81 FR 50729 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 81, Issue 148 (August 2, 2016)
Federal Register Volume 81, Issue 148 (August 2, 2016)
| 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
![Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices 50729
NUCLEAR REGULATORY Regulatory Commission, Washington DC submissions to remove such information
COMMISSION 20555–0001; telephone: 301–415–1927, before making the comment
email: Lynn.Ronewicz@nrc.gov. submissions available to the public or
[NRC–2016–0151]
entering the comment into ADAMS.
I. Obtaining Information and I. Notice of Consideration of Issuance
Biweekly Notice; Applications and Submitting Comments of Amendments to Facility Operating
Amendments to Facility Operating
A. Obtaining Information Licenses and Combined Licenses and
Licenses and Combined Licenses
Proposed No Significant Hazards
Involving No Significant Hazards Please refer to Docket ID NRC–2016–
Consideration Determination
Considerations 0151, facility name, unit number(s), The Commission has made a
plant docket number, application date, proposed determination that the
AGENCY: Nuclear Regulatory
and subject when contacting the NRC following amendment requests involve
Commission.
about the availability of information for no significant hazards consideration.
ACTION: Biweekly notice. this action. You may obtain publicly- Under the Commission’s regulations in
SUMMARY: Pursuant to Section 189a. (2) available information related to this § 50.92 of title 10 of the Code of Federal
of the Atomic Energy Act of 1954, as action by any of the following methods: Regulations (10 CFR), this means that
amended (the Act), the U.S. Nuclear • Federal Rulemaking Web site: Go to operation of the facility in accordance
Regulatory Commission (NRC) is http://www.regulations.gov and search with the proposed amendment would
publishing this regular biweekly notice. for Docket ID NRC–2016–0151. not (1) involve a significant increase in
The Act requires the Commission to • NRC’s Agencywide Documents the probability or consequences of an
publish notice of any amendments Access and Management System accident previously evaluated, or (2)
issued, or proposed to be issued, and (ADAMS): You may obtain publicly- create the possibility of a new or
grants the Commission the authority to available documents online in the different kind of accident from any
issue and make immediately effective ADAMS Public Documents collection at accident previously evaluated; or (3)
any amendment to an operating license http://www.nrc.gov/reading-rm/ involve a significant reduction in a
or combined license, as applicable, adams.html. To begin the search, select margin of safety. The basis for this
upon a determination by the ‘‘ADAMS Public Documents’’ and then proposed determination for each
Commission that such amendment select ‘‘Begin Web-based ADAMS amendment request is shown below.
involves no significant hazards Search.’’ For problems with ADAMS, The Commission is seeking public
consideration, notwithstanding the please contact the NRC’s Public comments on this proposed
pendency before the Commission of a Document Room (PDR) reference staff at determination. Any comments received
request for a hearing from any person. 1–800–397–4209, 301–415–4737, or by within 30 days after the date of
This biweekly notice includes all email to pdr.resource@nrc.gov. The publication of this notice will be
notices of amendments issued, or ADAMS accession number for each considered in making any final
proposed to be issued from July 5, 2019, document referenced (if it is available in determination.
to July 19, 2016. The last biweekly ADAMS) is provided the first time that Normally, the Commission will not
notice was published on July 19, 2016 it is mentioned in this document. issue the amendment until the
(81 FR 46958). • NRC’s PDR: You may examine and expiration of 60 days after the date of
purchase copies of public documents at publication of this notice. The
DATES: Comments must be filed by
the NRC’s PDR, Room O1–F21, One Commission may issue the license
September 1, 2016. A request for a
White Flint North, 11555 Rockville amendment before expiration of the 60-
hearing must be filed by October 3,
Pike, Rockville, Maryland 20852. day period provided that its final
2016.
B. Submitting Comments determination is that the amendment
ADDRESSES: You may submit comments involves no significant hazards
by any of the following methods: Please include Docket ID NRC–2016– consideration. In addition, the
• Federal Rulemaking Web site: Go to 0151, facility name, unit number(s), Commission may issue the amendment
http://www.regulations.gov and search plant docket number, application date, prior to the expiration of the 30-day
for Docket ID NRC–2016–0151. Address and subject in your comment comment period if circumstances
questions about NRC dockets to Carol submission. change during the 30-day comment
Gallagher; telephone: 301–415–3463; The NRC cautions you not to include period such that failure to act in a
email: Carol.Gallagher@nrc.gov. For identifying or contact information that timely way would result, for example in
technical questions, contact the you do not want to be publicly derating or shutdown of the facility. If
individual listed in the FOR FURTHER disclosed in your comment submission. the Commission takes action prior to the
INFORMATION CONTACT section of this The NRC will post all comment expiration of either the comment period
document. submissions at http:// or the notice period, it will publish in
• Mail comments to: Cindy Bladey, www.regulations.gov as well as enter the the Federal Register a notice of
Office of Administration, Mail Stop: comment submissions into ADAMS. issuance. If the Commission makes a
OWFN–12–H08, U.S. Nuclear The NRC does not routinely edit final no significant hazards
Regulatory Commission, Washington, comment submissions to remove consideration determination, any
DC 20555–0001. identifying or contact information. hearing will take place after issuance.
For additional direction on obtaining If you are requesting or aggregating The Commission expects that the need
information and submitting comments, comments from other persons for
mstockstill on DSK3G9T082PROD with NOTICES
to take this action will occur very
see ‘‘Obtaining Information and submission to the NRC, then you should infrequently.
Submitting Comments’’ in the inform those persons not to include
SUPPLEMENTARY INFORMATION section of identifying or contact information that A. Opportunity To Request a Hearing
this document. they do not want to be publicly and Petition for Leave To Intervene
FOR FURTHER INFORMATION CONTACT: disclosed in their comment submission. Within 60 days after the date of
Lynn Ronewicz, Office of Nuclear Your request should state that the NRC publication of this notice, any person(s)
Reactor Regulation, U.S. Nuclear does not routinely edit comment whose interest may be affected by this
VerDate Sep<11>2014 18:35 Aug 01, 2016 Jkt 238001 PO 00000 Frm 00048 Fmt 4703 Sfmt 4703 E:\FR\FM\02AUN1.SGM 02AUN1](https://thefederalregister.org/doc/81_FR_50877/page/1.svg)
![Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices 50733
proposed change does not alter the types of ADAMS under Accession No. consumption analyses demonstrate that the
inservice testing performed. In most cases, ML16190A118. EDG design continues to satisfy its safety
the frequency of inservice testing is Description of amendment request: function. The design basis limits for the
unchanged. However, the frequency of The amendment would update the accident and transient analyses will continue
testing would not result in a new or different to meet their design criteria.
Technical Specifications to revise the
kind of accident from any previously Therefore, the proposed change does not
evaluated since the testing methods are not emergency diesel generator (EDG) involve a significant reduction in a margin of
altered. engine-mounted fuel tank minimum safety.
Therefore, the proposed change does not volume from 200 gallons of fuel each to
create the possibility of a new or different 238 gallons of fuel each. The NRC staff has reviewed the
kind of accident from any previously Basis for proposed no significant licensee’s analysis and, based on this
evaluated. hazards consideration determination: review, it appears that the three
3. Does the proposed change involve a As required by 10 CFR 50.91(a), the standards of 50.92(c) are satisfied.
significant reduction in a margin of safety? licensee has provided its analysis of the Therefore, the NRC staff proposes to
Response: No issue of no significant hazards determine that the amendment request
The proposed change eliminates some involves no significant hazards
consideration, which is presented
requirements from the TS in lieu of consideration.
requirements in the ASME Code, as modified below:
Attorney for licensee: William S.
by use of Code Case OMN–20. Compliance 1. Does the proposed amendment involve Blair, Managing Attorney—Nuclear,
with the ASME Code is required by 10 CFR a significant increase in the probability or
50.55a. The proposed change also allows consequences of an accident previously Florida Power & Light Company, 700
inservice tests with frequencies greater than evaluated? Universe Boulevard, MS LAW/JB, Juno
2 years to be extended by 6 months to Response: No. Beach, FL 33408–0420.
facilitate test scheduling and consideration of The EDGs engine-mounted fuel oil tanks NRC Acting Branch Chief: Tracy J.
plant operating conditions that may not be are part of a system used to mitigate the Orf.
suitable for performance of the required consequences of an accident and do not
testing. The testing frequency extension will increase the probability of an accident Pacific Gas and Electric Company,
not affect the ability of the components to previously evaluated. The increase in Docket Nos. 50–275 and 50–323, Diablo
respond to an accident as the components are minimum fuel oil requirements enables Canyon Nuclear Power Plant, Units 1
required to be operable during the testing operation of the EDGs to remain unchanged and 2, San Luis Obispo County,
period extension. The proposed change will for ULSD [ultra low sulfur diesel] fuel oil, California
eliminate the existing TS SR 3.0.3 allowance thus the EDGs continue to be capable of
to defer performance of missed inservice tests performing their design functions. Date of amendment request: May 12,
up to the duration of the specified testing Acceptance criteria continue to be satisfied. 2016. A publicly-available version is in
frequency, and instead will require an Accordingly, the proposed change does not ADAMS under Package Accession No.
assessment of the missed test on equipment increase the consequences of an accident. ML16146A100.
operability. This assessment will consider Therefore, the proposed change does not Description of amendment request:
the effect on a margin of safety (equipment involve a significant increase in the The amendments would revise
operability). Should the component be probability or consequences of an accident Technical Specification (TS) 5.5.6,
inoperable, the Technical Specifications previously evaluated.
2. Does the proposed amendment create ‘‘Containment Leakage Rate Testing
provide actions to ensure that the margin of
the possibility of a new or different kind of Program,’’ to allow the following:
safety is protected. The proposed change also
eliminates a statement that nothing in the accident from any accident previously • Increase in the existing 10 CFR part
ASME Code should be construed to evaluated? 50, Appendix J, ‘‘Primary Reactor
supersede the requirements of any TS. The Response: No. Containment Leakage Testing for Water-
NRC has determined that statement to be No new accident scenarios, failure Cooled Power Reactors,’’ Type A test
incorrect. However, elimination of the mechanisms, or limiting single failures are interval from 10 years to 15 years in
statement will have no effect on plant introduced as a result of the increase in accordance with Nuclear Energy
operation or safety. minimum EDGs engine-mounted fuel oil tank Institute (NEI) 94–01, Revision 2–A,
Therefore, the proposed change does not volume. The proposed change has no adverse
effect on any safety-related system and does
‘‘Industry Guideline for Implementing
involve a significant reduction in a margin of Performance-Based Option of 10 CFR
safety. not change the performance or integrity of
any safety-related equipment. No new safety- part 50, Appendix J,’’ October 2008
The NRC staff has reviewed the related equipment is being added or replaced (ADAMS Accession No. ML100620847).
licensee’s analysis and, based on this as a result of the proposed change. • Adopt the use of American National
review, it appears that the three Therefore, the proposed change does not Standards Institute/American Nuclear
standards of 10 CFR 50.92(c) are create the possibility of a new or different Society (ANSI/ANS) 56.8–2002,
satisfied. Therefore, the NRC staff kind of accident from any previously ‘‘Containment System Leakage Testing
proposes to determine that the evaluated. Requirements,’’ as referenced in NEI 94–
3. Does the proposed amendment involve
amendment request involves no 01, Revision 2–A.
a significant reduction in a margin of safety?
significant hazards consideration. Response: No. • Adopt an allowable test interval
Attorney for licensee: David W. The calculation for EDG fuel consumption extension of 9 months, which is shorter
Jenkins, FirstEnergy Nuclear Operating shows that with the minimum day tank than the currently allowed 25 percent
Company, FirstEnergy Corporation, 76 volume of 238 gallons of ULSD fuel, the grace, for the 10 CFR 50, Appendix J,
South Main Street, Mail Stop A–GO–15, requirement for two day tanks to provide a Type A, Type B, and Type C leakage
Akron, OH 44308. usable volume which is sufficient for at least tests in accordance with NEI 94–01,
NRC Acting Branch Chief: G. Edward 1 hour 100% load operation of one diesel Revision 2–A.
mstockstill on DSK3G9T082PROD with NOTICES
Miller. generator set, plus a minimum margin of The proposed changes would revise
10% is met. The day tank minimum volumes
TS 5.5.16 to replace the reference to
Florida Power & Light Company, et al., with the DOST [diesel oil storage tank]
minimum volume is sufficient for the EDG NRC Regulatory Guide 1.163,
Docket No. 50–389, St. Lucie Plant, Unit ‘‘Performance-Based Containment Leak-
No. 2, St. Lucie County, Florida loading increase due to potential operation at
the upper frequency limit of 60.6 HZ [Hertz] Test Program,’’ September 1995
Date of amendment request: June 21, (60 HZ, +1%) and the EPU [extended power (ADAMS Accession No. ML003740058),
2016. A publicly-available version is in uprate] requirements. The EDG fuel and 10 CFR 50, Appendix J, Option B,
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![Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices 50735
adversely affected, nor do the changes South Carolina Electric and Gas Response: No.
described create any new accident Company and South Carolina Public The proposed change is to revise the
precursors. Service Authority, Docket Nos. 52–027 thickness of the floor above the CCS Valve
Therefore, the proposed amendment does room in the auxiliary building. The proposed
and 52–028, Virgil C. Summer Nuclear
not involve a significant increase in the changes do not change the design
Station (VCSNS), Units 2 and 3, requirements of the nuclear island structures.
probability or consequences of an accident
previously evaluated.
Fairfield County, South Carolina The proposed changes do not change the
2. Does the proposed amendment create Date of amendment request: July 5, design function, support, design, or operation
the possibility of a new or different kind of 2016. A publicly-available version is in of mechanical and fluid systems. The
accident from any accident previously ADAMS under Accession No. proposed changes do not result in a new
evaluated? failure mechanism for the nuclear island
ML16187A392.
Response: No. structures or new accident precursors. As a
Description of amendment request: result, the design function of the nuclear
The changes to UFSAR descriptions are The amendment request relates to
proposed to address changes in the detail island structures is not adversely affected by
changes to the slab thickness between the proposed change.
design of floors in the auxiliary building. The
thickness, geometry, and strength of the Column Lines I to J–1 and 2 to 4 at plant Therefore, the proposed amendment does
structures are not adversely altered. The elevation 153′-0″. The changes involve not create the possibility of a new or different
concrete and reinforcement materials are not changes to incorporated AP1000 Design kind of accident previously evaluated.
altered. The properties of the concrete are not Control Document Tier 1 information 3. Does the proposed amendment involve
altered. The changes to the design details of and corresponding departures to Tier 2* a significant reduction in a margin of safety?
the auxiliary building structure do not create Updated Final Safety Analysis Report Response: No.
any new accident precursors. As a result, the No safety analysis or design basis
information and conforming changes to acceptance limit/criterion is challenged or
design function of the auxiliary building the Combined License, Appendix C.
structure is not adversely affected by the exceeded by the proposed changes, thus, no
Basis for proposed no significant margin of safety is reduced.
proposed changes. hazards consideration determination:
Therefore, the proposed amendment does Therefore, the proposed amendment does
As required by 10 CFR 50.91(a), the not involve a significant reduction in a
not create the possibility of a new or different
kind of accident from any accident licensee has provided its analysis of the margin of safety previously evaluated.
previously evaluated. issue of no significant hazards The NRC staff has reviewed the
3. Does the proposed amendment involve consideration, which is presented licensee’s analysis and, based on this
a significant reduction in a margin of safety? below, with NRC staff edits in square review, it appears that the three
Response: No. brackets: standards of 10 CFR 50.92(c) are
The criteria and requirements of American 1. Does the proposed amendment involve satisfied. Therefore, the NRC staff
Concrete Institute (ACI) 349 and American a significant increase in the probability or
Institute of Steel Construction (AISC) N690 proposes to determine that the
consequences of an accident previously amendment request involves no
provide a margin of safety to structural evaluated?
failure. The design of the auxiliary building significant hazards consideration.
Response: No.
structure conforms to criteria and Attorney for licensee: Ms. Kathryn M.
The design functions of the nuclear island
requirements in ACI 349 and AISC N690 and structures are to provide support, protection, Sutton, Morgan, Lewis & Bockius LLC,
therefore maintains the margin of safety. and separation for the seismic Category I 1111 Pennsylvania Avenue NW.,
Analysis of the connection design confirms mechanical and electrical equipment located Washington, DC 20004–2514.
that code provisions are appropriate to the in the nuclear island. The nuclear island Acting NRC Branch Chief: Jennifer
floor to wall connection. The proposed structures are structurally designed to meet Dixon-Herrity.
changes to the UFSAR address changes in the seismic Category I requirements as defined in
detail design of floors in the auxiliary Regulatory Guide 1.29. The change of the Southern California Edison Company, et
building. The proposed changes also thickness of the floor above the [Component al., Docket Nos. 50–361 and 50–362,
incorporate the requirements for Cooling Water System (CCS)] Valve room in San Onofre Nuclear Generating Station,
development and anchoring of headed the auxiliary building meets criteria and Units 2 and 3, San Diego County,
reinforcement which were previously requirements of American Concrete Institute California
approved. There is no change to design (ACI) 349 and American Institute of Steel
requirements of the auxiliary building Construction (AISC) N690, does not have an
Date of amendment request: June 16,
structure. There is no change to the method adverse impact on the response of the 2016. A publicly-available version is in
of evaluation from that used in the design nuclear island structures to safe shutdown ADAMS under Accession No.
basis calculations. There is not a significant earthquake ground motions or loads due to ML16172A075.
change to the in structure response spectra. anticipated transients or postulated accident Description of amendment request:
Therefore, the proposed amendment does conditions. The proposed changes do not The amendments would extend the
not result in a significant reduction in a impact the support, design, or operation of scheduled implementation date for
margin of safety. mechanical and fluid systems. There is no Milestone 8 of the San Onofre Nuclear
change to plant systems or the response of Generating Station, Units 2 and 3, Cyber
The NRC staff has reviewed the systems to postulated accident conditions. Security Plan to December 31, 2019, in
licensee’s analysis and, based on this There is no change to the predicted
order to more fully reflect the
review, it appears that the three radioactive releases due to normal operation
or postulated accident conditions. The plant permanent shutdown status of the
standards of 10 CFR 50.92(c) are
response to previously evaluated accidents or facility and accommodate ongoing
satisfied. Therefore, the NRC staff
external events is not adversely affected, nor decommissioning activities.
proposes to determine that the Basis for proposed no significant
does the change described create any new
amendment request involves no accident precursors. hazards consideration determination:
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significant hazards consideration. Therefore, the proposed amendment does As required by 10 CFR 50.91(a), the
Attorney for licensee: M. Stanford not involve a significant increase in the licensee has provided its analysis of the
Blanton, Balch & Bingham LLP, 1710 probability or consequences of an accident issue of no significant hazards
Sixth Avenue North, Birmingham, AL previously evaluated. consideration, which is presented
35203–2015. 2. Does the proposed amendment create
the possibility of a new or different kind of
below:
NRC Acting Branch Chief: Jennifer accident from any accident previously 1. Does the proposed change involve a
Dixon-Herrity. evaluated? significant increase in the probability or
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![Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices 50737
and associated equipment from the kind of accident from any accident licensee has provided its analysis of the
Auxiliary Building to the Annex previously evaluated. issue of no significant hazards
Building, for each unit respectively. The 3. Does the proposed amendment involve consideration, which is presented
a significant reduction in a margin of safety?
proposed changes include information below:
Response: No.
in the Combined License, Appendix C. The VWS is a nonsafety-related system that 1. Does the proposed amendment involve
An exemption request relating to the performs the defense-in-depth function of a significant increase in the probability or
proposed changes to the AP1000 DCD providing a reliable source of chilled water consequences of an accident previously
Tier 1 is included with the request. to various HVAC subsystems and unit coolers evaluated?
Basis for proposed no significant and the safety-related function of providing Response: No.
hazards consideration determination: isolation of the VWS lines penetrating the The proposed change adds compensation,
As required by 10 CFR 50.91(a), the containment. The changes to the VWS do not for changes in reactor coolant density using
affect the VWS containment penetrations or the [delta T] power signal, to the reactor
licensee has provided its analysis of the
any other safety related equipment or fission coolant flow input signal for the low reactor
issue of no significant hazards product barriers. The requested changes will coolant flow reactor trip function of the RTS.
consideration, which is presented not affect any design code, function, design The proposed change also adds TS SR 3.3.1.3
below: analysis, safety analysis input or result, or to the surveillances required for the Reactor
1. Does the proposed amendment involve design/safety margin. No safety analysis or Coolant Flow-Low reactor trip specified in
a significant increase in the probability or design basis acceptance limit/criterion is TS Table 3.3.1–1. SR 3.3.1.3 compares the
consequences of an accident previously challenged or exceeded by the requested calorimetric heat balance to the calculated
evaluated? changes. [delta T] power in each Protection and Safety
Response: No. Therefore, the proposed amendment does Monitoring System (PMS) division every 24
The Central Chilled Water System (VWS) not involve a significant reduction in a hours to assure acceptable [delta T] power
performs the nonsafety-related function of margin of safety. calibration. As such, the surveillance is also
supplying chilled water to the heating, required to support operability of the Reactor
The NRC staff has reviewed the
ventilation, and air conditioning (HVAC) Coolant Flow-Low trip function. This change
systems. The only safety-related function of
licensee’s analysis and, based on this to the low reactor coolant flow trip input
the VWS is to provide isolation of the VWS review, it appears that the three signal assures that the reactor will trip on
lines penetrating the containment. The low standards of 10 CFR 50.92(c) are low reactor coolant flow when the requisite
capacity VWS subsystem is non-seismically satisfied. Therefore, the NRC staff conditions are met, and minimize spurious
designed. The change to relocate an air proposes to determine that the reactor trips and the accompanying plant
cooled chiller pump and associated amendment request involves no transients. The change to the COL Appendix
equipment and add a chemical feed tank to significant hazards consideration. A Table 3.3.1–1 aligns the surveillance of the
this pump does not adversely affect the Attorney for licensee: M. Stanford Reactor Coolant Flow-Low trip with the
capability of either low capacity VWS addition of the compensation, for changes in
subsystem loop to perform the system design
Blanton, Balch & Bingham LLP, 1710
reactor coolant density using [delta T] power
function. This change does not have an Sixth Avenue North, Birmingham, AL
to the flow input signal to the trip. These
adverse impact on the response to 35203–2015. changes do not affect the operation of any
anticipated transient or postulated accident NRC Acting Branch Chief: Jennifer systems or equipment that initiate an
conditions because the low capacity VWS Dixon-Herrity. analyzed accident or alter any structures,
subsystem is a nonsafety-related and non- systems, and components (SSC) accident
seismic system. No safety-related structure, Southern Nuclear Operating Company,
initiator or initiating sequence of events.
system, component (SSC) or function is Docket Nos. 52–025 and 52–026, Vogtle
These changes have no adverse impact on
involved with or affected by this change. The Electric Generating Plant, Units 3 and 4, the support, design, or operation of
changes to the low capacity VWS subsystem Burke County, Georgia mechanical and fluid systems. The response
do not involve an interface with any SSC of systems to postulated accident conditions
accident initiator or initiating sequence of Date of amendment request: May 27,
2016. A publicly-available version is in is not adversely affected and remains within
events, and thus, the probabilities of the response time assumed in the accident
accidents evaluated in the plant-specific ADAMS under Accession No.
analysis. There is no change to the predicted
UFSAR [Updated Final Safety Analysis ML16148A631. radioactive releases due to normal operation
Report] are not affected. The proposed VWS Description of amendment request: or postulated accident conditions.
change does not involve a change to the The amendment request proposes Consequently, the plant response to
predicted radiological releases due to changes to the Combined License (COL),
postulated accident conditions, thus, the previously evaluated accidents or external
Appendix A, Technical Specifications events is not adversely affected, nor does the
consequences of the accidents evaluated in
the UFSAR are not affected. (TSs), and Updated Final Safety proposed change create any new accident
Therefore, the proposed amendment does Analysis Report (UFSAR) in the form of precursors.
not involve a significant increase in the departures from the incorporated plant- Therefore, the proposed amendment does
probability or consequences of an accident specific Design Control Document Tier not involve a significant increase in the
previously evaluated. 2 information. Specifically, the probability or consequences of an accident
2. Does the proposed amendment create previously evaluated.
proposed departures consist of changes 2. Does the proposed amendment create
the possibility of a new or different kind of to the UFSAR adding compensation for
accident from any accident previously the possibility of a new or different kind of
evaluated?
changes in reactor coolant density using accident from any accident previously
Response: No. the ‘‘delta T’’ power signal to the reactor evaluated?
The proposed changes to the nonsafety- coolant flow input signal for the low Response: No.
related low capacity VWS subsystem do not reactor coolant flow trip function of the The proposed changes do not affect the
affect any safety-related equipment, nor do Reactor Trip System (RTS). operation of any systems or equipment that
they add any new interfaces to safety-related Additionally, TS Surveillance may initiate a new or different kind of
mstockstill on DSK3G9T082PROD with NOTICES
SSCs. No system or design function or Requirement (SR) 3.3.1.3 is added to the accident, or alter any SSC such that a new
equipment qualification is affected by these accident initiator or initiating sequence of
surveillances required for the Reactor events is created. The proposed change adds
changes. The changes do not introduce a new
failure mode, malfunction or sequence of Coolant Flow·Low reactor trip in TS compensation, for changes in reactor coolant
events that could affect safety related Table 3.3.1–1, Function 7. density using [delta T] power signal, to the
equipment. Basis for proposed no significant reactor coolant flow input signal to the low
Therefore, the proposed amendment does hazards consideration determination: reactor coolant flow reactor trip function of
not create the possibility of a new or different As required by 10 CFR 50.91(a), the the RTS. The proposed change also adds TS
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![50738 Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices
SR 3.3.1.3 to the surveillances required for Southern Nuclear Operating Company, thickness, geometry, and strength of the
the Reactor Coolant Flow-Low reactor trip Docket Nos. 52–025 and 52–026, Vogtle structures are not adversely altered. The
specified in TS Table 3.3.1–1. SR 3.3.1.3 Electric Generating Plant, Units 3 and 4, concrete and reinforcement materials are not
compares the calorimetric heat balance to the altered. The properties of the concrete are not
Burke County, Georgia
calculated [delta T] power in each PMS altered. The changes to the design details of
division every 24 hours to assure acceptable Date of amendment request: June 14, the auxiliary building structure do not create
[delta T] power calibration. As such, the 2016, as supplemented by letter dated any new accident precursors. As a result, the
July 1, 2016. Publicly-available versions design function of the auxiliary building
surveillance is also required to support
are in ADAMS under Accession Nos. structure is not adversely affected by the
operability of the Reactor Coolant Flow-Low
proposed changes.
trip function. The proposed change to the ML16166A409 and ML16183A394,
Therefore, the proposed amendment does
low reactor coolant flow reactor trip input respectively. not create the possibility of a new or different
signal does not alter the design function of Description of amendment request: kind of accident from any accident
the low flow reactor trip. The change to the The amendment request proposes previously evaluated.
COL Appendix A Table 3.3.1–1 aligns the changes to the Updated Final Safety 3. Does the proposed amendment involve
surveillance of the Reactor Coolant Flow-Low Analysis Report (UFSAR) in the form of a significant reduction in a margin of safety?
trip with the addition of compensation, for departures from the incorporated plant- Response: No.
changes in reactor coolant density using specific Design Control Document Tier The criteria and requirements of American
[delta T] power to the flow input signal to the 2* and associated Tier 2 information. Concrete Institute (ACI) 349 and American
trip. Consequently, because the low reactor Institute of Steel Construction (AISC) N690
Specifically, the proposed departures provide a margin of safety to structural
coolant flow trip functions are unchanged, consist of changes to the UFSAR to
there are no adverse effects that could create failure. The design of the auxiliary building
revise the details of the structural design structure conforms to criteria and
the possibility of a new or different kind of of auxiliary building floors within requirements in ACI 349 and AISC N690 and
accident from any previously evaluated in
module CA20 at approximate design therefore maintains the margin of safety.
the UFSAR. Analysis of the connection design confirms
elevations of 82′-6″ and 92′-6″.
Therefore, the proposed amendment does that code provisions are appropriate to the
Basis for proposed no significant
not create the possibility of a new or different floor to wall connection. The proposed
kind of accident from any accident
hazards consideration determination:
As required by 10 CFR 50.91(a), the changes to the UFSAR address changes in the
previously evaluated. detail design of floors in the auxiliary
4. Does the proposed amendment involve licensee has provided its analysis of the building. The proposed changes also
a significant reduction in a margin of safety? issue of no significant hazards incorporate the requirements for
Response: No. consideration, which is presented development and anchoring of headed
The proposed change adds compensation, below: reinforcement which were previously
for changes in reactor coolant density using 1. Does the proposed amendment involve approved. There is no change to design
[delta T] power signal, to the reactor coolant a significant increase in the probability or requirements of the auxiliary building
flow input signal for the low reactor coolant consequences of an accident previously structure. There is no change to the method
flow trip function of the RTS. The proposed evaluated? of evaluation from that used in the design
change also adds TS SR 3.3.1.3 to the Response: No. basis calculations. There is not a significant
surveillances required for the Reactor The design functions of the auxiliary change to the in structure response spectra.
Coolant Flow-Low reactor trip specified in building floors are to provide support, Therefore, the proposed amendment does
protection, and separation for the seismic not result in a significant reduction in a
TS Table 3.3.1–1. SR 3.3.1.3 compares the
Category I mechanical and electrical margin of safety.
calorimetric heat balance to the calculated
[delta T] power in each PMS division every equipment located in the auxiliary building. The NRC staff has reviewed the
24 hours to assure acceptable [delta T] power The auxiliary building is a seismic Category licensee’s analysis and, based on this
calibration. As such, the surveillance is also I structure and is designed for dead, live, review, it appears that the three
required to support operability of the Reactor thermal, pressure, safe shutdown earthquake
loads, and loads due to postulated pipe standards of 10 CFR 50.92(c) are
Coolant Flow-Low trip function. The satisfied. Therefore, the NRC staff
breaks. The proposed changes to UFSAR
proposed changes do not alter any applicable proposes to determine that the
descriptions are intended to address changes
design codes, code compliance, design in the detail design of floors in the auxiliary amendment request involves no
function, or safety analysis. Consequently, no building. The thickness and strength of the significant hazards consideration.
safety analysis or design basis acceptance auxiliary building floors are not reduced. As Attorney for licensee: M. Stanford
limit/criterion is challenged or exceeded by a result, the design function of the auxiliary Blanton, Balch & Bingham LLP, 1710
the proposed change, thus the margin of building structure is not adversely affected Sixth Avenue North, Birmingham, AL
safety is not reduced. by the proposed changes. There is no change
Therefore, the proposed amendment does
35203–2015.
to plant systems or the response of systems NRC Acting Branch Chief: Jennifer
not involve a significant reduction in a to postulated accident conditions. There is
margin of safety. no change to the predicted radioactive
Dixon-Herrity.
releases due to postulated accident Southern Nuclear Operating Company,
The NRC staff has reviewed the conditions. The plant response to previously Docket Nos. 52–025 and 52–026, Vogtle
licensee’s analysis and, based on this evaluated accidents or external events is not Electric Generating Plant, Units 3 and 4,
review, it appears that the three adversely affected, nor do the changes
Burke County, Georgia
standards of 10 CFR 50.92(c) are described create any new accident
satisfied. Therefore, the NRC staff precursors. Therefore, the proposed Date of amendment request: June 3,
proposes to determine that the amendment does not involve a significant 2016. A publicly-available version is in
increase in the probability or consequences ADAMS under Accession No.
amendment request involves no of an accident previously evaluated. ML16155A366.
mstockstill on DSK3G9T082PROD with NOTICES
significant hazards consideration. 2. Does the proposed amendment create Description of amendment request:
Attorney for licensee: M. Stanford the possibility of a new or different kind of
The amendment request proposes
Blanton, Balch & Bingham LLP, 1710 accident from any accident previously
evaluated? changes to correct editorial errors in
Sixth Avenue North, Birmingham, AL Combined License (COL) Appendix C
35203–2015. Response: No.
The changes to UFSAR descriptions are (and plant-specific Tier 1) and promote
NRC Acting Branch Chief: Jennifer proposed to address changes in the detail consistency with the Updated Final
Dixon-Herrity. design of floors in the auxiliary building. The Safety Analysis Report (UFSAR) Tier 2
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![50740 Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices
Tennessee Valley Authority, Docket probability or consequences of an accident ADAMS under Accession No.
Nos. 50–327 and 50–328, Sequoyah previously evaluated. ML16159A208.
Nuclear Plant (SQN), Units 1 and 2, 2. Does the proposed amendment create Description of amendment request:
the possibility of a new or different kind of The amendments would revise the
Hamilton County, Tennessee accident from any accident previously
Date of amendment request: May 26, evaluated? WBN, Unit 2, Technical Specification
2016. A publicly-available version is in Response: No. (TS) 3.7.10, ‘‘Control Room Emergency
ADAMS under Accession No. The proposed change does not involve a Ventilation System (CREVS),’’ to
ML16148A175. change in the plant design, system operation, include specific shutdown Required
or the use of the DGs. The proposed change Actions and associated Completion
Description of amendment request:
requires the DGs to meet SR acceptance Times during conditions to be taken due
The amendments would modify the criteria that envelope the actual demand
SQN, Units 1 and 2, Technical to a tornado warning. The proposed TS
requirements for the DGs during design basis
Specification (TS) 3.8.1, ‘‘AC conditions. These revised acceptance criteria
changes would be consistent with the
[Alternating Current] Sources— continue to demonstrate the capability and current TS 3.7.10 for WBN, Unit 1.
Operating,’’ by revising the acceptance capacity of the DGs to perform their required Additionally, the amendments would
criteria for the diesel generator (DG) functions. There are no new failure modes or revise several administrative-related
steady-state frequency acceptance mechanisms created due to testing the DGs inconsistencies identified in the WBN,
within the proposed acceptance criteria. Units 1 and 2, TSs.
criteria specified in the TS Surveillance Testing of the DGs at the proposed
Requirements (SRs). The frequency Basis for proposed no significant
acceptance criteria does not involve any hazards consideration determination:
would be changed to address the non- modification in the operational limits or
conservative TS recently identified. As required by 10 CFR 50.91(a), the
physical design of plant systems. There are
Basis for proposed no significant no new accident precursors generated due to licensee has provided its analysis of the
hazards consideration determination: the proposed test loadings. issue of no significant hazards
As required by 10 CFR 50.91(a), the Therefore, the proposed change does not consideration, which is presented
licensee has provided its analysis of the create the possibility of a new or different below:
issue of no significant hazards kind of accident from any accident 1. Does the proposed amendment involve
previously evaluated. a significant increase in the probability or
consideration, which is presented 3. Does the proposed amendment involve
below: consequences of an accident previously
a significant reduction in a margin of safety? evaluated?
1. Does the proposed amendment involve Response: No. Response: No.
a significant increase in the probability or The proposed change will continue to The proposed changes modify WBN Unit 1
consequence of an accident previously demonstrate that the DGs meet the TS TS 3.7.10 to resolve a potential conflict in
evaluated? definition of operability, that is, the proposed applying the appropriate actions for not
Response: No. acceptance criteria will continue to meeting the Required Action and associated
The DGs are required to be operable in the demonstrate that the DGs will perform their Completion Time of Condition E and request
event of a design basis accident coincident safety function. The proposed testing will administrative changes to correct
with a loss of offsite power to mitigate the also continue to demonstrate the capability inconsistencies in TS Applicability
consequences of the accident. The DGs are and capacity of the DGs to supply their statements.
not accident initiators and, therefore, these required loads for mitigating a design basis The proposed changes do not affect the
changes do not involve a significant increase accident. structures, systems, or components (SSCs) of
in the probability of an accident previously The proposed change does not alter the the plant, affect plant operations, or any
evaluated. manner in which safety limits, limiting safety design function or an analysis that verifies
The accident analyses assume that at least system settings or limiting conditions for
the capability of an SSC to perform a design
the boards in one load group are provided operation are determined. The safety analysis
function. No change is being made to any of
with power either from the offsite circuits or acceptance criteria are not affected by this
the previously evaluated accidents in the
the DGs. The change proposed in this license change. The proposed change will not result
WBN Unit 1 Updated Final Safety Analysis
amendment request will continue to assure in plant operation in a configuration outside
that the DGs have the capacity and capability Report (UFSAR) and the WBN Unit 2 FSAR
the design basis.
to assume their maximum design basis [Final Safety Analysis Report]. These
Therefore, the proposed change does not
accident loads. The proposed change does proposed changes are administrative or
involve a significant reduction in a margin of
not significantly alter how the plant would provide specific shutdown actions instead of
safety.
mitigate an accident previously evaluated. using default shutdown actions.
The proposed change does not adversely The NRC staff has reviewed the The proposed changes do not (1) require
affect accident initiators or precursors nor licensee’s analysis and, based on this physical changes to plant systems, structures,
alter the design assumptions, conditions, and review, it appears that the three or components; (2) prevent the safety
configuration of the facility or the manner in standards of 10 CFR 50.92(c) are function of any safety-related system,
which the plant is operated and maintained. satisfied. Therefore, the NRC staff structure, or component during a design basis
The proposed change does not adversely event; (3) alter, degrade, or prevent action
proposes to determine that the described or assumed in any accident
affect the ability of structures, systems, and amendment request involves no
components (SSC) to perform their intended described in the WBN Unit 1 UFSAR and the
safety function to mitigate the consequences significant hazards consideration. WBN Unit 2 FSAR from being perform[ed]
of an initiating event within the assumed Attorney for licensee: General because the safety-related systems,
acceptance limits. The proposed change does Counsel, Tennessee Valley Authority, structures, or components are not modified;
not affect the source term, containment 400 West Summit Hill Drive, 6A West (4) alter any assumptions previously made in
isolation, or radiological release assumptions Tower, Knoxville, TN 37902. evaluating radiological consequences; or (5)
used in evaluating the radiological NRC Acting Branch Chief: Tracy J. affect the integrity of any fission product
consequences of any accident previously Orf. barrier.
mstockstill on DSK3G9T082PROD with NOTICES
evaluated. Further, the proposed change does Therefore, the proposed change does not
not increase the types and amounts of Tennessee Valley Authority, Docket involve a significant increase in the
radioactive effluent that may be released Nos. 50–390 and 50–391, Watts Bar probability or consequences of an accident
offsite, nor significantly increase individual Nuclear Plant (WBN), Units 1 and 2, previously evaluated.
or cumulative occupational/public radiation Rhea County, Tennessee 2. Does the proposed amendment create
exposure. the possibility of a new or different kind of
Therefore, the proposed change does not Date of amendment request: June 7, accident from any accident previously
involve a significant increase in the 2016. A publicly-available version is in evaluated?
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![50742 Federal Register / Vol. 81, No. 148 / Tuesday, August 2, 2016 / Notices
Date of issuance: July 5, 2016. South Carolina Electric and Gas Effective date: As of the date of
Effective date: As of the date of Company and the South Carolina Public issuance and shall be implemented
issuance and shall be implemented Service Authority, Docket Nos. 52–027 within 120 days of issuance.
within 60 days of issuance. and 52–028, Virgil C. Summer Nuclear Amendment Nos.: 181 (Unit 1) and
Amendments Nos.: 308 (Unit 2) and Station (VCSNS), Units 2 and 3, 162 (Unit 2). A publicly-available
312 (Unit 3). A publicly-available Fairfield County, South Carolina version is in ADAMS under Accession
version is in ADAMS under Accession No. ML16165A195; documents related
Date of amendment request: October
No. ML16159A148; documents related to these amendments are listed in the
1, 2015.
to these amendments are listed in the Safety Evaluation enclosed with the
Brief description of amendment: The
Safety Evaluation enclosed with the amendments.
amendments consisted of changes to the
amendments. Renewed Facility Operating License
Facility Combined License, Appendix C,
Renewed Facility Operating License Nos. NPF–68 and NPF–81: Amendments
‘‘Inspections, Tests, Analyses, and
Nos. DPR–44 and DPR–56: The revised the Renewed Facility Operating
Acceptance Criteria [ITAAC].’’
amendments revised the Renewed Licenses and Technical Specifications.
Specifically, the changes to the plant-
Facility Operating Licenses and TSs. Date of initial notice in Federal
Date of initial notice in Federal specific Emergency Planning ITAAC
removed and replaced current Register: March 3, 2015 (80 FR 11480).
Register: December 8, 2015 (80 FR The supplemental letters dated February
76320). The supplemental letter dated references to AP1000 Design Control
Document Table 7.5–1, and Final Safety 27, 2015; May 2, 2016; and June 14,
March 23, 2016, provided additional 2016, provided additional information
information that clarified the Analysis Report (FSAR) Table 7.5–201
on the post-accident monitoring system, that clarified the application, did not
application, did not expand the scope of expand the scope of the application as
the application as originally noticed, with references to proposed updated
FSAR Table 7.5–1 in Table C.3.8–1 for originally noticed, and did not change
and did not change the staff’s original the staff’s original proposed no
proposed no significant hazards ITAAC Numbers C.3.8.01.01.01,
C.3.8.01.05.01.05, and C.3.8.01.05.02.04. significant hazards consideration
consideration determination as determination as published in the
published in the Federal Register. Date of issuance: May 2, 2016.
Effective date: As of the date of Federal Register.
The Commission’s related evaluation The Commission’s related evaluation
of the amendments is contained in a issuance and shall be implemented
within 30 days of issuance. of the amendments is contained in a
Safety Evaluation dated July 5, 2016. Safety Evaluation dated July 15, 2016.
No significant hazards consideration Amendment Nos.: 46. A publicly-
available version is in ADAMS under No significant hazards consideration
comments received: No. comments received: No.
Package Accession No. ML16074A234.
NextEra Energy Duane Arnold, LLC, Documents related to these amendments Dated at Rockville, Maryland, this 22nd
Docket No. 50–331, Duane Arnold are listed in the Safety Evaluation day of July 2016.
Energy Center, Linn County, Iowa enclosed with the amendments. For the Nuclear Regulatory Commission.
Date of amendment request: July 24, Facility Combined License Nos. NPF– Anne T. Boland,
2015. 93 and NPF–94: Amendments revised Director, Division of Operating Reactor
Brief description of amendment: The the Facility Combined Licenses. Licensing, Office of Nuclear Reactor
amendment revised Technical Date of initial notice in Federal Regulation.
Specification 1.4, ‘‘Frequency,’’ by Register: November 24, 2015 (80 FR [FR Doc. 2016–18290 Filed 8–1–16; 8:45 am]
correcting Example 1.4–1 to be 73241). BILLING CODE 7590–01–P
consistent with Technical Specifications The Commission’s related evaluation
Task Force (TSTF) Traveler TSTF–485, of the amendment is contained in a
‘‘Correct Example 1.4–1,’’ Revision 0. In Safety Evaluation dated May 2, 2016. NUCLEAR REGULATORY
addition, the amendment revised No significant hazards consideration COMMISSION
Example 1.4–5 and Example 1.4–6 to be comments received: No.
[NRC–2016–0143]
consistent with Amendment No. 258 to Southern Nuclear Operating Company,
the Renewed Facility Operating License. Inc., Docket Nos. 50–424 and 50–425, Applications and Amendments to
Date of issuance: July 13, 2016. Vogtle Electric Generating Plant, Units 1 Facility Operating Licenses and
Effective date: As of the date of and 2, Burke County, Georgia Combined Licenses Involving
issuance and shall be implemented Proposed No Significant Hazards
within 60 days of issuance. Date of amendment request: July 18,
Considerations and Containing
Amendment No.: 293. A publicly- 2014, as supplemented by letters dated
Sensitive Unclassified Non-Safeguards
available version is in ADAMS under February 27, 2015; May 2, 2016; and
Information and Order Imposing
Accession No. ML15246A408; June 14, 2016.
Procedures for Access to Sensitive
documents related to this amendment Brief description of amendments: The
Unclassified Non-Safeguards
are listed in the Safety Evaluation amendments changed Technical
Information
enclosed with the amendment. Specification 3.9.4, ‘‘Containment
Renewed Facility Operating License Penetrations,’’ to allow containment AGENCY: Nuclear Regulatory
No. DPR–49: The amendment revised penetrations to be un-isolated under Commission.
the Technical Specifications. administrative controls during core ACTION: License amendment request;
Date of initial notice in Federal alterations or movement of irradiated
mstockstill on DSK3G9T082PROD with NOTICES
notice of opportunity to comment,
Register: November 10, 2015 (80 FR fuel assemblies within containment by request a hearing, and petition for leave
69713). adopting a previously NRC-approved to intervene; order imposing
The Commission’s related evaluation Technical Specification Task Force procedures.
of the amendment is contained in a (TSTF) Change Traveler TSTF–312,
Safety Evaluation dated July 13, 2016. Revision 1, ‘‘Administratively Control SUMMARY: The U.S. Nuclear Regulatory
No significant hazards consideration Containment Penetrations.’’ Commission (NRC) received and is
comments received: No. Date of issuance: July 15, 2016. considering approval of four
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Document Created: 2016-08-02 01:47:55
Document Modified: 2016-08-02 01:47:55
| 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 |
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