| Page Range | 50729-50742 | |
| FR Document | 2016-18290 |
[Federal Register Volume 81, Number 148 (Tuesday, August 2, 2016)] [Notices] [Pages 50729-50742] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2016-18290] [[Page 50729]] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2016-0151] 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 July 5, 2019, to July 19, 2016. The last biweekly notice was published on July 19, 2016 (81 FR 46958). DATES: Comments must be filed by September 1, 2016. A request for a hearing must be filed by October 3, 2016. ADDRESSES: You may submit comments by any of the following methods:Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0151. 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: Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected]. I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0151, 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-2016-0151. 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-2016-0151, facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. I. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this [[Page 50730]] action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested person(s) 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 request for a hearing or petition for leave to intervene 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 request and/or 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 for leave to intervene 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 requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/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 requestor/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 requestor/ petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion to support its position on this 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 amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/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)-(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 September 19, 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 for leave to intervene 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 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, 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 [[Page 50731]] 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 request or petition for hearing (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 detailed in the NRC's ``Guidance for Electronic Submission to the NRC,'' which is available on the agency's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.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. If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance 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 request/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 hearing request and petition to intervene 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. 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 obtaining information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station (CNS), Units 1 and 2, York County, South Carolina Date of amendment request: May 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16147A105. Description of amendment request: The amendments would revise Sections 8.3.1, ``AC Power Systems''; 9.2.1, ``Nuclear Service Water System''; 9.4.1, ``Control Room Area Ventilation''; and 9.4.3, ``Auxiliary Building Ventilation System,'' of the updated final safety analysis report (UFSAR), to clarify how a shutdown unit supplying either its normal or emergency power source may be credited for operability of shared components supporting the operating unit. 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? [[Page 50732]] Response: No. The proposed change only involves a change to the UFSAR to reflect how shared systems at CNS can be powered from offsite or onsite power sources. The proposed change does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change does not impact the consequence of any analyzed accident since the change does not adversely affect any equipment related to accident mitigation. 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 only involves a change to the UFSAR to reflect how shared systems at CNS can be powered from offsite or onsite power sources. The proposed change does not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system set points are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed change and the proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change only involves a change to the UFSAR to reflect how shared systems at CNS can be powered from offsite or onsite power sources. The proposed change to the UFSAR does not affect any of the assumptions used in the CNS accident analysis, nor does it affect any operability requirements for equipment important to safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 28202-1802. NRC Branch Chief: Michael T. Markley. FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, Pennsylvania FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: May 24, 2016. A publicly-available version is in ADAMS under Accession No. ML16148A047. Description of amendment request: The amendment would eliminate Technical Specification (TS), Section 5.5, ``Inservice Testing Program,'' to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test Frequency.'' A new defined term, ``INSERVICE TESTING PROGRAM,'' will be 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 Usage Rule Application to Section 5.5 Testing.'' Using the consolidated line-item improvement process, the NRC staff issued a notice of availability in the Federal Register on March 28, 2016 (81 FR 17208), for a possible proposed change that modifies the Standard Technical Specification (STS) to eliminate Chapter 5.0, ``Administrative Controls,'' specification Section 5.5, ``Inservice Testing Program,'' to remove requirements duplicated in ASME Code, Case OMN-20, ``Inservice Test Frequency.'' ASME Code, Case OMN-20, provides similar definitions and allowances as in the current STS Inservice Testing Program. The notice of availability added a new defined term, ``Inservice Testing Program (IST),'' to the STS, Section 1.1, ``Definitions.'' Also, the STS, Section 3.0, ``Surveillance Requirement (SR) Applicability,'' and STS Bases were revised to explain the application of the usage rules to the Section 5.5 testing requirements. Existing uses of the term ``Inservice Testing Program'' in the STS and STS Bases were capitalized to indicate that it is now a defined term. The FR notice included the model application, No Significant Hazards Consideration (NSHC) Determination, and the model safety evaluation for referencing in license amendment applications. The licensee affirmed the applicability of the model NSHC determination in its application dated May 24, 2016, which is presented below. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, along with NRC edits in square brackets: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS Chapter 5, ``Administrative Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating the ``Inservice Testing Program'' specification. Most requirements in the Inservice Testing 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 Inservice Testing 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 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 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 [[Page 50733]] 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 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 SR 3.0.3 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 a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications 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: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Mail Stop A-GO-15, Akron, OH 44308. NRC Acting Branch Chief: G. Edward Miller. Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida Date of amendment request: June 21, 2016. A publicly-available version is in ADAMS under Accession No. ML16190A118. Description of amendment request: The amendment would update the Technical Specifications to revise the emergency diesel generator (EDG) engine-mounted fuel tank minimum volume from 200 gallons of fuel each to 238 gallons of fuel each. 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 EDGs engine-mounted fuel oil tanks are part of a system used to mitigate the consequences of an accident and do not increase the probability of an accident previously evaluated. The increase in minimum fuel oil requirements enables operation of the EDGs to remain unchanged for ULSD [ultra low sulfur diesel] fuel oil, thus the EDGs continue to be capable of performing their design functions. Acceptance criteria continue to be satisfied. Accordingly, the proposed change does not increase the consequences of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the increase in minimum EDGs engine-mounted fuel oil tank volume. The proposed change has no adverse effect on any safety-related system and does not change the performance or integrity of any safety-related equipment. No new safety-related equipment is being added or replaced as a result of the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The calculation for EDG fuel consumption shows that with the minimum day tank volume of 238 gallons of ULSD fuel, the requirement for two day tanks to provide a usable volume which is sufficient for at least 1 hour 100% load operation of one diesel generator set, plus a minimum margin of 10% is met. The day tank minimum volumes with the DOST [diesel oil storage tank] minimum volume is sufficient for the EDG loading increase due to potential operation at the upper frequency limit of 60.6 HZ [Hertz] (60 HZ, +1%) and the EPU [extended power uprate] requirements. The EDG fuel consumption analyses demonstrate that the EDG design continues to satisfy its safety function. The design basis limits for the accident and transient analyses will continue to meet their design criteria. 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 S. Blair, Managing Attorney-- Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/ JB, Juno Beach, FL 33408-0420. NRC Acting Branch Chief: Tracy J. Orf. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California Date of amendment request: May 12, 2016. A publicly-available version is in ADAMS under Package Accession No. ML16146A100. Description of amendment request: The amendments would revise Technical Specification (TS) 5.5.6, ``Containment Leakage Rate Testing Program,'' to allow the following: Increase in the existing 10 CFR part 50, Appendix J, ``Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,'' Type A test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 2-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' October 2008 (ADAMS Accession No. ML100620847). Adopt the use of American National Standards Institute/ American Nuclear Society (ANSI/ANS) 56.8-2002, ``Containment System Leakage Testing Requirements,'' as referenced in NEI 94-01, Revision 2- A. Adopt an allowable test interval extension of 9 months, which is shorter than the currently allowed 25 percent grace, for the 10 CFR 50, Appendix J, Type A, Type B, and Type C leakage tests in accordance with NEI 94-01, Revision 2-A. The proposed changes would revise TS 5.5.16 to replace the reference to NRC Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test Program,'' September 1995 (ADAMS Accession No. ML003740058), and 10 CFR 50, Appendix J, Option B, [[Page 50734]] ``Performance-Based Requirements,'' with a reference to NEI 94-01, Revision 2-A. In addition, the proposed amendments would modify TS 5.5.16 to remove an exception under paragraph 5.16.a.3 for a one-time 15-year Type A test interval beginning May 4, 1994, for Unit 1 and April 30, 1993, for Unit 2. 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 license amendment adopts the Nuclear Regulatory Commission (NRC)-accepted guidelines of Nuclear Energy Institute (NEI) Report 94-01, Revision 2-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' for development of the Diablo Canyon Power Plant (DCPP) Units 1 and 2 performance-based Technical Specification 5.5.16, ``Containment Leakage Rate Testing Program.'' NEI 94-01 allows, based on risk and performance, an extension of Type A containment leak test intervals. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses. The findings of the DCPP risk assessment confirm the general findings of previous studies that the risk impact with extending the containment leak rate is small, per the guidance provided in Regulatory Guide (RG) 1.174, Revision 2 ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant- Specific Changes to the Licensing Basis,'' May 2011 (ADAMS Accession No. ML100910006). Since the license amendment is implementing a performance-based containment testing program, the proposed license amendment 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 requirements for leakage rate tests and acceptance criteria will not be changed by this license amendment. Therefore, the containment will continue to perform its design function as a barrier to fission product releases. The proposed license amendment also deletes an exception previously granted to allow one time extensions of the Type A test frequency for DCPP. This exception was for an activity that has already taken place; therefore, the deletion is solely an administrative action that has no effect on any component and no physical impact on how the units are operated. 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 accident from any accident previously evaluated? Response: No. The proposed license amendment to implement a performance-based Type A testing program does not change the design or operation of structures, systems, or components of the plant. In addition, the proposed changes would not impact any other plant system or component. The proposed license amendment would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by the proposed changes. The proposed license amendment also deletes an exception previously granted to allow one time extensions of the Type A test frequency for DCPP. This exception was for an activity that has already taken place; therefore, the deletion is solely an administrative action and does not change how the units are operated or maintained. Therefore, the proposed license amendment does not create the possibility of a new or different accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed license amendment to implement the performance- based Type A testing program does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation. The specific requirements and conditions of Technical Specification 5.5.16, ``Containment Leakage 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 the Technical Specifications is maintained. This ensures that the margin of safety in the plant safety analysis is maintained. The proposed amendment will ensure that the design, operation, testing methods and acceptance criteria for Type A tests specified in applicable codes and standards would continue to be met since these are not affected by implementation of a performance based Type A testing interval. The proposed amendment also deletes an exception previously granted to allow one time extensions of the Type A test frequency for DCPP. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how the unit is operated and maintained. 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: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, CA 94120. NRC Branch Chief: Robert J. Pascarelli. South Carolina Electric and 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: June 16, 2016, as supplemented by letter dated July 7, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16168A282 and ML16189A453, respectively. Description of amendment request: The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* and associated Tier 2 information. Specifically, the proposed departures consist of changes to the UFSAR to revise the details of the structural design of auxiliary building floors within module CA20 at approximate design elevations of 82'-6'' and 92'-6''. 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 design functions of the auxiliary building floors are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the auxiliary building. The auxiliary building is a seismic Category I structure and is designed for dead, live, thermal, pressure, safe shutdown earthquake loads, and loads due to postulated pipe breaks. The proposed changes to UFSAR descriptions are intended to address changes in the detail design of floors in the auxiliary building. The thickness and strength of the auxiliary building floors are not reduced. As a result, the design function of the auxiliary building structure is not adversely affected by the proposed changes. 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 [[Page 50735]] adversely affected, nor do the changes described 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 changes to UFSAR descriptions are proposed to address changes in the detail design of floors in the auxiliary building. The thickness, geometry, and strength of the structures are not adversely altered. The concrete and reinforcement materials are not altered. The properties of the concrete are not altered. The changes to the design details of the auxiliary building structure do not create any new accident precursors. As a result, the design function of the auxiliary building structure is not 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. The criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349 and AISC N690 and therefore maintains the margin of safety. Analysis of the connection design confirms that code provisions are appropriate to the floor to wall connection. The proposed changes to the UFSAR address changes in the detail design of floors in the auxiliary building. The proposed changes also incorporate the requirements for development and anchoring of headed reinforcement which were previously approved. There is no change to design requirements of the auxiliary building structure. There is no change to the method of evaluation from that used in the design basis calculations. There is not a significant change to the in structure response spectra. Therefore, the proposed amendment does not result in 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 Acting Branch Chief: Jennifer Dixon-Herrity. South Carolina Electric and Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina Date of amendment request: July 5, 2016. A publicly-available version is in ADAMS under Accession No. ML16187A392. Description of amendment request: The amendment request relates to changes to the slab thickness between Column Lines I to J-1 and 2 to 4 at plant elevation 153'-0''. The changes involve changes to incorporated AP1000 Design Control Document Tier 1 information and corresponding departures to Tier 2* Updated Final Safety Analysis Report information and conforming changes to the Combined License, Appendix C. 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 design functions of the nuclear island structures are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the nuclear island. The nuclear island structures are structurally designed to meet seismic Category I requirements as defined in Regulatory Guide 1.29. The change of the thickness of the floor above the [Component Cooling Water System (CCS)] Valve room in the auxiliary building meets criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690, does not have an adverse impact on the response of the nuclear island structures to safe shutdown earthquake ground motions or loads due to anticipated transients or postulated accident conditions. The proposed changes do not impact the support, design, or operation of mechanical and fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor does the change described 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 change is to revise the thickness of the floor above the CCS Valve room in the auxiliary building. The proposed changes do not change the design requirements of the nuclear island structures. The proposed changes do not change the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism for the nuclear island structures or new accident precursors. As a result, the design function of the nuclear island structures is not adversely affected by the proposed change. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus, no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety previously evaluated. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. Acting NRC Branch Chief: Jennifer Dixon-Herrity. Southern California Edison Company, et al., Docket Nos. 50-361 and 50- 362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California Date of amendment request: June 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16172A075. Description of amendment request: The amendments would extend the scheduled implementation date for Milestone 8 of the San Onofre Nuclear Generating Station, Units 2 and 3, Cyber Security Plan to December 31, 2019, in order to more fully reflect the permanent shutdown status of the facility and accommodate ongoing decommissioning activities. 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 [[Page 50736]] consequences of an accident previously evaluated? Response: No. The proposed change to the San Onofre Nuclear Generating Station (SONGS) Cyber Security Plan Implementation Schedule is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components (SSCs) relied upon to mitigate the consequences of postulated accidents, and has no impact on the probability or consequences of an accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the SONGS Cyber Security Plan Implementation Schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the SSCs relied upon to mitigate the consequences of postulated accidents, and does not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the SONGS Cyber Security Plan Implementation Schedule is administrative in nature. Since the proposed change is administrative in nature, there is no change to these established safety margins. 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: Walker A. Matthews, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, CA 91770. NRC Branch Chief: Bruce Watson. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: March 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16064A352. Description of amendment request: The amendment proposes to change the VEGP, Units 3 and 4, License Conditions 2.D(12)(d) and submits the new plant-specific Emergency Action Level (EAL) scheme for both units. 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 requested amendment proposes changes to the Vogtle Electric Generating Plant (VEGP) Units 3 and 4 License Conditions 2.D(12)(d) and submits the new plant-specific Emergency Action Level (EAL) scheme for both units. The proposed changes, including the modification of VEGP Units 3 and 4 License Condition 2.D(12)(d) and submittal of the new plant-specific EALs for both units, do not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes. Therefore, the requested 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, including the modification of VEGP Units 3 and 4 License Conditions 2.D(12)(d) and submittal of the new plant- specific EALs for both units, do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes will not introduce failure modes that could result in a new accident, and the changes do not alter assumptions made in the safety analysis. The proposed changes are not initiators of any accidents. Therefore, the requested amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes to the plant- specific EALs and the modification of VEGP Units 3 and 4 License Conditions 2.D(12)(d) do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these proposed changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. 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 Acting Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: April 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16117A531. Description of amendment request: The amendments would change the certified AP1000 Design Control Document (DCD) Tier 1 information and depart from the plant-specific Tier 2 and Tier 2* information in the Updated Final Safety Analysis Report (UFSAR) for VEGP, Units 3 and 4, by modifying the overall design of the Central Chilled Water subsystem to relocate the Air Cooled Chiller Pump 3 (VWS-MP-03) [[Page 50737]] and associated equipment from the Auxiliary Building to the Annex Building, for each unit respectively. The proposed changes include information in the Combined License, Appendix C. An exemption request relating to the proposed changes to the AP1000 DCD Tier 1 is included with the request. 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 Central Chilled Water System (VWS) performs the nonsafety- related function of supplying chilled water to the heating, ventilation, and air conditioning (HVAC) systems. The only safety- related function of the VWS is to provide isolation of the VWS lines penetrating the containment. The low capacity VWS subsystem is non- seismically designed. The change to relocate an air cooled chiller pump and associated equipment and add a chemical feed tank to this pump does not adversely affect the capability of either low capacity VWS subsystem loop to perform the system design function. This change does not have an adverse impact on the response to anticipated transient or postulated accident conditions because the low capacity VWS subsystem is a nonsafety-related and non-seismic system. No safety-related structure, system, component (SSC) or function is involved with or affected by this change. The changes to the low capacity VWS subsystem do not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the plant-specific UFSAR [Updated Final Safety Analysis Report] are not affected. The proposed VWS change does not involve a change to the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the nonsafety-related low capacity VWS subsystem do not affect any safety-related equipment, nor do they add any new interfaces to safety-related SSCs. No system or design function or equipment qualification is affected by these changes. The changes do not introduce a new failure mode, malfunction or sequence of events that could affect 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 VWS is a nonsafety-related system that performs the defense- in-depth function of providing a reliable source of chilled water to various HVAC subsystems and unit coolers and the safety-related function of providing isolation of the VWS lines penetrating the containment. The changes to the VWS do not affect the VWS containment penetrations or any other safety related equipment or fission product barriers. The requested changes will not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Acting Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: May 27, 2016. A publicly-available version is in ADAMS under Accession No. ML16148A631. Description of amendment request: The amendment request proposes changes to the Combined License (COL), Appendix A, Technical Specifications (TSs), and Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information. Specifically, the proposed departures consist of changes to the UFSAR adding compensation for changes in reactor coolant density using the ``delta T'' power signal to the reactor coolant flow input signal for the low reactor coolant flow trip function of the Reactor Trip System (RTS). Additionally, TS Surveillance Requirement (SR) 3.3.1.3 is added to the surveillances required for the Reactor Coolant Flow[middot]Low reactor trip in TS Table 3.3.1-1, Function 7. 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 adds compensation, for changes in reactor coolant density using the [delta T] power signal, to the reactor coolant flow input signal for the low reactor coolant flow reactor trip function of the RTS. The proposed change also adds TS SR 3.3.1.3 to the surveillances required for the Reactor Coolant Flow- Low reactor trip specified in TS Table 3.3.1-1. SR 3.3.1.3 compares the calorimetric heat balance to the calculated [delta T] power in each Protection and Safety Monitoring System (PMS) division every 24 hours to assure acceptable [delta T] power calibration. As such, the surveillance is also required to support operability of the Reactor Coolant Flow-Low trip function. This change to the low reactor coolant flow trip input signal assures that the reactor will trip on low reactor coolant flow when the requisite conditions are met, and minimize spurious reactor trips and the accompanying plant transients. The change to the COL Appendix A Table 3.3.1-1 aligns the surveillance of the Reactor Coolant Flow-Low trip with the addition of the compensation, for changes in reactor coolant density using [delta T] power to the flow input signal to the trip. These changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) 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 and remains within response time assumed in the accident analysis. 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 or external events is not adversely affected, 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 change adds compensation, for changes in reactor coolant density using [delta T] power signal, to the reactor coolant flow input signal to the low reactor coolant flow reactor trip function of the RTS. The proposed change also adds TS [[Page 50738]] SR 3.3.1.3 to the surveillances required for the Reactor Coolant Flow-Low reactor trip specified in TS Table 3.3.1-1. SR 3.3.1.3 compares the calorimetric heat balance to the calculated [delta T] power in each PMS division every 24 hours to assure acceptable [delta T] power calibration. As such, the surveillance is also required to support operability of the Reactor Coolant Flow-Low trip function. The proposed change to the low reactor coolant flow reactor trip input signal does not alter the design function of the low flow reactor trip. The change to the COL Appendix A Table 3.3.1- 1 aligns the surveillance of the Reactor Coolant Flow-Low trip with the addition of compensation, for changes in reactor coolant density using [delta T] power to the flow input signal to the trip. Consequently, because the low reactor coolant flow trip functions are unchanged, there are no adverse effects that could create the possibility of a new or different kind of accident from any previously evaluated in the UFSAR. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 4. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change adds compensation, for changes in reactor coolant density using [delta T] power signal, to the reactor coolant flow input signal for the low reactor coolant flow trip function of the RTS. The proposed change also adds TS SR 3.3.1.3 to the surveillances required for the Reactor Coolant Flow-Low reactor trip specified in TS Table 3.3.1-1. SR 3.3.1.3 compares the calorimetric heat balance to the calculated [delta T] power in each PMS division every 24 hours to assure acceptable [delta T] power calibration. As such, the surveillance is also required to support operability of the Reactor Coolant Flow-Low trip function. 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. 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 Acting Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: June 14, 2016, as supplemented by letter dated July 1, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16166A409 and ML16183A394, respectively. 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* and associated Tier 2 information. Specifically, the proposed departures consist of changes to the UFSAR to revise the details of the structural design of auxiliary building floors within module CA20 at approximate design elevations of 82'-6'' and 92'-6''. 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 design functions of the auxiliary building floors are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the auxiliary building. The auxiliary building is a seismic Category I structure and is designed for dead, live, thermal, pressure, safe shutdown earthquake loads, and loads due to postulated pipe breaks. The proposed changes to UFSAR descriptions are intended to address changes in the detail design of floors in the auxiliary building. The thickness and strength of the auxiliary building floors are not reduced. As a result, the design function of the auxiliary building structure is not adversely affected by the proposed changes. 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 changes described 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 changes to UFSAR descriptions are proposed to address changes in the detail design of floors in the auxiliary building. The thickness, geometry, and strength of the structures are not adversely altered. The concrete and reinforcement materials are not altered. The properties of the concrete are not altered. The changes to the design details of the auxiliary building structure do not create any new accident precursors. As a result, the design function of the auxiliary building structure is not 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. The criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349 and AISC N690 and therefore maintains the margin of safety. Analysis of the connection design confirms that code provisions are appropriate to the floor to wall connection. The proposed changes to the UFSAR address changes in the detail design of floors in the auxiliary building. The proposed changes also incorporate the requirements for development and anchoring of headed reinforcement which were previously approved. There is no change to design requirements of the auxiliary building structure. There is no change to the method of evaluation from that used in the design basis calculations. There is not a significant change to the in structure response spectra. Therefore, the proposed amendment does not result in 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 Acting Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: June 3, 2016. A publicly-available version is in ADAMS under Accession No. ML16155A366. Description of amendment request: The amendment request proposes changes to correct editorial errors in Combined License (COL) Appendix C (and plant-specific Tier 1) and promote consistency with the Updated Final Safety Analysis Report (UFSAR) Tier 2 [[Page 50739]] information. Additionally, one of the proposed changes to plant- specific Tier 1 information also requires an involved change to UFSAR Tier 2 information. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Tier 1 material departures. The requested amendment also contains a proposed editorial correction to COL paragraph 2.D. 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 consistency and editorial Combined License (COL) Appendix C (and plant-specific Tier 1) and involved Tier 2 changes, along with one COL paragraph 2.D change, do not involve a technical change, (e.g. there is no design parameter or requirement, calculation, analysis, function or qualification change). No structure, system, component design or function would be affected. No design or safety analysis would be affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed consistency and editorial COL Appendix C (and plant-specific Tier 1) and involved Tier 2 changes, along with one COL paragraph 2.D change, would not affect the design or function of any structure, system, component (SSC), but will instead provide consistency between the SSC designs and functions currently presented in the Updated Final Safety Analysis Report (UFSAR) and the Tier 1 information. The proposed changes would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed consistency and editorial COL Appendix C (and plant-specific Tier 1) and involved Tier 2 update, along with one COL paragraph 2.D change, is non-technical, thus would not affect any design parameter, function or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit/ criterion is involved. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. 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 Acting Branch Chief: Jennifer Dixon-Herrity. Tennessee Valley Authority Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant (BFN), Unit 1, 2 and 3, Limestone County Alabama Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee Date of amendment request: April 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16105A287. Description of amendment request: The amendments would revise the BFN Units 1, 2, and 3, and the SQN, Units 1 and 2, Technical Specification (TS) 5.3, ``Unit Staff Qualifications,'' to delete the references to Regulatory Guide 1.8, Revision 2, and replace it with references to the TVA Nuclear Quality Assurance Plan (NQAP). The proposed changes would ensure consistent regulatory requirements regarding staff qualifications for the TVA nuclear fleet. The proposed changes would further allow TVA to implement standard procedures related to staff qualifications. Additionally, the proposed TS changes are consistent with the intent of NRC Administrative Letter 95-06 in that the relocated requirements are adequately controlled by 10 CFR 50, Appendix B, and the quality assurance change control process in 10 CFR 50.54(a). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated? Response: No. The Unit Staff Qualifications that are being removed from BFN TS 5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in Appendix B to the TVA NQAP and are consistent with the Watts Bar (WBN) Unit 1 and Unit 2 Technical Specifications (TS). Changes to the TVA NQAP are controlled by 10 CFR 50.54(a). These changes do not affect any of the design basis accidents. Therefore, the proposed changes do not involve an increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The Unit Staff Qualifications that are being removed from BFN TS 5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in Appendix B to the TVA NQAP and are consistent with the WBN Unit 1 and Unit 2 TS. Changes to the TVA NQAP are controlled by 10 CFR 50.54(a). These changes do not affect any of the design basis accidents. No modifications to any plant equipment are involved. There is no effect on system interactions made by these changes. 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 Unit Staff Qualifications that are being removed from BFN TS 5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in Appendix B to the TVA NQAP and are consistent with the WBN Unit 1 and Unit 2 TS. Changes to the TVA NQAP are controlled by 10 CFR 50.54(a). The margin of safety as reported in the basis for the TS is not reduced. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Acting Branch Chief: Tracy J. Orf. [[Page 50740]] Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee Date of amendment request: May 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16148A175. Description of amendment request: The amendments would modify the SQN, Units 1 and 2, Technical Specification (TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' by revising the acceptance criteria for the diesel generator (DG) steady-state frequency acceptance criteria specified in the TS Surveillance Requirements (SRs). The frequency would be changed to address the non-conservative TS recently identified. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated? Response: No. The DGs are required to be operable in the event of a design basis accident coincident with a loss of offsite power to mitigate the consequences of the accident. The DGs are not accident initiators and, therefore, these changes do not involve a significant increase in the probability of an accident previously evaluated. The accident analyses assume that at least the boards in one load group are provided with power either from the offsite circuits or the DGs. The change proposed in this license amendment request will continue to assure that the DGs have the capacity and capability to assume their maximum design basis accident loads. The proposed change does not significantly alter how the plant would mitigate an accident previously evaluated. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed change does not adversely affect the ability of structures, systems, and components (SSC) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed change does not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/ public radiation exposure. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a change in the plant design, system operation, or the use of the DGs. The proposed change requires the DGs to meet SR acceptance criteria that envelope the actual demand requirements for the DGs during design basis conditions. These revised acceptance criteria continue to demonstrate the capability and capacity of the DGs to perform their required functions. There are no new failure modes or mechanisms created due to testing the DGs within the proposed acceptance criteria. Testing of the DGs at the proposed acceptance criteria does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the proposed test loadings. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change will continue to demonstrate that the DGs meet the TS definition of operability, that is, the proposed acceptance criteria will continue to demonstrate that the DGs will perform their safety function. The proposed testing will also continue to demonstrate the capability and capacity of the DGs to supply their required loads for mitigating a design basis accident. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Acting Branch Chief: Tracy J. Orf. Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee Date of amendment request: June 7, 2016. A publicly-available version is in ADAMS under Accession No. ML16159A208. Description of amendment request: The amendments would revise the WBN, Unit 2, Technical Specification (TS) 3.7.10, ``Control Room Emergency Ventilation System (CREVS),'' to include specific shutdown Required Actions and associated Completion Times during conditions to be taken due to a tornado warning. The proposed TS changes would be consistent with the current TS 3.7.10 for WBN, Unit 1. Additionally, the amendments would revise several administrative-related inconsistencies identified in the WBN, Units 1 and 2, TSs. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes modify WBN Unit 1 TS 3.7.10 to resolve a potential conflict in applying the appropriate actions for not meeting the Required Action and associated Completion Time of Condition E and request administrative changes to correct inconsistencies in TS Applicability statements. The proposed changes do not affect the structures, systems, or components (SSCs) of the plant, affect plant operations, or any design function or an analysis that verifies the capability of an SSC to perform a design function. No change is being made to any of the previously evaluated accidents in the WBN Unit 1 Updated Final Safety Analysis Report (UFSAR) and the WBN Unit 2 FSAR [Final Safety Analysis Report]. These proposed changes are administrative or provide specific shutdown actions instead of using default shutdown actions. The proposed changes do not (1) require physical changes to plant systems, structures, or components; (2) prevent the safety function of any safety-related system, structure, or component during a design basis event; (3) alter, degrade, or prevent action described or assumed in any accident described in the WBN Unit 1 UFSAR and the WBN Unit 2 FSAR from being perform[ed] because the safety-related systems, structures, or components are not modified; (4) alter any assumptions previously made in evaluating radiological consequences; or (5) affect the integrity of any fission product barrier. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? [[Page 50741]] Response: No. The proposed changes do not introduce any new accident causal mechanisms, since no physical changes are being made to the plant, nor do they impact any plant systems that are potential accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed changes will have no effect on the availability, operability, or performance of safety-related systems and components. The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis. The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely affect plant-operating margins or the reliability of equipment credited in the safety analyses. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Sherry Quirk, Executive Vice President and General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., 6A West Tower, Knoxville, TN 37902. NRC Acting Branch Chief: Tracy J. Orf. III. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration. For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice. Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: August 18, 2015, as supplemented by letters dated September 29, 2015; February 5, 2016; April 28, 2016; and May 19, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15236A265 (Package), ML15272A443, ML16036A091, ML16119A326, and ML16141A048, respectively. Brief description of amendment request: The amendment would revise the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute document NEI 04-10, ``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). Additionally, a new program, the Surveillance Frequency Control Program, would be added to TS Section 6, ``Administrative Controls.'' Date of publication of individual notice in Federal Register: July 15, 2016 (81 FR 46119). Expiration date of individual notice: August 15, 2016 (public comments); September 13, 2016 (hearing requests). Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date of amendment request: May 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16138A247. Brief description of amendment request: The amendments would revise the Cyber Security Plan implementation schedule for Milestone 8 and revise the associated license condition in the Facility Operating Licenses. Date of publication of individual notice in the Federal Register: July 8, 2016 (81 FR 44665). Expiration date of individual notice: August 8, 2016 (public comments); September 6, 2016 (hearing requests). IV. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: October 2, 2015, as supplemented by letter dated March 23, 2016. Brief description of amendments: The amendments (1) revised the allowable test pressure band in the technical specification (TS) surveillance requirements (SRs) for the pump flow testing of the high pressure coolant injection system and the reactor core isolation system; (2) revised the surveillance frequency requirements for verifying the sodium pentaborate enrichment of the standby liquid control system; and (3) deleted SRs associated with verifying the manual transfer capability of the normal and alternate power supplies for certain motor-operated valves associated with the suppression pool spray and drywell spray sub-systems of the residual heat removal system. [[Page 50742]] Date of issuance: July 5, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendments Nos.: 308 (Unit 2) and 312 (Unit 3). A publicly- available version is in ADAMS under Accession No. ML16159A148; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: December 8, 2015 (80 FR 76320). The supplemental letter dated March 23, 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 July 5, 2016. No significant hazards consideration comments received: No. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: July 24, 2015. Brief description of amendment: The amendment revised Technical Specification 1.4, ``Frequency,'' by correcting Example 1.4-1 to be consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-485, ``Correct Example 1.4-1,'' Revision 0. In addition, the amendment revised Example 1.4-5 and Example 1.4-6 to be consistent with Amendment No. 258 to the Renewed Facility Operating License. Date of issuance: July 13, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 293. A publicly-available version is in ADAMS under Accession No. ML15246A408; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-49: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: November 10, 2015 (80 FR 69713). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 13, 2016. No significant hazards consideration comments received: No. South Carolina Electric and Gas Company and the South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina Date of amendment request: October 1, 2015. Brief description of amendment: The amendments consisted of changes to the Facility Combined License, Appendix C, ``Inspections, Tests, Analyses, and Acceptance Criteria [ITAAC].'' Specifically, the changes to the plant-specific Emergency Planning ITAAC removed and replaced current references to AP1000 Design Control Document Table 7.5-1, and Final Safety Analysis Report (FSAR) Table 7.5-201 on the post-accident monitoring system, with references to proposed updated FSAR Table 7.5-1 in Table C.3.8-1 for ITAAC Numbers C.3.8.01.01.01, C.3.8.01.05.01.05, and C.3.8.01.05.02.04. Date of issuance: May 2, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 46. A publicly-available version is in ADAMS under Package Accession No. ML16074A234. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Combined License Nos. NPF-93 and NPF-94: Amendments revised the Facility Combined Licenses. Date of initial notice in Federal Register: November 24, 2015 (80 FR 73241). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 2, 2016. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50- 425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: July 18, 2014, as supplemented by letters dated February 27, 2015; May 2, 2016; and June 14, 2016. Brief description of amendments: The amendments changed Technical Specification 3.9.4, ``Containment Penetrations,'' to allow containment penetrations to be un-isolated under administrative controls during core alterations or movement of irradiated fuel assemblies within containment by adopting a previously NRC-approved Technical Specification Task Force (TSTF) Change Traveler TSTF-312, Revision 1, ``Administratively Control Containment Penetrations.'' Date of issuance: July 15, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 181 (Unit 1) and 162 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16165A195; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: March 3, 2015 (80 FR 11480). The supplemental letters dated February 27, 2015; May 2, 2016; and June 14, 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 July 15, 2016. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 22nd day of July 2016. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2016-18290 Filed 8-1-16; 8:45 am] BILLING CODE 7590-01-P
| 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 September 1, 2016. A request for a hearing must be filed by October 3, 2016. | |
| Contact | Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected] | |
| FR Citation | 81 FR 50729 |