degree of assurance that the containment interval currently authorized within NEI 94–
FitzPatrick Nuclear Power Plant] Type A would not degrade in a manner that is 01, Revision 3–A. Industry experience
containment test interval to 15 years and the detectable only by a Type A test. Based on supports the conclusion that Type B and C
extension of the Type C test interval to 75 the above, the proposed extensions do not testing detects a large percentage of
months. The current Type A test interval of significantly increase the consequences of an containment leakage paths and that the
120 months (10 years) would be extended on accident previously evaluated. percentage of containment leakage paths that
a permanent basis to no longer than 15 years The proposed amendment also deletes are detected only by Type A testing is small.
from the last Type A test. The current Type exceptions previously granted to allow one The containment inspections performed in
C test interval of 60 months for selected time extensions of the ILRT test frequency for accordance with ASME Section Xl, TS and
components would be extended on a JAF. These exceptions were for activities that the Maintenance Rule serve to provide a high
performance basis to no longer than 75 would have already taken place by the time degree of assurance that the containment
months. Extensions of up to nine months this amendment is approved; therefore, their would not degrade in a manner that is
(total maximum interval of 84 months for deletion is solely an administrative action detectable only by Type A testing. The
Type C tests) are permissible only for non- that has no effect on any component and no combination of these factors ensures that the
routine emergent conditions. The proposed impact on how the unit is operated. margin of safety in the plant safety analysis
extension does not involve either a physical Therefore, the proposed change does not is maintained. The design, operation, testing
change to the plant or a change in the manner result in a significant increase in the methods and acceptance criteria for Type A,
in which the plant is operated or controlled. probability or consequences of an accident
B, and C containment leakage tests specified
The containment is designed to provide an previously evaluated.
in applicable codes and standards would
essentially leak tight barrier against the 2. Does the proposed change create the
continue to be met, with the acceptance of
uncontrolled release of radioactivity to the possibility of a new or different kind of
this proposed change, since these are not
environment for postulated accidents. As accident from any accident previously
evaluated? affected by changes to the Type A and Type
such, the containment and the testing C test intervals.
requirements invoked to periodically Response: No.
The proposed amendment to the TS The proposed amendment also deletes
demonstrate the integrity of the containment exceptions previously granted to allow one
exist to ensure the plant’s ability to mitigate involves the extension of the JAF Type A
containment test interval to 15 years and the time extensions of the ILRT test frequency for
the consequences of an accident, and do not JAF. These exceptions were for activities that
extension of the Type C test interval to 75
involve the prevention or identification of would have already taken place by the time
months. The containment and the testing
any precursors of an accident. The change in this amendment is approved; therefore, their
requirements to periodically demonstrate the
dose risk for changing the Type A test deletion is solely an administrative action
integrity of the containment exist to ensure
frequency from three-per-ten years to once- and does not change how the unit is operated
the plant’s ability to mitigate the
per-fifteen-years, measured as an increase to consequences of an accident do not involve and maintained. Thus, there is no reduction
the total integrated plant risk for those any accident precursors or initiators. The in any margin of safety.
accident sequences influenced by Type A proposed change does not involve a physical Therefore, the proposed change does not
testing, is 0.0087 person rem/year. EPRI change to the plant (i.e., no new or different involve a significant reduction in a margin of
[Electric Power Research Institute] Report type of equipment will be installed) or a safety.
No. 1009325, Revision 2–A states that a very change to the manner in which the plant is
small population dose is defined as an The NRC staff has reviewed the
operated or controlled.
increase of ≤ 1.0 person-rem per year, or ≤ The proposed amendment also deletes
licensee’s analysis and, based on this
1% of the total population dose, whichever exceptions previously granted to allow one review, it appears that the three
is less restrictive for the risk impact time extensions of the ILRT test frequency for standards of 10 CFR 50.92(c) are
assessment of the extended ILRT intervals. JAF. These exceptions were for activities that satisfied. Therefore, the NRC staff
The results of the risk assessment for this would have already taken place by the time proposes to determine that the
amendment meet these criteria. Moreover, this amendment is approved; therefore, their amendment request involves no
the risk impact for the ILRT extension when deletion is solely an administrative action
compared to other severe accident risks is
significant hazards consideration. Based
that does not result in any change in how the on this review, it appears that the three
negligible. Therefore, this proposed unit is operated.
extension does not involve a significant Therefore, the proposed change does not
standards of 10 CFR 50.92(c) are
increase in the probability of an accident create the possibility of a new or different satisfied. Therefore the NRC staff
previously evaluated. kind of accident from any previously proposes to determine that the
As documented in NUREG–1493, Type B evaluated. amendment request involves no
and C tests have identified a very large 3. Does the proposed change involve a significant hazards consideration.
percentage of containment leakage paths, and significant reduction in a margin of safety? Attorney for licensee: Ms. Jeanne Cho,
the percentage of containment leakage paths Response: No. Assistant General Counsel, Entergy
that are detected only by Type A testing is The proposed amendment to TS 5.5.6
very small. The JAF Type A test history
Nuclear Operations, Inc., 440 Hamilton
involves the extension of the JAF Type A
supports this conclusion. Avenue, White Plains, NY 10601.
containment test interval to 15 years and the
The integrity of the containment is subject extension of the Type C test interval to 75 NRC Branch Chief: Travis L. Tate.
to two types of failure mechanisms that can months for selected components. This Exelon Generation Company, LLC,
be categorized as: (1) Activity based, and; (2) amendment does not alter the manner in Docket No. 50–461, Clinton Power
time based. Activity based failure which safety limits, limiting safety system set
mechanisms are defined as degradation due points, or limiting conditions for operation
Station (CPS), Unit No.1, DeWitt
to system and/or component modifications or are determined. The specific requirements County, Illinois
asabaliauskas on DSK3SPTVN1PROD with NOTICES
maintenance. Local leak rate test and conditions of the TS Containment Leak Date of amendment request: July 28,
requirements and administrative controls Rate Testing Program exist to ensure that the 2016. A publicly-available version is in
such as configuration management and degree of containment structural integrity ADAMS under Accession No.
procedural requirements for system and leak-tightness that is considered in the
restoration ensure that containment integrity plant safety analysis is maintained. The ML16210A300.
is not degraded by plant modifications or overall containment leak rate limit specified Description of amendment request:
maintenance activities. The design and by TS is maintained. The proposed changes supports changes
construction requirements of the The proposed change involves only the to the organization, staffing, and
containment combined with the containment extension of the interval between Type A training requirements contained in
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Section 5.0 of the technical the focus of nuclear safety concerns to those specified in 10 CFR 50.82(a)(2), the 10 CFR
specifications (TSs) after the license no associated with safely maintaining spent 50 license for CPS will no longer authorize
longer authorizes operation of the nuclear fuel. These changes remove the operation of the reactor or emplacement or
implication that CPS can return to operation retention of fuel into the reactor vessel
reactor or placement or retention of fuel once the final certification required by 10 following submittal of the certifications
in the reactor pressure vessel. CFR 50.82(a)(1)(ii) is submitted to the NRC. required by 10 CFR 50.82(a)(1). As a result,
Basis for proposed no significant Any event involving safe storage of spent the occurrence of certain design basis
hazards consideration determination: irradiated fuel or the methods used for postulated accidents are no longer
As required by 10 CFR 50.91(a), the handling and storage of such fuel in the SFP considered credible when the reactor is
licensee has provided its analysis of the would evolve slowly enough that no permanently defueled. The only remaining
issue of no significant hazards immediate response would be required to credible accidents are the FHA, the
consideration which is presented below: protect the health and safety of the public or Postulated Radioactive Releases Due to
station personnel. Adequate communications Liquid Radwaste Tank Failures, and the Cask
1. Does the proposed amendment involve capability is provided to allow facility
a significant increase in the probability or Drop Accident. The FHA is the limiting
personnel to safely manage storage and
consequences of an accident previously Chapter 15 dose event for CPS in its
handling of irradiated fuel. As a result, no
evaluated? decommissioned state.
changes to radiological release parameters are
Response: No. The proposed changes do not adversely
involved. There is no effect on the type or
The proposed changes would not take amount of radiation released, and there is no affect the inputs or assumptions of any of the
effect until CPS has permanently ceased effect on predicted offsite doses in the event design basis analyses that impact the FHA.
operation and entered a permanently of an accident. The proposed changes are limited to those
defueled condition. The proposed changes Therefore, the proposed changes do not portions of the TS administrative controls
would revise the CPS TS by deleting or involve a significant increase in the that are not related to the safe storage and
modifying certain portions of the TS probability or consequence of an accident maintenance of spent irradiated fuel.
administrative controls described in Section previously evaluated. These proposed changes do not directly
5.0 of the TS that are no longer applicable to 2. Does the proposed change create the involve any physical equipment limits or
a permanently shutdown and defueled possibility of a new or different kind of parameters. The requirements that are
facility. accident from any accident previously proposed to be revised and/or deleted from
The proposed changes do not involve any evaluated? the CPS TS are not credited in the existing
physical changes to plant structures, systems, accident analysis for the remaining
Response: No.
and components (SSCs) or the manner in applicable postulated accidents; therefore,
The proposed changes to delete and/or
which SSCs are operated, maintained, they do not contribute to the margin of safety
modify certain TS administrative controls
modified, tested, or inspected. The proposed associated with the accident analysis. Certain
have no impact on facility SSCs affecting the
changes do not involve a change to any safety postulated DBAs [design-basis accidents]
safe storage of spent irradiated fuel, or on the
limits, limiting safety system settings, involving the reactor are no longer possible
methods of operation of such SSCs, or on the
limiting control settings, limiting conditions because the reactor will be permanently shut
handling and storage of spent irradiated fuel
for operation, surveillance requirements, or down and defueled and CPS will no longer
itself. The proposed changes do not result in
design features. be authorized to operate the reactor.
different or more adverse failure modes or
The deletion and modification of Therefore, the proposed changes do not
provisions of the facility administrative accidents than previously evaluated because
the reactor will be permanently shut down involve a significant reduction in the margin
controls do not affect the design of SSCs of safety.
necessary for safe storage of spent irradiated and defueled and CPS will no longer be
fuel or the methods used for handling and authorized to operate the reactor.
The proposed changes will continue to The NRC staff has reviewed the
storage of such fuel in the Spent Fuel Pool
(SFP). The proposed changes are require proper control and monitoring of licensee’s analysis and, based on this
administrative in nature and do not affect safety significant parameters and activities. review, it appears that the three
any accidents applicable to the safe The proposed changes do not result in any standards of 10 CFR 50.92(c) are
management of spent irradiated fuel or the new mechanisms that could initiate damage satisfied. Therefore, the NRC staff
permanently shutdown and defueled to the remaining relevant safety barriers in proposes to determine that the
condition of the reactor. support of maintaining the plant in a
amendment request involves no
In a permanently defueled condition, the permanently shutdown and defueled
condition (e.g., fuel cladding and SFP significant hazards consideration.
only credible accidents are the Fuel Handling
Accident (FHA), Postulated Radioactive cooling). Since extended operation in a Attorney for licensee: Bradley J.
Releases Due to Liquid Radwaste Tank defueled condition will be the only operation Fewell, Associate General Counsel,
Failures, and Cask Drop Accident. Other allowed, and therefore bounded by the Exelon Nuclear,. 4300 Winfield Road,
accidents such as Loss of Coolant Accident, existing analyses, such a condition does not Warrenville, IL 60555.
Loss of Feedwater, and Reactivity and Power create the possibility of a new or different
Distribution Anomalies will no longer be kind of accident. Acting NRC Branch Chief: G. Edward
applicable to a permanently defueled reactor The proposed changes do not alter the Miller.
plant. protection system design or create new NextEra Energy Duane Arnold, LLC,
The probability of occurrence of previously failure modes. The proposed changes do not
involve a physical alteration of the plant, and
Docket No. 50–331, Duane Arnold
evaluated accidents is not increased, since
no new or different kind of equipment will Energy Center, Linn County, Iowa
extended operation in a permanently
defueled condition will be the only operation be installed. Consequently, there are no new NextEra Energy Point Beach, LLC,
allowed, and therefore, bounded by the initiators that could result in a new or Docket Nos. 50–266 and 50–301, Point
existing analyses. Additionally, the different kind of accident. Beach Nuclear Plant, Units 1 and 2,
occurrence of postulated accidents associated Therefore, the proposed changes do not Town of Two Creeks, Manitowoc
with reactor operation is no longer credible create the possibility of a new or different
asabaliauskas on DSK3SPTVN1PROD with NOTICES
kind of accident from any accident
County, Wisconsin
in a permanently defueled reactor. This
significantly reduces the scope of applicable previously evaluated. NextEra Energy Seabrook, LLC, Docket
accidents. 3. Does the proposed change involve a No. 50–443, Seabrook Station, Unit No.
The proposed changes in the significant reduction in a margin of safety? 1, Rockingham County, New Hampshire
administrative controls do not affect the Response: No.
ability to successfully respond to previously The proposed changes involve deleting Florida Power & Light Company, et al.,
evaluated accidents and do not affect and/or modifying certain TS administrative Docket Nos. 50–335 and 50–389, St.
radiological assumptions used in the controls once the CPS facility has been Lucie Plant, Unit Nos. 1 and 2, St. Lucie
evaluations. The proposed changes narrow permanently shutdown and defueled. As County, Florida
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![Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices 70181
Florida Power and Light Company, availability of the affected components, as Attorney for licensee: William Blair,
Docket Nos. 50–250 and 50–251, Turkey well as their ability to mitigate the Managing Attorney—Nuclear, Florida
Point Nuclear Generating Unit Nos. 3 consequences of accidents previously Power & Light Company, P.O. Box
evaluated, is not affected.
and 4, Miami-Dade County, Florida Therefore, the proposed change does not
14000, Juno Beach, FL 33408–0420.
Date of amendment request: July 28, involve a significant increase in the
Acting NRC Branch Chief: Jeanne A.
2016. A publicly-available version is in probability or consequences of an accident Dion.
ADAMS under Accession No. previously evaluated. Northern States Power Company—
ML16214A276. 2. Does the proposed change create the
Minnesota (NSPM), Docket No. 50–263,
Description of amendment request: possibility of a new or different kind of
accident from any previously evaluated? Monticello Nuclear Generating Plant,
The amendments would revise the Response: No. Wright County, Minnesota
Technical Specifications (TS) consistent The proposed change does not alter the Date of amendment request: July 28,
with Technical Specifications Task design or configuration of the plant. The 2016. A publicly-available version is in
Force Traveler 545, Revision 3, ‘‘TS proposed change does not involve a physical ADAMS under Accession No.
Inservice Testing [IST] Program alteration of the plant; no new or different
kind of equipment will be installed. The ML16210A030.
Removal & Clarify SR [Surveillance Description of amendment request:
Requirement] Usage Rule Application to proposed change does not alter the types of
[IST] performed. In most cases, the frequency The proposed amendment would
Section 5.5 Testing’’ (ADAMS eliminate technical specification (TS),
of [IST] is unchanged. However, the
Accession No. ML15294A555). frequency of testing would not result in a Section 5.5.5, ‘‘Inservice Testing [IST]
Basis for proposed no significant new or different kind of accident from any Program,’’ to remove requirements
hazards consideration determination: previously evaluated since the testing duplicated in American Society of
As required by 10 CFR 50.91(a), the methods are not altered. Mechanical Engineers (ASME) Code for
licensee has provided its analysis of the Therefore, the proposed change does not
Operation and Maintenance of Nuclear
issue of no significant hazards create the possibility of a new or different
kind of accident from any previously Power Plants (OM Code), Case OMN–20,
consideration, which is presented
evaluated. ‘‘Inservice Test Frequency.’’ A new
below:
3. Does the proposed change involve a defined term, ‘‘Inservice Testing
1. Does the proposed change involve a significant reduction in a margin of safety? Program,’’ is added to TS Section 1.1,
significant increase in the probability or Response: No. ‘‘Definitions.’’ The proposed change to
consequences of an accident previously The proposed change eliminates some
evaluated?
the TS is consistent with TSTF–545,
requirements from the TS in lieu of Revision 3, ‘‘TS Inservice Testing
Response: No. requirements in the ASME Code, as modified
The proposed change eliminates Technical Program Removal & Clarify SR
by use of Code Case OMN–20. Compliance
Specifications (TS) Section 5.5.6 and 5.5.7, with the ASME Code is required by 10 CFR [surveillance requirement] Usage Rule
‘‘Inservice Testing Program,’’ for Duane 50.55a. The proposed change also allows Application to Section 5.5 Testing.’’ TS
Arnold and Point Beach, respectively, and inservice tests with frequencies greater than SRs that currently refer to the IST
eliminates TS Section 6.8.4.i, ‘‘Inservice 2 years to be extended by 6 months to Program from Section 5.5.6 would be
Testing Program’’ for St. Lucie Units 1 and facilitate test scheduling and consideration of revised to refer to the new defined term,
2. The proposed change eliminates the plant operating conditions that may not be
requirements regarding [IST] from TS 4.0.5 in
‘‘Inservice Testing Program.’’
suitable for performance of the required Basis for proposed no significant
the Seabrook and Turkey Point TS. Most testing. The testing frequency extension will
requirements in the [IST] Program are hazards consideration determination:
not affect the ability of the components to
removed, as they are duplicative of As required by 10 CFR 50.91(a), the
respond to an accident as the components are
requirements in the ASME OM [American required to be operable during the testing licensee has provided its analysis of the
Society of Mechanical Engineers Operation period extension. The proposed change will issue of no significant hazards
and Maintenance] Code, as clarified by Code eliminate the existing TS allowance to defer consideration, which is presented
Case OMN–20, ‘‘Inservice Test Frequency.’’ performance of missed inservice tests up to below:
The remaining requirements related to the the duration of the specified testing
IST Program are eliminated because the NRC 1. Does the proposed change involve a
frequency, and instead will require an significant increase in the probability or
has determined their inclusion in the TS is
assessment of the missed test on equipment consequences of an accident previously
contrary to regulations. A new defined term,
operability. This assessment will consider evaluated?
‘‘Inservice Testing Program,’’ is added to the
the effect on margin of safety (equipment Response: No.
TS, which references the requirements of 10
operability). Should the component be The proposed change revises TS Chapter 5,
CFR 50.55a(f).
inoperable, the TS provide actions to ensure ‘‘Administrative Controls,’’ Section 5.5,
Performance of [IST] is not an initiator to
that the margin of safety is protected. The ‘‘Programs and Manuals,’’ by eliminating the
any accident previously evaluated. As a
proposed change also eliminates a statement ‘‘Inservice Testing Program’’ specification.
result, the probability of occurrence of an
that nothing in the ASME Code should be Most requirements in the IST Program are
accident is not significantly affected by the
construed to supersede the requirements of removed as they are duplicative of
proposed change. Inservice test frequencies
under Code Case OMN–20 are equivalent to any TS. The NRC has determined that requirements in the ASME OM Code, as
the current testing period allowed by the TS statement to be incorrect. However, clarified by Code Case OMN–20, ‘‘Inservice
with the exception that testing frequencies elimination of the statement will have no Test Frequency.’’ The remaining
greater than 2 years may be extended by up effect on plant operation or safety. requirements in the Section 5.5 IST Program
to 6 months to facilitate test scheduling and Therefore, the proposed change does not are eliminated because the NRC has
consideration of plant operating conditions involve a significant reduction in a margin of determined their inclusion in the TS is
that may not be suitable for performance of safety. contrary to the regulations. A new defined
asabaliauskas on DSK3SPTVN1PROD with NOTICES
the required testing. The testing frequency The NRC staff has reviewed the term, ‘‘Inservice Testing Program,’’ is added
extension will not affect the ability of the licensee’s analysis and, based on this to the TS, which references the requirements
components to mitigate any accident of 10 CFR 50.55a(f).
review, it appears that the three
previously evaluated as the components are Performance of inservice testing is not an
required to be operable during the testing
standards of 50.92(c) are satisfied. initiator to any accident previously
period extension. Performance of inservice Therefore, the NRC staff proposes to evaluated. As a result, the probability of
tests utilizing the allowances in OMN–20 determine that the amendment request occurrence of an accident is not significantly
will not significantly affect the reliability of involves no significant hazards affected by the proposed change. Inservice
the tested components. As a result, the consideration. test frequencies under Code Case OMN–20
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![Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices 70183
satisfied. Therefore, the NRC staff These changes have no adverse impact on South Carolina Electric & Gas Company,
proposes to determine that the the support, design, or operation of South Carolina Public Service
amendment request involves no mechanical and fluid systems. The response Authority, Docket No. 50–395, Virgil C.
of systems to postulated accident conditions
significant hazards consideration. is not adversely affected by the proposed
Summer Nuclear Station, Unit No. 1,
Attorney for licensee: Ms. Kathryn M. changes. There is no change to the predicted Fairfield County, South Carolina
Sutton, Morgan, Lewis & Bockius LLC, radioactive releases due to normal operation Date of amendment request: August
1111 Pennsylvania Avenue NW., or postulated accident conditions.
Consequently, the plant response to
29, 2016. A publicly-available version is
Washington, DC, 20004–2514.
previously evaluated accidents is not in ADAMS under Accession No.
NRC Branch Chief: Jennifer Dixon-
impacted, nor does the proposed change ML16243A463.
Herrity.
create any new accident precursors. Description of amendment request:
South Carolina Electric & Gas Company Therefore, the proposed amendment does The amendment would remove the
and South Carolina Public Service not involve a significant increase in the administrative controls associated with
Authority, Docket Nos. 52–027 and 52– probability or consequences of an accident the Limiting Condition for Operation
previously evaluated.
028, Virgil C. Summer Nuclear Station, 2. Does the proposed amendment create (LCO) of Technical Specification (TS)
Units 2 and 3, Fairfield, South Carolina the possibility of a new or different kind of 3.5.4, ‘‘Refueling Water Storage Tank.’’
Date of amendment request: accident from any accident previously Basis for proposed no significant
September 8, 2016. A publicly-available evaluated? hazards consideration determination:
Response: No. As required by 10 CFR 50.91(a), the
version is in ADAMS under Accession The proposed changes do not affect the
No. ML16252A200. licensee has provided its analysis of the
operation of any systems or equipment that
Description of amendment request: may initiate a new or different kind of
issue of no significant hazards
The amendment request proposes accident, or alter any SSC such that a new consideration, which is presented
changes to the Fire Pump Head and accident initiator or initiating sequence of below, with NRC staff edits in square
Diesel Fuel Day Tank. Because, this events is created. The proposed changes to brackets:
proposed change requires a departure the fire pump performance specifications and 1. Does the proposed change involve a
fire pump fuel day tank volume do not affect significant increase in the probability or
from Tier 1 information in the
any safety-related equipment, nor do they consequences of an accident previously
Westinghouse Electric Company’s add any new interface to safety-related SSCs.
AP1000 Design Control Document evaluated?
No system or design function or equipment Response: No.
(DCD), the licensee also requested an qualification is affected by this change. The The proposed change removes an
exemption from the requirements of the changes do not introduce a new failure mode, administrative note added by Amendment
Generic DCD Tier 1 in accordance with malfunction, or sequence of events that could
No. 192. The administrative control applied
10 CFR 52.63(b)(1). affect safety or safety-related equipment.
by Amendment No. 192 was issued to
Basis for proposed no significant Therefore, the proposed amendment does
prevent or reduce the risk for drainage of the
not create the possibility of a new or different
hazards consideration determination: Reactor Water Storage Tank (RWST) when
kind of accident from any accident
As required by 10 CFR 50.91(a), the aligned to the non-safety, non-seismic
previously evaluated.
licensee has provided its analysis of the purification system. The station has
3. Does the proposed amendment involve
issue of no significant hazards implemented a modification that qualifies
a significant reduction in a margin of safety?
consideration, which is presented the interconnection of the RWST to the
Response: No.
purification system. The installed design
below: The proposed changes maintain
compliance with the applicable Codes and prevents the RWST being drained below the
1. Does the proposed amendment involve Standards, thereby maintaining the margin of current Technical Specifications minimum
a significant increase in the probability or safety associated with these SSCs. The volume requirement due to a failure in the
consequences of an accident previously proposed changes do not alter any applicable non-safety purification system. The RWST
evaluated? design codes, code compliance, design will continue to perform its safety function
Response: No. function, or safety analysis. Consequently, no and the overall system performance has not
The increase in head pressure by the safety analysis or design basis acceptance been affected [by] this proposed amendment.
proposed change to the fire protection system limit/criterion is challenged or exceeded by Assumptions previously made in evaluating
(FPS) motor-driven and diesel-driven fire the proposed change, thus the margin of the consequences of the accident are not
pumps maintains compliance with National safety is not reduced. altered, and the consequences of the accident
Fire Protection Association (NFPA) Standard Because no safety analysis or design basis are not increased. Therefore, the proposed
NFPA–14, Standard for the Installation of acceptance limit/criterion is challenged or change does not involve a significant
Standpipe, Private Hydrants, and Hose exceeded by these changes, no margin of increase in the probability or consequences
Systems, 2000 Edition, requirements by safety is reduced. of an accident previously evaluated. The
providing adequate pressure in the standpipe Therefore, the proposed amendment does Purification Loop supports the Spent Fuel
and automatic sprinkler system to maintain not involve a significant reduction in a System and is not credited for safe shutdown
the ability to fight and/or contain a margin of safety. of the plant or accident mitigation. Therefore,
postulated fire. The proposed change to the the proposed change has insignificant impact
diesel-driven fire pump fuel day tank volume The NRC staff has reviewed the on the probability and consequences of an
maintains the availability of the diesel-driven licensee’s analysis and, based on this accident previously evaluated. A
fire pump for service upon failure of the review, it appears that the three combination of design and administrative
electric motor-driven fire pump or a loss of standards of 10 CFR 50.92(c) are controls ensure that the Purification Loop
offsite power by providing a fuel day tank satisfied. Therefore, the NRC staff maintains RWST boron concentration and
that is reserved exclusively for the diesel- water volume requirements whenever the
asabaliauskas on DSK3SPTVN1PROD with NOTICES
proposes to determine that the
driven pump and meets the minimum amendment request involves no contents of the RWST are processed through
capacity requirements of NFPA 20, Standard the system. The RWST is operated under
for the Installation of Stationary Pumps for
significant hazards consideration. System Operating Procedure for the Spent
Fire Protection, 1999 Edition. These changes
Attorney for licensee: Ms. Kathryn M. Fuel Cooling System and is protected by
do not affect the operation of any systems or Sutton, Morgan, Lewis & Bockius, LLC, maintaining the isolation valve for the lower
equipment that initiate an analyzed accident 1111 Pennsylvania NW., Washington, return line locked closed in modes 1 through
or alter any structures, systems, and DC 20004–2514. 4.
component’s (SSC’s) accident initiator or NRC Branch Chief: Jennifer Dixon- 2. Does the proposed change create the
initiating sequence of events. Herrity. possibility of a new or different kind of
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![70184 Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices
accident from any accident previously in ADAMS under Accession No. proposes to determine that the
evaluated? ML16245A257. amendment request involves no
Response: No. Description of amendment request: significant hazards consideration.
The proposed change does not introduce a The amendment would revise the values Attorney for licensee: Jennifer M.
new or different accident previously
for the reactor core Safety Limit 2.1.1.2 Buettner, Associate General Counsel,
evaluated. The station implemented a
qualified design that prevents the RWST for Minimum Critical Power Ratios for Southern Nuclear Operating Company,
from being drained below the current TS both single and dual recirculation loop 40 Iverness Center Parkway,
3.5.4.a minimum volume requirement. The operation. Birmingham, AL 35242.
proposed change does not alter the design Basis for proposed no significant NRC Branch Chief: Michael T.
requirements of the RWST or any Structure, hazards consideration determination: Markley.
System or Component or its function during As required by 10 CFR 50.91(a), the
accident conditions. The changes do not alter Southern Nuclear Operating Company,
licensee has provided its analysis of the
assumptions made in the safety analysis and Inc., Docket Nos. 52–025 and 52–026,
issue of no significant hazards
the current TS LCO are maintained. The Vogtle Electric Generating Plant (VEGP),
consideration, which is presented
Purification Loop supports the Spent Fuel Units 3 and 4, Burke County, Georgia
below, with NRC staff edits in brackets:
System and is not credited for safe shutdown Date of amendment request: July 29,
of the plant or accident mitigation. The 1. Does the proposed change involve a
significant increase in the probability or 2016. A publicly-available version is in
proposed change removes a note added by
Amendment No. 192 that applied an consequences of an accident previously ADAMS under Accession No.
administrative control to manage the risk of evaluated? ML16211A436.
a postulated RWST drainage scenario by the Response: No. Description of amendment request:
purification system. The Safety Limit Minimum Critical Power The amendment request proposes to add
Therefore, the proposed change does not Ratio (SLMCPR) ensures that 99.9% of the to License Condition 2.D.(1) of the
create the possibility of a new or different fuel rods in the core will not be susceptible VEGP Units 3 and 4 combined licenses
kind of accident from any previously to boiling transition during normal operation an Interim Amendment Request process
evaluated. or the most limiting postulated design-basis for changes during construction when
3. Does the proposed change involve a transient event. The new SLMCPR values
preserve the existing margin to the onset of
emergent conditions are present.
significant reduction in a margin of safety? Basis for proposed no significant
Response: No. transition boiling; therefore, the probability
The proposed change removes a note of fuel damage is not increased as a result of hazards consideration determination:
added by Amendment No. 192. The proposed this proposed change. The determination of As required by 10 CFR 50.91(a), the
change does not alter the safety limits, the revised HNP Unit 2 SLMCPRs has been licensee has provided its analysis of the
limiting safety system settings or limiting performed using NRC-approved methods of issue of no significant hazards
conditions for operation of the RWST. The evaluation. These plant-specific calculations consideration, which is presented below
modification preserved the current licensing are performed each operating cycle and may with NRC staff’s edits in square
and design bases of the RWST, therefore the require changes for future cycles. The revised brackets:
margin of safety for the RWST are not SLMCPR values do not change the method of
affected. The proposed changes do not operating the plant; therefore, they have no 1. Does the proposed amendment involve
adversely affect systems that respond to effect on the probability of an accident, a significant increase in the probability or
safely shutdown the plant and to maintain initiating event, or transient: consequences of an accident previously
2. Does the proposed change create the evaluated?
the plant in a safe shutdown condition. The
possibility of a new or different kind of Response: No.
Purification Loop supports the Spent Fuel
accident from any accident previously The proposed amendment would add an
System and is not credited for safe shutdown
evaluated? Interim Amendment Request process to
of the plant or accident mitigation.
Response: No. Condition 2.0.(1) of the Vogtle 3 and 4 COLs
Therefore, the proposed change does not [combined licenses] to allow construction to
involve a significant reduction in a margin of The proposed changes result only from a
specific analysis for the HNP Unit 2 core continue, at SNC’s [Southern Nuclear
safety. Operating Company] own risk, in emergent
reload design. These changes do not involve
The NRC staff has reviewed the conditions, where a non-conforming
any new or different methods for operating
licensee’s analysis and, based on this condition that has little or no safety
the facility. No new initiating events or
review, it appears that the three significance is discovered and the work
transients result from these changes.
activity cannot be adjusted. The Interim
standards of 10 CFR 50.92(c) are 3. Does the proposed change involve a Amendment Request process would require
satisfied. Therefore, the NRC staff significant reduction in a margin of safety? SNC to submit a Nuclear Construction Safety
proposes to determine that the Response: No. Assessment which (1) identifies the proposed
amendment request involves no The new SLMCPRs have been calculated change; (2) evaluates whether emergent
significant hazards consideration. using NRC-approved methods of evaluation conditions are present; (3) evaluates whether
Attorney for licensee: Kathryn M. with plant and cycle-specific input values for the change would result in any material
the fuel and core design for the upcoming decrease in safety; and (4) evaluates whether
Sutton, Morgan, Lewis & Bockius LLP, cycle of operation. The SLMCPR values
1111 Pennsylvania Avenue NW., continued construction would make the non-
ensure that 99.9% of the fuel rods in the core conforming condition irreversible. Only if the
Washington, DC 20004. will not be susceptible to boiling transition continued construction would have no
NRC Branch Chief: Michael T. during normal operation or the most limiting material decrease in safety would the NRC
Markley. postulated design-basis transient event. The issue a determination that construction could
operating MCPR limit is set appropriately continue pending SNC’s initiation of the
Southern Nuclear Operating Company, above the safety limit value to ensure COL–ISG–025 PAR [preliminary amendment
Inc. (SNC); Georgia Power Company; adequate margin when the cycle-specific
asabaliauskas on DSK3SPTVN1PROD with NOTICES
request]/LAR [license amendment request]
Oglethorpe Power Corporation; transients are evaluated. Accordingly, the process. The requirement to include a
Municipal Electric Authority of Georgia; margin of safety is maintained with the Nuclear Construction Safety Assessment
City of Dalton, Georgia, Docket No. 50– revised values. ensures that the proposed amendment would
366, Edwin I. Hatch Nuclear Plant The NRC staff has reviewed the not involve a significant increase in the
(HNP), Unit No. 2, Appling County, probability or consequences of an accident
licensee’s analysis and, based on this previously evaluated. If the continued
Georgia review, it appears that the three construction would result a material decrease
Date of amendment request: August standards of 10 CFR 50.92(c) are in safety, then continued construction would
29, 2016. A publicly-available version is satisfied. Therefore, the NRC staff not be authorized.
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![70186 Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices
3. Does the proposed amendment involve 1. Does the proposed amendment involve Therefore, the proposed amendment does
a significant reduction in a margin of safety? a significant increase in the probability or not create the possibility of a new or different
Response: No. consequences of an accident previously kind of accident from any accident
There is no safety-related structure, system evaluated? previously evaluated.
or component (SSC) or function adversely Response: No. 3. Does the proposed amendment involve
affected by the proposed change to the roles The proposed changes to revise plant- a significant reduction in a margin of safety?
of the QDPS and safety-related displays, nor specific Tier 1, COL Appendix C, and Response: No.
by the change to add Division A and Division [Updated Final Safety Analysis Report There is no safety-related [structure,
D of the MCR safety-related displays to the (UFSAR)] information concerning details of system, and component (SSC)] or function
listing of Protection and Safety Monitoring the IDS, specifically the addition of seven adversely affected by the proposed change to
System (PMS) equipment. The proposed Class 1E fuse isolation panels at the add IDS fuse isolation panels to non-Class 1E
changes do not alter the mechanisms by interconnection of the non-Class 1E IDS IDS battery monitors and Class 1E IDS
which system components are actuated or battery monitors and Class 1E IDS circuits, circuits. No safety analysis or design basis
controlled. Because no safety analysis or are necessary to conform to Regulatory Guide acceptance limit/criterion is challenged or
design basis acceptance limit/criterion is 1.75 Rev. 2 (consistent with UFSAR exceeded by the proposed changes and no
challenged or exceeded by the proposed Appendix 1A exceptions) and IEEE 384–1981 margin or safety is reduced.
changes, no margin of safety is reduced. to prevent a fault on non-Class 1E circuits or Therefore, the proposed amendment does
Therefore, the proposed amendment does equipment from degrading the operation of not involve a significant reduction in a
not involve a significant reduction in a Class 1E IDS circuits and equipment below margin of safety.
margin of safety. an acceptable level. The proposed changes do
not adversely affect the design functions of The NRC staff has reviewed the
The NRC staff has reviewed the the IDS, including the Class 1E battery banks licensee’s analysis and, based on this
licensee’s analysis and, based on this and the battery monitors. review, it appears that the three
review, it appears that the three These proposed changes to revise plant- standards of 10 CFR 50.92(c) are
standards of 10 CFR 50.92(c) are specific Tier 1, COL Appendix C, and UFSAR satisfied. Therefore, the NRC staff
satisfied. Therefore, the NRC staff information concerning details of the IDS, proposes to determine that the
proposes to determine that the specifically the addition of seven Class 1E
fuse isolation panels at the interconnection of amendment request involves no
amendment request involves no significant hazards consideration.
significant hazards consideration. the non-Class 1E IDS battery monitors and
Class 1E IDS circuits as described in the Attorney for licensee: Mr. M. Stanford
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710
current licensing basis do not have an
Blanton, Balch & Bingham LLP, 1710 adverse effect on any of the design functions Sixth Avenue North, Birmingham, AL
Sixth Avenue North, Birmingham, AL of any plant systems. The proposed changes 35203–2015.
35203–2015. do not adversely affect any plant electrical NRC Branch Chief: Jennifer Dixon-
NRC Branch Chief: Jennifer Dixon- system and do not affect the support, design, Herrity.
Herrity. or operation of mechanical and fluid systems
required to mitigate the consequences of an Southern Nuclear Operating Company,
Southern Nuclear Operating Company, accident. There is no change to plant systems Inc., Docket Nos. 52–025 and 52–026,
Inc., Docket Nos. 52–025 and 52–026, or the response of systems to postulated Vogtle Electric Generating Plant, Units 3
Vogtle Electric Generating Plant Units 3 accident conditions. There is no change to and 4, Burke County, Georgia
and 4, Burke County, Georgia the predicted radioactive releases due to
postulated accident conditions. The plant Date of amendment request:
Date of amendment request:
response to previously evaluated accidents or September 13, 2016. A publicly-
September 9, 2016. A publicly-available external events is not adversely affected, nor available version is in ADAMS under
version is in ADAMS under Accession do the proposed changes create any new Accession No. ML16257A711.
No. ML16253A204. accident precursors. Description of amendment request:
Description of amendment request: Therefore, the proposed amendment does
The amendment request proposes The amendment request proposes
not involve a significant increase in the
changes to revise plant-specific Tier 1, probability or consequences of an accident changes to the Updated Final Safety
plant-specific Tier 2, and combined previously evaluated. Analysis Report (UFSAR) in the form of
license (COL) Appendix C information 2. Does the proposed amendment create departures from the incorporated plant-
concerning the details of the Class 1E the possibility of a new or different kind of specific Design Control Document Tier
direct current and uninterruptible accident from any accident previously 2* information. The proposed departure
evaluated? consists of changes to Tier 2*
power supply system (IDS), specifically Response: No.
adding seven Class 1E fuse panels to the information in the UFSAR to change the
The proposed changes to revise plant- provided minimum reinforcement area
IDS design. These proposed changes specific Tier 1, COL Appendix C, and UFSAR
provide electrical isolation between the information concerning details of the IDS,
in the column line 7.3 wall from
non-Class 1E IDS battery monitors and specifically the addition of seven Class 1E elevation 82’–6’’ to elevation 100’–0’’.
fuse isolation panels at the interconnection of Basis for proposed no significant
their respective Class 1E battery banks.
the non-Class 1E IDS battery monitors and hazards consideration determination:
Because, this proposed change requires
Class 1E IDS circuits, are necessary to As required by 10 CFR 50.91(a), the
a departure from Tier 1 information in
conform to Regulatory Guide 1.75 Rev. 2 licensee has provided its analysis of the
the Westinghouse Electric Company’s (consistent with UFSAR Appendix 1A issue of no significant hazards
AP1000 Design Control Document exceptions) and IEEE 384–1981 to prevent a consideration, which is presented
(DCD), the licensee also requested an fault on non-Class 1E circuits or equipment below:
exemption from the requirements of the from degrading the operation of Class 1E IDS
Generic DCD Tier 1 in accordance with circuits and equipment below an acceptable 1. Does the proposed amendment involve
asabaliauskas on DSK3SPTVN1PROD with NOTICES
10 CFR 52.63(b)(1). level. The proposed changes do not adversely a significant increase in the probability or
Basis for proposed no significant affect any plant electrical system and do not consequences of an accident previously
hazards consideration determination: adversely affect the design function, support, evaluated?
design, or operation of mechanical and fluid Response: No.
As required by 10 CFR 50.91(a), the systems. The proposed changes do not result As indicated in the UFSAR Subsection
licensee has provided its analysis of the in a new failure mechanism or introduce any 3H.5.1.2, the wall at column line 7.3 is a
issue of no significant hazards new accident precursors. No design function shear wall that connects the shield building
consideration, which is presented below described in the UFSAR is adversely affected and the nuclear island exterior wall at
with NRC staff edits in square brackets: by the proposed changes. column line I. Deviations were identified in
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![70188 Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices
(PG) lockout from being applicable in Therefore, the proposed change does not Duke Energy Carolinas, LLC, Docket
Refueling Shutdown and Cold involve a significant reduction in a margin of Nos. 50–413 and 50–414, Catawba
Shutdown to being applicable in safety. Nuclear Station, Units 1 and 2, York
Refueling Shutdown, Cold Shutdown, The NRC staff has reviewed the County, South Carolina
Intermediate Shutdown, and Hot licensee’s analysis and, based on this
review, it appears that the three Date of amendment request: January
Shutdown (except during the approach
standards of 10 CFR 50.92(c) are 18, 2016, as supplemented by letter
to critical and within 1 hour following
satisfied. Therefore, the NRC staff dated June 20, 2016.
reactor shutdown from reactor critical or
power operation). proposes to determine that the Brief description of amendments: The
amendment request involves no amendments revised Technical
Basis for proposed no significant Specification (TS) 5.5.2, ‘‘Containment
hazards consideration determination: significant hazards consideration.
Attorney for licensee: Lillian M. Leakage Rate Testing Program,’’ to allow
As required by 10 CFR 50.91(a), the (1) an increase in the existing Type A
Cuoco, Senior Counsel, Dominion
licensee has provided its analysis of the Integrated Leakage Rate Testing Program
Resources Services, Inc., 120 Tredegar
issue of no significant hazards test interval from 10 years to 15 years,
St., RS–2, Richmond, VA 23219.
consideration, which is presented NRC Branch Chief: Michael T. in accordance with Nuclear Energy
below: Markley. Institute (NEI) Topical Report NEI 94–
1. Does the proposed change involve a 01, Revision 3–A, ‘‘Industry Guideline
significant increase in the probability or III. Notice of Issuance of Amendments for Implementing Performance-Based
consequences of an accident previously to Facility Operating Licenses and
Option of 10 CFR part 50, appendix J,’’
evaluated? Combined Licenses
and the conditions and limitations
Response: No. During the period since publication of specified in NEI 94–01, Revision 2–A;
The proposed change conservatively the last biweekly notice, the (2) adoption of an extension of the
imposes additional operational controls on Commission has issued the following
the highest capacity flow path of PG to the containment isolation valve leakage
Reactor Coolant System (RCS). These
amendments. The Commission has testing (Type C) frequency from the 60
controls are currently credited in the boron determined for each of these months currently permitted by 10 CFR
dilution analysis in Refueling Shutdown and amendments that the application part 50, appendix J, Option B, to a 75-
Cold Shutdown modes. The proposed change complies with the standards and month frequency for Type C leakage rate
extends these controls into Intermediate and requirements of the Atomic Energy Act testing of selected components, in
Hot Shutdown modes. As such, the change of 1954, as amended (the Act), and the accordance with NEI 94–01, Revision 3–
will provide defense against rapid reactivity Commission’s rules and regulations. A; (3) adoption of the use of American
insertions due to boron dilution events and The Commission has made appropriate National Standards Institute/American
reduce the probability of boron dilution findings as required by the Act and the Nuclear Society (ANSI/ANS)-56.8–2002,
events. The proposed change will have no Commission’s rules and regulations in
impact on normal operating plant releases ‘‘Containment System Leakage Testing
10 CFR Chapter I, which are set forth in Requirements’’; and (4) adoption of a
and will not increase the predicted
radiological consequences of accidents the license amendment. more conservative grace interval of 9
postulated in the UFSAR [Updated Final A notice of consideration of issuance months for Type A, Type B, and Type
Safety Analysis Report]. The proposed of amendment to facility operating C leakage tests, in accordance with NEI
change makes no physical modifications and license or combined license, as 94–01, Revision 3–A.
does not change plant design. applicable, proposed no significant The amendments also made the
Therefore, neither the probability of hazards consideration determination, following administrative changes: (1)
occurrence nor the consequences of any and opportunity for a hearing in
accident previously evaluated is significantly Deletion of the information regarding
connection with these actions, was the performance of containment visual
increased. published in the Federal Register as
2. Does the proposed change create the inspections as required by Regulatory
possibility of a new or different kind of
indicated. Position C.3, as the containment
Unless otherwise indicated, the
accident from any accident previously inspections are addressed in TS
evaluated? Commission has determined that these
Surveillance Requirement 3.6.1.1, and
Response: No. amendments satisfy the criteria for
(2) deletion of the information regarding
The proposed change is an extension of categorical exclusion in accordance
the performance of the next Catawba
existing operational controls on PG flow to with 10 CFR 51.22. Therefore, pursuant
the RCS to include additional operating
Nuclear Station, Unit 1, Type A test no
to 10 CFR 51.22(b), no environmental
modes. The change precludes high flow rate later than November 13, 2015, and the
impact statement or environmental
boron dilutions in Intermediate and Hot next Catawba Nuclear Station, Unit 2,
assessment need be prepared for these
Shutdown modes similar to the current TS Type A test no later than February 6,
amendments. If the Commission has
requirement in Refueling and Cold Shutdown 2008, as both Type A tests have already
prepared an environmental assessment
modes. It does not affect the operation of the occurred.
emergency boration function of the Chemical under the special circumstances
provision in 10 CFR 51.22(b) and has Date of issuance: September 12, 2016.
and Volume Control System (CVCS).
Therefore, the proposed change does not made a determination based on that Effective date: As of the date of
create the possibility of a new or different assessment, it is so indicated. issuance and shall be implemented
kind of accident from any previously For further details with respect to the within 120 days of issuance.
analyzed. action see (1) the applications for Amendment Nos.: 286 (Unit 1) and
asabaliauskas on DSK3SPTVN1PROD with NOTICES
3. Does the proposed change involve a amendment, (2) the amendment, and (3) 282 (Unit 2). A publicly-available
significant reduction in a margin of safety? the Commission’s related letter, Safety version is in ADAMS under Accession
Response: No. No. ML16229A113; documents related
Evaluation and/or Environmental
The proposed change provides defense to these amendments are listed in the
against rapid reactivity insertions to potential Assessment as indicated. All of these
boron dilution events in shutdown operating items can be accessed as described in Safety Evaluation enclosed with the
modes and reduces the probability of boron the ‘‘Obtaining Information and amendments.
dilution events. As such, it increases the Submitting Comments’’ section of this Renewed Facility Operating License
margin of safety for the boron dilution event. document. Nos. NPF–35 and NPF–52: Amendments
VerDate Sep<11>2014 20:12 Oct 07, 2016 Jkt 241001 PO 00000 Frm 00104 Fmt 4703 Sfmt 4703 E:\FR\FM\11OCN1.SGM 11OCN1](https://thefederalregister.org/doc/81_FR_70371/page/14.svg)
![70190 Federal Register / Vol. 81, No. 196 / Tuesday, October 11, 2016 / Notices
FirstEnergy Nuclear Operating approved Westinghouse PARAGON Technical Specification 3⁄4.6.5, ‘‘Ice
Company, Docket No. 50–440, Perry Topical Report WCAP–16045–P–A, Condenser.’’
Nuclear Power Plant (PNPP), Unit No. 1, Revision 0, ‘‘Qualification of the Two- DATES: October 11, 2016.
Lake County, Ohio Dimensional Transport Code ADDRESSES: Please refer to Docket ID
Date of amendment request: October PARAGON,’’ to the Technical NRC–2014–0045 when contacting the
29, 2015, as supplemented by letter Specification list of Core Operating NRC about the availability of
dated April 22, 2016. Limits Report methodologies. information regarding this document.
Date of issuance: September 19, 2016.
Brief description of amendment: The You may obtain publicly-available
Effective date: As of the date of
amendment revised the PNPP information related to this document
issuance and shall be implemented
emergency action level (EAL) scheme to using any of the following methods:
within 90 days of issuance.
one based on the Nuclear Energy • Federal Rulemaking Web site: Go to
Amendment Nos.: 235 and 185. A
Institute (NEI) guidance in NEI 99–01, http://www.regulations.gov and search
publicly-available version is in ADAMS
Revision 6, ‘‘Development of Emergency for Docket ID NRC–2014–0045. Address
under Accession No. ML16183A138;
Action Levels for Non-Passive questions about NRC dockets to Carol
documents related to these amendments
Reactors.’’ Gallagher; telephone: 301–415–3463;
are listed in the Safety Evaluation (SE)
Date of issuance: September 14, 2016. email: Carol.Gallagher@nrc.gov. For
enclosed with the amendments.
Effective date: As of the date of Renewed Facility Operating License technical questions, contact the
issuance and shall be implemented individual listed in the FOR FURTHER
Nos. DPR–67 and NPF–16: Amendments
within 180 days of issuance. INFORMATION CONTACT section of this
revised the Renewed Facility Operating
Amendment No.: 173. A publicly- document.
Licenses and Technical Specifications. • NRC’s Agencywide Documents
available version is in ADAMS under Date of initial notice in Federal
Accession No. ML16158A331; Access and Management System
Register: March 29, 2016 (81 FR
documents related to this amendment (ADAMS): You may obtain publicly-
17506). The supplemental letter dated
are listed in the Safety Evaluation available documents online in the
May 6, 2016, provided additional
enclosed with the amendment. ADAMS Public Documents collection at
information that clarified the
Facility Operating License No. NPF– http://www.nrc.gov/reading-rm/
application, did not expand the scope of
58: The amendment revised the Facility adams.html. To begin the search, select
the application as originally noticed,
Operating License to authorize revision ‘‘ADAMS Public Documents’’ and then
and did not change the staff’s original
to the PNPP emergency plan. select ‘‘Begin Web-based ADAMS
proposed no significant hazards
Date of initial notice in Federal Search.’’ For problems with ADAMS,
consideration determination as
Register: December 22, 2015 (80 FR please contact the NRC’s Public
published in the Federal Register.
79620). The supplemental letter dated The Commission’s related evaluation Document Room (PDR) reference staff at
April 22, 2016, provided additional of the amendment is contained in an SE 1–800–397–4209, 301–415–4737, or by
information that clarified the dated September 19, 2016. email to pdr.resource@nrc.gov. The
application, did not expand the scope of No significant hazards consideration ADAMS accession number for each
the application as originally noticed, comments received: No. document referenced (if that document
and did not change the staff’s original is available in ADAMS) is provided the
Dated at Rockville, Maryland, this 28th day first time that a document is referenced.
proposed no significant hazards of September 2016.
consideration determination as • NRC’s PDR: You may examine and
For the Nuclear Regulatory Commission. purchase copies of public documents at
published in the Federal Register.
Anne T. Boland, the NRC’s PDR, Room O1–F21, One
The Commission’s related evaluation
of the amendment is contained in a Director, Division of Operating Reactor White Flint North, 11555 Rockville
Licensing, Office of Nuclear Reactor Pike, Rockville, Maryland 20852.
Safety Evaluation dated September 14, Regulation.
2016. FOR FURTHER INFORMATION CONTACT:
[FR Doc. 2016–24321 Filed 10–7–16; 8:45 am] Andrew Hon, Office of Nuclear Reactor
No significant hazards consideration
comments received: No. BILLING CODE 7590–01–P Regulation, U.S. Nuclear Regulatory
Commission, Washington DC 20555–
Florida Power & Light Company, et al., 0001; telephone: 301–415–8480, email:
Docket Nos. 50–335 and 50–389, St. NUCLEAR REGULATORY Andrew.Hon@nrc.gov.
Lucie Plant Unit Nos. 1 and 2, St. Lucie COMMISSION
SUPPLEMENTARY INFORMATION: The NRC
County, Florida [Docket Nos. 50–327 and 50–328; NRC– has granted the request of Tennessee
Date of amendment request: January 2014–0045] Valley Authority (the licensee) to
19, 2016, as supplemented by a letter withdraw its July 3, 2013, application
dated May 6, 2016. Tennessee Valley Authority, Sequoyah (ADAMS Accession No. ML13199A281)
Brief description of amendments: The Nuclear Plant, Units 1 and 2 for proposed amendment to Facility
amendments revised the Operating AGENCY: Nuclear Regulatory Operating License Nos. DPR–77 and
Licenses’ licensing basis to allow Commission. DPR–79 issued to the licensee for
elimination of the end-of-cycle ACTION: License amendment application; operation of the Sequoyah Nuclear
moderator temperature coefficient withdrawal by applicant. Plant, Units 1 and 2, located in
(MTC) surveillance test as supported by Hamilton County, Tennessee.
asabaliauskas on DSK3SPTVN1PROD with NOTICES
NRC-Approved Topical Report CE SUMMARY: The U.S. Nuclear Regulatory The licensee requested to revise
NPSD–91 1–A and Amendment 1–A, Commission (NRC) has granted the Sequoyah Nuclear Plant, Units 1 and 2
‘‘Analysis of Moderator Temperature request of Tennessee Valley Authority Technical Specification 3⁄4.6.5, ‘‘Ice
Coefficients in Support of a Change in (the licensee) to withdraw its Condenser,’’ to increase the total ice
the Technical Specification End of application dated July 3, 2013, for a weight from 2,225,880 pounds to
Cycle Negative MTC Limit,’’ and St. proposed amendment to DPR–77 and 2,540,808 pounds. This proposed
Lucie specific supporting information. DPR–79. The proposed amendment amendment request was noticed in the
The amendments also add NRC- would have revised Units 1 and 2 Federal Register (79 FR 12246) dated
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Document Created: 2018-02-13 16:33:30
Document Modified: 2018-02-13 16:33:30
| 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 November 10, 2016. A request for a hearing must be filed by December 12, 2016. | |
| Contact | Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected] | |
| FR Citation | 81 FR 70175 |
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