81 FR 28891 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 81, Issue 90 (May 10, 2016)
Federal Register Volume 81, Issue 90 (May 10, 2016)
| Page Range | 28891-28905 | |
| FR Document | 2016-10949 |
[Federal Register Volume 81, Number 90 (Tuesday, May 10, 2016)] [Notices] [Pages 28891-28905] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2016-10949] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2016-0093] Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission. ACTION: Biweekly notice. ----------------------------------------------------------------------- SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from April 12 to April 25, 2016. The last biweekly notice was published on April 26, 2016 (81 FR 24659). DATES: Comments must be filed by June 9, 2016. A request for a hearing must be filed by July 11, 2016. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0093. 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: Shirley Rohrer, Office of Nuclear [[Page 28892]] Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-5411, email: [email protected]. I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0093 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-2016-0093. 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 the SUPPLEMENTARY INFORMATION section. 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-2016-0093, facility name, unit number(s), application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (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 should 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. Should the Commission take 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. Should the Commission make 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 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 [[Page 28893]] 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. 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 NRC 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 July 11, 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 Sec. 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. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by July 11, 2016. 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). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least ten 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,'' which is available on the agency's public Web site at http://www.nrc.gov/site-help/e-submittals.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 Meta System 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. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. 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/e-submittals.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 [[Page 28894]] 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 Meta System 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 Meta System Help Desk is available between 8 a.m. and 8 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 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 request 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 accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power Station, Unit No. 1 (MPS1), New London County, Connecticut Date of amendment request: March 28, 2014. A publicly-available version is in the ADAMS under Accession No. ML14093A028. Description of amendment request: The amendment would make changes to the MPS1 Permanently Defueled Technical Specifications (PDTSs) by deleting the Table of Contents section and making administrative changes to the PDTSs. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1) Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes are administrative in nature. The proposed changes remove the PDTS Table of Contents section and make two other administrative changes to the PDTSs. Furthermore, MPS1 has permanently ceased operation and is being maintained in a defueled condition. Therefore, the only credible design basis accident is a fuel handling accident. The administrative changes proposed herein are not initiators of any fuel handling accident previously evaluated, and, consequently, the probability and consequences of a fuel handling accident previously evaluated is not significantly increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2) Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are administrative in nature, therefore no new or different accidents result from the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The changes do not alter assumptions made in the safety analysis. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3) Do the proposed changes involve a significant reduction in the margin of safety? Response: No. The proposed administrative changes do not involve a change in the method of plant operation, do not affect any accident analyses, and do not relax any safety system settings. 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: Lillian M. Cuoco, Senior Counsel, Dominion Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Bruce A. Watson. Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: February 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16076A413. Description of amendment request: The amendment would allow a one- time extension to the 10-year frequency of the McGuire Nuclear Station, Units 1 [[Page 28895]] and 2, containment leakage rate tests. The change would extend the period from 10 years to 10.5 years for each 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment to the Technical Specifications (TS) involves the extension of the McGuire Nuclear Station (MNS) Type A containment integrated leak rate test interval to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to no longer than 10.5 years from the last Type A test. This extension is bounded by the 15 month extension, permissible only for non-routine emergent conditions, allowed in accordance with NEI [Nuclear Energy Institute] 94-01 revision 0. The proposed extension also does not change the test method or procedure. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from 10 years to 10.5 years, measured, as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.023 person-rem/year. EPRI [Electric Power Research Institute] Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of <= 1.0 person-rem per year, or <= 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT [integrated leak rate test] intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493, Performance-Based Containment Leak- Test Program, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The MNS Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and; (2) time based as previously discussed. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME Section XI, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the MNS Type A containment integrated leak rate test interval from 10 years to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to 10.5 years from the last Type A test. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. The proposed amendment to TS 5.5.2 involves the extension of the MNS Type A containment integrated leak rate test interval to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to no longer than 10.5 years from the last Type A test. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests for MNS. The proposed surveillance interval extension is bounded by the 15-year ILRT interval currently authorized within NEI 94-01, Revisions 2-A and 3- A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the approval of this proposed change, since these are not affected by changes to the Type A test intervals. 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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 28202. NRC Branch Chief: Michael T. Markley. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No. 1, DeWitt County, Illinois Date of amendment request: January 25, 2016, as supplemented by letter dated March 31, 2016. A publicly-available version is in ADAMS under Accession Nos. ML16025A182 and ML16076A077. Description of amendment request: The proposed amendment would revise the technical specifications (TSs) to allow a permanent extension of the Type ``A'' integrated leak rate testing and Type ``C'' leak rate testing frequencies. This request also proposes to delete information in TS 5.5.13 regarding a completed requirement to perform Type ``C'' testing in 2008. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed activity involves the extension of the Clinton Power Station (CPS), Unit 1, Type A containment test interval to [[Page 28896]] 15 years, and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A Integrated Leak Rate Test (ILRT) interval from three-per-ten years to once-per- fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI [Electric Power Research Institute] guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/ yr or less than 1% person-rem/yr. The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance for change in population dose of less than 1.0 person- rem/yr or less than 1% person-rem/yr. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. In addition, as documented in NUREG-1493, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CPS, Unit 1 Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and, (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Section XI, and Technical Specifications (TS) requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS 5.5.13, ``Primary Containment Leakage Rate Testing Program,'' involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment. The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.13 involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leaktightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves the extension of the interval between Type A containment leak rate tests and Type C tests for CPS, Unit 1. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI [Nuclear Energy Institute] 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals. The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for CPS, Unit 1. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how CPS is operated and maintained. Thus, there is no reduction in any margin of safety. 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: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555. NRC Acting Branch Chief: Justin C. Poole. Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16054A359. Description of amendment request: The amendment would revise the Technical Specifications to incorporate previously NRC-approved Industry/Technical Specification Task Force 439 (TSTF-439), Revision 2, ``Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO.'' 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 [[Page 28897]] 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 eliminates certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Time are no different than the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed change does not alter or prevent the ability of SSCs [systems, structures, and components] from performing their intended 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 an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change does not alter any assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of anew or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to delete the second Completion Time 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 of 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station (NMPNS), Units 1 and 2, Oswego County, New York Date of amendment request: March 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16078A065. Description of amendment request: The amendment would revise the Technical Specifications (TS) concerning a change to the method of calculating core reactivity for the purpose of performing the Reactivity Anomalies surveillance at NMPNS, Units 1 and 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed TS changes do not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the Reactivity Anomalies surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the TS requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of k eff . Therefore, since the Reactivity Anomalies surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. This TS amendment request does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the Reactivity Anomalies surveillance will only provide a new, more efficient method of detecting an unexpected change in core reactivity. Since all systems continue to be operated within their design bases, no new failure modes are introduced and the possibility of a new or different kind of accident is not created. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. This proposed TS amendment proposes to change the method for performing the Reactivity Anomalies surveillance from a comparison of predicted to actual control rod density to a comparison of predicted to monitored keff . The direct comparison of keff provides a technically superior method of calculating any differences in the expected core reactivity. The Reactivity Anomalies surveillance will continue to be performed at the same frequency as is currently required by the TS, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16055A149. Description of amendment request: The amendment would (1) revise Technical Specification (TS) 4.2.1, ``Reactor Core, Fuel Assemblies,'' to add Optimized ZIRLO\TM\, as an approved fuel rod cladding material, (2) revise TS 5.6.5.b to add the Westinghouse topical reports for Optimized ZIRLO\TM\ and ZIRLO[supreg], and (3) revise TS 5.6.5.b with a non-technical change to the Reference 11 title (replace a semicolon with a period). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: [[Page 28898]] EGC [Exelon Generation Company] has evaluated the proposed changes for Braidwood and Byron, using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration. Criteria 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change would allow the use of Optimized ZIRLO\TM\ clad nuclear fuel in the reactors. The NRC approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, ``Optimized ZIRLO\TM\ prepared by Westinghouse Electric Company LLC (Westinghouse), addresses Optimized ZIRLO\TM\ and demonstrates that Optimized ZIRLO\TM\ has essentially the same properties as currently licensed ZIRLO[supreg]. The fuel cladding itself is not an accident initiator and does not affect accident probability. With the approved exemption, use of Optimized ZIRLO\TM\ fuel cladding will continue to meet all 10 CFR 50.46 acceptance criteria and, therefore, will 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 change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Use of Optimized ZIRLO\TM\ clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of Optimized ZIRLO\TM\ are similar to those of standard ZIRLO[supreg]. Therefore, Optimized ZIRLO\TM\ fuel rod cladding will perform similarly to those fabricated from standard ZIRLO[supreg] thus precluding the possibility of the fuel cladding becoming an accident initiator and causing a new or different type of accident. 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 will not involve a significant reduction in the margin of safety. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of the Optimized ZIRLO\TM\ are not significantly different from those of standard ZIRLO[supreg]. Optimized ZIRLO\TM\ is expected to perform similarly to standard ZIRLO[supreg] for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference is Optimized ZIRLO\TM\ material properties relative to standard ZIRLO[supreg] could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO\TM\ properties will demonstrate that the acceptance criteria of 10 CFR 50.46 have been satisfied. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, EGC concludes that the proposed amendment to allow the use of Optimized ZIRLO\TM\ fuel cladding material does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant hazards consideration'' is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555. NRC Acting Branch Chief: Justin C. Poole. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: March 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16075A411. Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.6.2.2, ``Suppression Pool Water Level,'' as well as TS surveillance requirements 3.6.2.4.1 and 3.6.2.4.4 associated with TS 3.6.2.4, ``Suppression Pool Makeup System (SPMU),'' to allow installation of the reactor well to steam dryer storage pool gate in the upper containment pool (UCP) in MODES 1, 2, and 3. The proposed amendment would also create new special operations TS 3.10.9, ``Suppression Pool Makeup--MODE 3 Upper Containment Pool Drain-Down,'' to allow draining of the reactor well portion of the UCP in MODE 3. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The changes proposed in the license amendment request specify different water level requirements in the upper containment pool and suppression pool to permit gate installation in MODES 1, 2, and 3, and drain-down of the reactor well in MODE 3. The probability of an accident previously evaluated is unrelated to the water level in these pools, since they are mitigating systems. The operation or failure of a mitigating system does not contribute to the occurrence of an accident. No active or passive failure mechanisms that could lead to an accident are affected by these proposed changes. Suppression pool water levels are increased during upper pool gate installation in MODES 1, 2, and 3 and during reactor well drain-down in MODE 3, with a potential for an increased probability of drywell flooding during an inadvertent dump of the upper containment pool. An inadvertent dump of the upper pool during any period of operation with a pressurized vessel does not represent, in and of itself, any significant hazard to the public, the plant operating personnel, or any plant equipment. The piping components which would be affected in this event have been analyzed for the flooding effect, and it has been determined that this event could not initiate a loss of coolant accident (LOCA). The changes have no impact on the ability of any of the emergency core cooling systems (ECCS) to function adequately, since adequate net positive suction head (NPSH) is maintained. The increase in suppression pool water level to compensate for the reduction in UCP volume will provide reasonable assurance that the minimum post-accident vent coverage is adequate to assure the pressure suppression function of the suppression pool is accomplished. The suppression pool water level will be raised above the current high water level for the proposed reactor well drain- down activity only after the reactor pressure has been reduced sufficiently to assure that the hydrodynamic loads from a loss of coolant accident will not exceed the design values. The reduced reactor pressure will also ensure that the loads due to main steam safety relief valve actuation with an elevated pool level are within the design loads. Relative to dose rates on the refuel floor, the resultant dose rates from the reactor in MODES 3 and 4 are the same regardless of a drain-down of the upper pool reactor well. Relative to a low pressure LOCA in MODE 3, the reduced post-LOCA containment pressure and the decay time to reach MODE 3 conditions ensures that post- accident dose consequences are bounded by the design-basis accident LOCA. Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from an accident previously evaluated? Response: No. The proposed changes specify different water level requirements in the upper containment pool and suppression pool to permit gate installation in MODES 1, 2, and 3, and drain-down of the reactor well in MODE 3. These changes do not affect or alter the ability of the suppression pool makeup [[Page 28899]] (SPMU) system to perform its design function. The proposed change in the pool water levels will maintain the design function of mitigating the pressure and temperature increase generated by a LOCA, and will maintain the required drywell vent coverage during post-accident ECCS draw down. The altered water levels in the pools do not create a different type of accident than presently evaluated. With the reduced pressure in the reactor coolant system, the GOTHIC computer program simulations demonstrate that the accident responses at defined conditions with the reactor well drained in MODE 3 are bounded by the current design basis accidents. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes to the UCP and the suppression pool water levels do not introduce any new setpoints at which protective or mitigating actions are initiated. Current instrument setpoints remain unaltered by this change. Although the water levels are adjusted for the UCP gate installation and the reactor well drain- down activity, the design and functioning of the containment pressure suppression system remains unchanged. The proposed total water volume is sufficient to provide high confidence that the pressure suppression and containment systems will be capable of mitigating large and small break accidents. All analyzed accident results remain within the design values for the structures and equipment. 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: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: David J. Wrona. Pacific Gas and Electric Company (PG&E), 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: March 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16084A588. Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.4.12, ``Low Temperature Overpressure Protection (LTOP) System,'' to reflect the mass input transient analysis that assumes an emergency core cooling system (ECCS) centrifugal charging pump (CCP) and the normal charging pump (NCP) capable of simultaneously injecting into the reactor coolant system (RCS) during TS 3.4.12 applicability. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS 3.4.12 to allow an ECCS CCP and the NCP aligned to LTOP orifice to be capable of injecting into the RCS during low RCS pressures and temperatures. The LCO [Limiting Condition for Operation] provides RCS overpressure protection by having a minimum coolant input capability and have adequate pressure relief capability. Analyses have demonstrated that one power operated relief valve (PORV) or an RCS vent of at least 2.07 square inches is capable of limiting the RCS pressure excursions below the 10 CFR 50, Appendix G limits for the design basis LTOP limits. 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 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. The NRC has previously evaluated a similar LAR [license amendment request] related to Wolf Creek Generating Station. In Amendment No. 207, the NRC concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated [ADAMS Accession No. ML13282A534]. In 2007, PG&E replaced the Unit 1 non-safety-related PDP [positive displacement pump] with a non-safety-related CCP, called the NCP, in order to alleviate operational issues associated with the PDP. In 2008, PG&E performed the replacement on Unit 2. PG&E also designed, tested, and installed an FCO [flow choking orifice] called the LTOP orifice to be used during LTOP operation to ensure that the total maximum mass injection capability with the NCP remained bounded by the LTOP mass injection analysis. These changes were implemented under 10 CFR 50.59. However, no physical changes are being made to the plant as a result of the proposed license amendment. 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 change revises TS 3.4.12 to allow an ECCS CCP and the NCP aligned to LTOP orifice to be capable of simultaneously injecting into the RCS during low RCS pressures and temperatures. The LCO provides RCS overpressure protection by having a minimum coolant input capability and have adequate pressure relief capability. Analyses have demonstrated that one PORV or an RCS vent of at least 2.07 square inches is capable of limiting the RCS pressure excursions below the 10 CFR 50, Appendix G limits for the design basis LTOP limits. The proposed change will not physically alter the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. The proposed change does not introduce new accident initiators or impact assumptions made in the safety analysis. Testing requirements continue to demonstrate that the LCOs are met and the system components are functional. Therefore, the proposed change 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 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 impacted 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 requests involve 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. [[Page 28900]] South Carolina Electric and Gas Company, 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: March 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16067A145. Description of amendment request: The proposed changes, if approved, would amend Combined License (COL) No. NPF-93 and NPR-94 for the VCSNS. The requested amendment proposed changes would depart from the approved AP1000 Design Control Document (DCD) ``Tier 2'' and ``Tier 2*'' information as currently incorporated into the VCSNS Updated Final Safety Analysis Report (UFSAR). The changes relate to updating the UFSAR text and tables; and information incorporated by reference related to Westinghouse Electric Company's Reports WCAP-16096, ``Software Program Manual for Common QTM Systems,'' (also known as the Common Q SPM) Revision 4, WCAP-16097, ``Common Qualified Platform Topical Report,'' (also known as the Common Q Topical Report) Revision 3, and WCAP-15927, ``Design Process for AP1000 Common Q Safety Systems,'' Revision 4; and associated documents and references such as a reference to the NRC's Regulatory Guide 1.152, ``Criteria for Use of Computers in Safety Systems of Nuclear Power Plants'' (Revision 3, July 2011), and its associated exceptions. The proposed changes also include removal of Tier 2* WCAP-17201-P, ``AC160 High Speed Link Communication Compliance to DI&C-ISG-04 Staff Positions 9, 12, 13 and 15 Technical Report,'' as a UFSAR incorporated by reference document. 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. WCAP-16096 (Common Q Software Program Manual) was updated to Revision 4 to reference later NRC endorsed regulatory guides and standards and update the requirements for the software design and development processes for the Common Q portion of the AP1000 Protection and Safety Monitoring System (PMS). WCAP-16097 (Common Q Topical Report) was updated to Revision 3 to describe new Common Q components and standards currently used for the AP1000 PMS implementation of the Common Q platform. These two WCAPs have been reviewed and approved by the NRC in Safety Evaluations dated February 7, 2013. WCAP-15927 was updated to reference the newest revisions of WCAP-16096 and WCAP-16097 and for editorial corrections. The proposed activity adopts the updated versions as incorporated by reference documents into the Updated Final Safety Analysis Report. Other proposed document changes support the implementation of the updated versions of WCAP-16096, WCAP-16097, and WCAP-15927. The Common Q platform is an acceptable platform for nuclear safety-related applications. The Common Q system meets the requirements of 10 CFR part 50, Appendix A, General Design Criteria (Criteria 1, 2, 4, 13, 19, 20, 21, 22, 23, 24, and 25), the Institute of Electrical and Electronics Engineers (IEEE) Standard 603-1991 for the design of safety-related reactor protection systems, engineered safety features systems and other plant systems, and the guidelines of Regulatory Guide 1.152 and supporting industry standards for the design of digital systems. Because the Common Q platform and the Protection and Safety Monitoring System (PMS) implementation of the Common Q platform meet the criteria in the applicable General Design Criteria, the revisions to these documents do not affect the prevention and mitigation of abnormal events, such as accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses as described in the licensing basis. The incorporation of the updated documents does not adversely affect the interface with any structure, system, or component (SSC) accident initiator or initiating sequence of events. Thus, the probabilities of the accidents previously evaluated in the UFSAR are not affected. Therefore, the proposed activity does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to adopt the updated WCAP-16096, WCAP- 16097, and WCAP-15927 into the UFSAR do not adversely affect the design or operation of safety-related equipment or equipment whose failure could initiate an accident beyond what is already described in the licensing basis. These changes do not adversely affect fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change. Therefore, this activity 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 proposed changes to adopt the updated WCAP-16096, WCAP- 16097, and WCAP-15927 into the UFSAR do not adversely affect the design, construction, or operation of any plant SSCs, including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514. NRC Acting Branch Chief: John McKirgan. South Carolina Electric and Gas Company, 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: March 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16075A264. Description of amendment request: The proposed change would amend the Combined License (COL) No. NPF-93 and NPF-94 for the VCSNS. The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information (text, tables, and figures) and involved Tier 2* information (as incorporated into the Updated Final Safety Analysis Report as plant specific DCD information), and also involves a change to the plant-specific Technical Specifications. Specifically, the amendment request proposes changes to the plant- specific AP1000 fuel system design, nuclear design, thermal hydraulic design, and accident analyses as described in the licensing basis documents. These proposed changes are consistent with those generically approved in WCAP-17524-P-A, Revision 1, ``AP1000 Core Reference Report.'' 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? [[Page 28901]] Response: No. The proposed changes will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses. The UFSAR [Updated Final Safety Analysis Report] Chapter 15 accident analyses describe the analyses of various design basis transients and accidents to demonstrate compliance of the AP1000 design with the acceptance criteria for these events. The acceptance criteria for the various events are based on meeting the relevant regulations, general design criteria, the Standard Review Plan, and are a function of the anticipated frequency of occurrence of the event and potential radiological consequences to the public. As such, each design-basis event is categorized accordingly based on these considerations. As discussed in Section 5.3 of WCAP-17524-P-A Revision 1, the revised accident analyses maintain their plant conditions, and thus their frequency designation and consequence level as previously evaluated. As confirmed in the Safety Evaluation Report (SER), the revised analyses meet the applicable guidelines in the Standard Review Plan. 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 will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses. The proposed changes would not introduce a new failure mode, fault, or sequence of events that could result in a radioactive material release. The proposed changes do not alter the design, configuration, or method of operation of the plant beyond standard functional capabilities of the equipment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses. Safety margins are applied at many levels to the design and licensing basis functions and to the controlling values of parameters to account for various uncertainties and to avoid exceeding regulatory or licensing limits. UFSAR Subsection 4.1.1 presents the Principle Design Requirements imposed on the fuel and control rod mechanism design to ensure that the performance and safety criteria described in UFSAR Chapter 4 and Chapter 15 are met. The revised fuel system design, nuclear design, thermal hydraulic design, and accident analyses maintain the same Principle Design Requirements, and further, satisfy the applicable regulations, general design criteria, and Standard Review Plan. The effects of the changes do not result in a significant reduction in margin for any safety function, and were evaluated in the Safety Evaluation Report for WCAP-17524-P-A Revision 1 and found to be acceptable. 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: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514. NRC Acting Branch Chief: John McKirgan. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16054A585. Description of amendment request: The amendment would revise the WBN Dual Unit Fire Protection Report and would revise the associated License Condition regarding the WBN fire protection program. Specifically, the amendment requests approval of a deviation from the physical separation requirements of 10 CFR part 50, appendix R, section III.G.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. A fire hazards analysis was performed for the areas under the scope of this amendment. This fire hazards analysis demonstrates that one train of safe shutdown equipment will remain functional in the event of an Appendix R fire, even though a radiant energy shield will not be provided for two raceway containing safe shutdown circuits. 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. A fire hazards analysis was performed for the areas under the scope of this amendment. This fire hazards analysis demonstrates that one train of safe shutdown equipment will remain functional in the event of an Appendix R fire, even though a radiant energy shield will not be provided for two raceway containing safe shutdown circuits. Based on this, the proposed amendment will not alter the requirements or function for systems required during accident conditions. 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. A fire hazards analysis was performed for the areas under the scope of this amendment. This fire hazards analysis demonstrates that one train of safe shutdown equipment will remain functional in the event of an Appendix R fire, even though a radiant energy shield will not be provided for two raceway containing safe shutdown circuits. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Sherry A. Quirk, Executive Vice President and General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, Knoxville, TN 37902. NRC Branch Chief: Benjamin G. Beasley. 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. [[Page 28902]] Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of amendment request: March 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16064A488. Brief description of amendment request: The amendment would revise the Cyber Security Plan implementation schedule for Milestone 8 and would revise the associated license condition in the Facility Operating License. Date of publication of individual notice in Federal Register: April 19, 2016 (81 FR 23011). Expiration date of individual notice: May 19, 2016 (public comments); June 20, 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. DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan Date of amendment request: September 24, 2015. Brief description of amendment: The amendment revises Surveillance Requirements (SRs) to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification. The changes address the concerns discussed in NRC Generic Letter (GL) 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'' as described in NRC-approved Technical Specifications Task Force (TSTF)-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.'' Date of issuance: April 20, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 204. A publicly-available version is in ADAMS under Accession. No. ML16069A006; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-43: This amendment revises the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: January 5, 2016 (81 FR 260). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 20, 2016. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: April 30, 2015, as supplemented by letter dated February 19, 2016. Brief description of amendments: The amendments approved adoption of an emergency action level scheme based on Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors,'' for the Catawba Nuclear Station, Units 1 and 2. Date of issuance: April 18, 2016. Effective date: As of the date of issuance and shall be implemented by March 10, 2017. Amendment Nos.: 279 for Unit 1 and 275 for Unit 2. A publicly- available version is in ADAMS under Accession No. ML16082A038; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-35 and NPF-52: The amendments revised the Renewed Facility Operating License. Date of initial notice in Federal Register: June 23, 2015 (80 FR 35980). The supplemental letter dated February 19, 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 April 18, 2016. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-369, 50-370, 50-413, and 50- 414, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina and Catawba Nuclear Station, Units 1 and 2, York County, SC Date of amendment request: June 23, 2015. Brief description of amendments: The amendments remove superseded TS requirements. Date of issuance: April 8, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 283, 262, 278, and 274. A publicly-available version is in ADAMS under Accession No. ML16060A229; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-9, NPF-17, NPF-35, and NPF-52: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: August 4, 2015 (80 FR 46347). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 8, 2016. No significant hazards consideration comments received: No. 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: April 30, 2015, as supplemented by letters dated November 19, 2015, and January 28, 2016. Brief description of amendment: The amendment adopted the NRC- endorsed [[Page 28903]] Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the Development of Emergency Action Levels for Non-Passive Reactors.'' Date of issuance: April 13, 2016. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment No.: 149. A publicly-available version is in ADAMS under Accession No. ML16057A838; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment. Facility Operating License No. NPF-63: The amendment revised the Emergency Action Level Technical Bases document. Date of initial notice in Federal Register: July 21, 2015 (80 FR 43128). The supplemental letters dated November 19, 2015, and January 28, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in an SE dated April 13, 2016. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50- 286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester County, New York Date of amendment request: June 16, 2015. Brief description of amendments: The amendments revised the Cyber Security Plan Milestone 8 full implementation date by extending the full implementation date from June 30, 2016, to December 31, 2017. Date of issuance: April 12, 2016. Effective date: As of the date of issuance, and shall be implemented within 30 days of issuance. Amendment Nos.: 59 (Unit No. 1), 284 (Unit No. 2), and 260 (Unit No. 3). A publicly-available version is in ADAMS under Accession No. ML16064A215; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Provisional Operating License No. DPR-5 and Facility Operating License Nos. DPR-26 and DPR-64: The amendments revised the Provisional Operating License for Unit No. 1 and the Facility Operating Licenses for Unit Nos. 2 and 3. Date of initial notice in Federal Register: August 4, 2015 (80 FR 46348). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 12, 2016. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland Date of amendment request: November 5, 2015. Brief description of amendments: The amendments revise the Surveillance Requirement (SR) frequencies for SRs 3.4.6.4, 3.4.7.4, 3.4.8.3, 3.5.2.10, 3.6.6.9, 3.9.4.2, and 3.9.5.4. The changes to the SR frequencies relocate the frequencies to the Surveillance Frequency Control Program. Date of issuance: April 11, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 317 and 295. A publicly-available version is in ADAMS under Accession No. ML16060A401; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: January 5, 2016 (81 FR 261). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 11, 2016. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of amendment request: March 23, 2015, as supplemented by letters dated January 8, 2016, and March 21, 2016. Brief description of amendment: The amendment revised the technical specifications (TS) and relocated the secondary containment bypass leakage paths table from the TS to the Technical Requirements Manual. Date of issuance: April 19, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment No.: 156. A publicly-available version is in ADAMS under Accession No. ML16088A053; documents related to this amendment is listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: September 29, 2015 (80 FR 58517). The supplemental letters dated January 8, 2016, and March 21, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 2016. No significant hazards consideration comments received: No. Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2 (PSL-2), St. Lucie County, Florida Date of amendment request: December 30, 2014, as supplemented by letters dated March 23, June 2, June 18, July 30, October 2, November 3, 2015; and December 8, 2015. Brief description of amendment: The amendment revised the Technical Specifications (TSs) to allow the use of AREVA fuel and AREVA M5[supreg] material as an approved fuel rod cladding at PSL-2. Date of issuance: April 19, 2016. Effective date: As of the date of issuance and shall be implemented upon the start of the PSL-2 Cycle 23 spring 2017 refueling outage to support the AREVA fuel transition project plan. Amendment No.: 182. A publicly-available version is in ADAMS under Accession No. ML16063A121; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-16: Amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: June 9, 2015 (80 FR 32620). The supplements dated June 2, June 18, July 30, October 2, November 3, and December 8, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 2016. No significant hazards consideration comments received: No. [[Page 28904]] Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of application for amendments: June 26, 2013, as supplemented by letters dated September 29, October 27, October 29, November 26, and December 31, 2014; February 25 (two letters), May 7, October 15, and December 31, 2015; and January 28, 2016. Brief description of amendments: The amendments permit the PG&E (the licensee) to adopt a new fire protection licensing basis based on National Fire Protection Association (NFPA) Standard 805, ``Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition),'' at Diablo Canyon Power Plant, Units 1 and 2, that complies with the requirements of 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide 1.205, ``Risk Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,'' December 2009. Date of issuance: April 14, 2016. Effective date: As of its date of issuance and shall be implemented as described in the transition license conditions. Amendment Nos.: Unit 1--225; Unit 2--227. A publicly-available version is in ADAMS under Accession No. ML16035A441; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: December 26, 2013 (78 FR 78408). The supplemental letters dated October 3, 2013; September 29, October 27, October 29, November 26, and December 31, 2014; February 25 (two letters), May 7, October 15, and December 31, 2015; and January 28, 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 April 14, 2016. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: September 1, 2015. Brief description of amendment: The amendment authorized changes to the VEGP Units 3 and 4 plant specific emergency planning inspections, tests, analyses, and acceptance criteria (ITAAC) in Appendix C of VEGP Units 3 and 4 Combined Operating Licenses (COLs). The changes authorize the removal of the copy of Updated Final Safety Analysis Report Table 7.5-1, ``Post-Accident Monitoring System'' from ITAAC in Appendix C of the VEGP Units 3 and 4 COLs. Date of issuance: March 30, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 47. A publicly-available version is in ADAMS under Accession No. ML16061A220; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses. Date of initial notice in Federal Register: October 27, 2015 (80 FR 65807). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2015. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: January 13, 2015, as supplemented by letters dated June 16 and November 24, 2015. Brief description of amendments: The amendments adopt Technical Specification Task Force change number 523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation,'' for the Hatch Nuclear Plant, Unit Nos 1 and 2, technical specifications. The change revised or added surveillance requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification. Date of issuance: April 14, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 278 and 222. A publicly-available version is in ADAMS under Accession No. ML16090A174; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: March 17, 2015 (80 FR 13911). The supplemental letters dated June 16 and November 24, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 14, 2016. No significant hazards consideration comments received: No. Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: September 23, 2015. Brief description of amendment: The amendment revised the diesel generator (DG) full load rejection test and endurance and margin test specified by Technical Specification (TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' Surveillance Requirements (SR) 3.8.1.10 and 3.8.1.14, respectively. The change adds a new Note to SR 3.8.1.10 and SR 3.8.1.14, consistent with Technical Specification Task Force (TSTF) traveler TSTF-276-A, Revision 2, ``Revise DG full load rejection test.'' The Note allows the full load rejection test and endurance and margin test to be performed at the specified power factor with clarifications addressing situations when the power factor cannot be achieved. Date of issuance: April 15, 2016. Effective date: As of its date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 215. A publicly-available version is in ADAMS under Accession No. ML16081A194; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-42. The amendment revised the Operating License and Technical Specifications. Date of initial notice in Federal Register: November 24, 2015 (80 FR 73242). [[Page 28905]] The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 15, 2016. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 2nd day of May 2016. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2016-10949 Filed 5-9-16; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices 28891
Applied Sciences Advisory Committee (NASA) hereby gives notice of its intent NUCLEAR REGULATORY
(ASAC). This Committee functions in an to grant an exclusive license in the COMMISSION
advisory capacity to the Director, Earth United State to practice the inventions
[NRC–2016–0093]
Science Division, in the NASA Science described and claimed in U.S. Patent
Mission Directorate. The meeting will Application Number 14/658,584, titled Applications and Amendments to
be held for the purpose of soliciting, ‘‘Infrasonic Stethoscope for Monitoring Facility Operating Licenses and
from the applied sciences community Physiological Processes,’’ NASA Case Combined Licenses Involving No
and other persons, scientific and Number LAR–18509–1, to Infrasonix, Significant Hazards Considerations
technical information relevant to Inc., having its principal place of
program planning. AGENCY: Nuclear Regulatory
business in Lawrenceville, GA. Certain
DATES: Tuesday, May 31, 2016, 12:00 Commission.
patent rights in this invention have been
p.m. to 3:00 p.m., Eastern Daylight Time assigned to the United States of ACTION: Biweekly notice.
(EDT). America, as represented by the SUMMARY: Pursuant to Section 189a. (2)
FOR FURTHER INFORMATION CONTACT: Ms. Administrator of the National of the Atomic Energy Act of 1954, as
Ann Delo, Science Mission Directorate, Aeronautics and Space Administration. amended (the Act), the U.S. Nuclear
NASA Headquarters, Washington, DC The prospective exclusive license will Regulatory Commission (NRC) is
20546, (202) 358–0750, fax (202) 358– comply with the terms and conditions publishing this regular biweekly notice.
2779, or ann.b.delo@nasa.gov. of 35 U.S.C. 209 and 27 CFR 404.7. The Act requires the Commission to
SUPPLEMENTARY INFORMATION: This publish notice of any amendments
meeting will be open to the public DATES: The prospective exclusive
issued, or proposed to be issued, and
telephonically and by WebEx. You must license may be granted unless, within
grants the Commission the authority to
use a touch-tone phone to participate in fifteen (15) days from the date of this
issue and make immediately effective
this meeting. Any interested person may published notice, NASA receives
any amendment to an operating license
dial the USA toll free conference call written objections including evidence or combined license, as applicable,
number (888) 469–2034, passcode and argument that establish that the upon a determination by the
1671423, followed by the # sign, to grant of the license would not be Commission that such amendment
participate in this meeting by telephone. consistent with the requirements of 35 involves no significant hazards
The WebEx link is https:// U.S.C. 209 and 37 CFR. 404.7. consideration, notwithstanding the
nasa.webex.com/; the meeting number Competing applications completed and pendency before the Commission of a
is 997 185 050 and the password is @ received by NASA within fifteen (15) request for a hearing from any person.
May31st. days of the date of this published notice This biweekly notice includes all
The agenda for the meeting includes will also be treated as objections to the notices of amendments issued, or
the following topics: grant of the contemplated exclusive proposed to be issued from April 12 to
• Overview of 2016 Applied Sciences license. April 25, 2016. The last biweekly notice
Program Budget Objections submitted in response to was published on April 26, 2016 (81 FR
• Continuity Study this notice will not be made available to 24659).
• Status of User Working Groups and the public for inspection and, to the DATES: Comments must be filed by June
Science Teams extent permitted by law, will not be 9, 2016. A request for a hearing must be
• Update on Status of Decadal Survey released under the Freedom of filed by July 11, 2016.
It is imperative that the meeting be Information Act, 5 U.S.C. 552. ADDRESSES: You may submit comments
held on this date to accommodate the by any of the following methods (unless
scheduling priorities of the key ADDRESSES: Objections relating to the this document describes a different
participants. prospective license may be submitted to method for submitting comments on a
Patent Counsel, Office of Chief Counsel, specific subject):
Patricia D. Rausch,
MS 30, NASA Langley Research Center, • Federal Rulemaking Web site: Go to
Advisory Committee Management Officer, Hampton, Virginia 23681, (757) 864–
National Aeronautics and Space
http://www.regulations.gov and search
Administration. 3221 (phone), (757) 864–9190 (fax). for Docket ID NRC–2016–0093. Address
questions about NRC dockets to Carol
[FR Doc. 2016–10842 Filed 5–9–16; 8:45 am] FOR FURTHER INFORMATION CONTACT:
Gallagher; telephone: 301–415–3463;
BILLING CODE 7510–13–P Andrea Z. Warmbier, Patent Attorney,
email: Carol.Gallagher@nrc.gov. For
Office of Chief Counsel, MS 30, NASA
technical questions, contact the
Langley Research Center, Hampton, individual listed in the FOR FURTHER
NATIONAL AERONAUTICS AND Virginia 23681, (757) 864–3221 (phone);
SPACE ADMINISTRATION INFORMATION CONTACT section of this
(757) 864–9190 (fax); document.
[Notice (16–035)] Andrea.Z.Warmbier@nasa.gov. • Mail comments to: Cindy Bladey,
Information about other NASA Office of Administration, Mail Stop:
Notice of Intent To Grant an Exclusive inventions available for licensing can be OWFN–12–H08, U.S. Nuclear
License found online at http:// Regulatory Commission, Washington,
National Aeronautics and technology.nasa.gov. DC 20555–0001.
asabaliauskas on DSK3SPTVN1PROD with NOTICES
AGENCY:
Space Administration For additional direction on obtaining
Mark P. Dvorscak,
information and submitting comments,
ACTION: Notice of intent to grant Agency Counsel for Intellectual Property. see ‘‘Obtaining Information and
exclusive license. [FR Doc. 2016–10929 Filed 5–9–16; 8:45 am] Submitting Comments’’ in the
SUMMARY: This notice is issued in BILLING CODE 7510–13–P SUPPLEMENTARY INFORMATION section of
accordance with 35 U.S.C. 209(e) and 37 this document.
CFR 404.7(a)(1)(i). The National FOR FURTHER INFORMATION CONTACT:
Aeronautics and Space Administration Shirley Rohrer, Office of Nuclear
VerDate Sep<11>2014 17:33 May 09, 2016 Jkt 238001 PO 00000 Frm 00074 Fmt 4703 Sfmt 4703 E:\FR\FM\10MYN1.SGM 10MYN1](https://thefederalregister.org/doc/81_FR_28981/page/1.svg)
![Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices 28895
and 2, containment leakage rate tests. procedural requirements for system testing detects a large percentage of
The change would extend the period restoration ensure that containment integrity containment leakage paths and that the
from 10 years to 10.5 years for each unit. is not degraded by plant modifications or percentage of containment leakage paths that
Basis for proposed no significant maintenance activities. The design and are detected only by Type A testing is small.
construction requirements of the The containment inspections performed in
hazards consideration determination: containment combined with the containment accordance with ASME Section XI, and TS
As required by 10 CFR 50.91(a), the inspections performed in accordance with serve to provide a high degree of assurance
licensee has provided its analysis of the ASME Section XI, the Maintenance Rule, and that the containment would not degrade in a
issue of no significant hazards TS requirements serve to provide a high manner that is detectable only by Type A
consideration, which is presented degree of assurance that the containment testing. The combination of these factors
below: would not degrade in a manner that is ensures that the margin of safety in the plant
detectable only by a Type A test. Based on safety analysis is maintained. The design,
1. Does the proposed amendment involve the above, the proposed extensions do not operation, testing methods and acceptance
a significant increase in the probability or significantly increase the consequences of an criteria for Type A, B, and C containment
consequences of an accident previously accident previously evaluated. leakage tests specified in applicable codes
evaluated? Therefore, the proposed change does not and standards would continue to be met,
Response: No. result in a significant increase in the with the approval of this proposed change,
The proposed amendment to the Technical probability or consequences of an accident since these are not affected by changes to the
Specifications (TS) involves the extension of previously evaluated.
the McGuire Nuclear Station (MNS) Type A Type A test intervals.
2. Does the proposed amendment create Therefore, the proposed change does not
containment integrated leak rate test interval the possibility of a new or different kind of
to 10.5 years. The current Type A test involve a significant reduction in a margin of
accident from any accident previously safety.
interval of 120 months (10 years) would be evaluated?
extended on a one-time basis to no longer Response: No. The NRC staff has reviewed the
than 10.5 years from the last Type A test. The proposed amendment to the TS licensee’s analysis and, based on this
This extension is bounded by the 15 month involves the extension of the MNS Type A review, it appears that the three
extension, permissible only for non-routine
containment integrated leak rate test interval standards of 10 CFR 50.92(c) are
emergent conditions, allowed in accordance
from 10 years to 10.5 years. The current Type satisfied. Therefore, the NRC staff
with NEI [Nuclear Energy Institute] 94–01
A test interval of 120 months (10 years) proposes to determine that the
revision 0. The proposed extension also does
would be extended on a one-time basis to amendment request involves no
not change the test method or procedure. The
10.5 years from the last Type A test. The
containment is designed to provide an significant hazards consideration.
containment and the testing requirements to
essentially leak tight barrier against the
periodically demonstrate the integrity of the
Attorney for licensee: Lara S. Nichols,
uncontrolled release of radioactivity to the Deputy General Counsel, Duke Energy
containment exist to ensure the plant’s
environment for postulated accidents. The Corporation, 526 South Church Street—
ability to mitigate the consequences of an
containment and the testing requirements EC07H, Charlotte, NC 28202.
invoked to periodically demonstrate the accident do not involve any accident
precursors or initiators. The proposed change NRC Branch Chief: Michael T.
integrity of the containment exist to ensure
the plant’s ability to mitigate the does not involve a physical change to the Markley.
consequences of an accident, and do not plant (i.e., no new or different type of
equipment will be installed) or a change to Exelon Generation Company, LLC,
involve the prevention or identification of Docket No. 50–461, Clinton Power
any precursors of an accident. The change in the manner in which the plant is operated or
controlled. Station (CPS), Unit No. 1, DeWitt
dose risk for changing the Type A test
frequency from 10 years to 10.5 years, Therefore, the proposed change does not County, Illinois
measured, as an increase to the total create the possibility of a new or different Date of amendment request: January
integrated plant risk for those accident kind of accident from any previously
evaluated.
25, 2016, as supplemented by letter
sequences influenced by Type A testing, is dated March 31, 2016. A publicly-
0.023 person-rem/year. EPRI [Electric Power 3. Does the proposed amendment involve
a significant reduction in the margin of available version is in ADAMS under
Research Institute] Report No. 1009325,
Revision 2–A states that a very small safety? Accession Nos. ML16025A182 and
population dose is defined as an increase of Response: No. ML16076A077.
≤ 1.0 person-rem per year, or ≤ 1% of the The proposed amendment to TS 5.5.2 Description of amendment request:
total population dose, whichever is less involves the extension of the MNS Type A The proposed amendment would revise
restrictive for the risk impact assessment of containment integrated leak rate test interval the technical specifications (TSs) to
the extended ILRT [integrated leak rate test] to 10.5 years. The current Type A test allow a permanent extension of the
intervals. Therefore, this proposed extension interval of 120 months (10 years) would be
Type ‘‘A’’ integrated leak rate testing
does not involve a significant increase in the extended on a one-time basis to no longer
than 10.5 years from the last Type A test. and Type ‘‘C’’ leak rate testing
probability of an accident previously
This amendment does not alter the manner frequencies. This request also proposes
evaluated.
As documented in NUREG–1493, in which safety limits, limiting safety system to delete information in TS 5.5.13
Performance-Based Containment Leak-Test set points, or limiting conditions for regarding a completed requirement to
Program, Type B and C tests have identified operation are determined. The specific perform Type ‘‘C’’ testing in 2008.
a very large percentage of containment requirements and conditions of the TS Basis for proposed no significant
leakage paths, and the percentage of Containment Leak Rate Testing Program exist hazards consideration determination:
containment leakage paths that are detected to ensure that the degree of containment As required by 10 CFR 50.91(a), the
only by Type A testing is very small. The structural integrity and leak tightness that is licensee has provided its analysis of the
MNS Type A test history supports this considered in the plant safety analysis is
issue of no significant hazards
conclusion. maintained. The overall containment leak
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The integrity of the containment is subject rate limit specified by TS is maintained. consideration which is presented below:
to two types of failure mechanisms that can The proposed change involves only the 1. Does the proposed amendment involve
be categorized as: (1) Activity based, and; (2) extension of the interval between Type A a significant increase in the probability or
time based as previously discussed. Activity containment leak rate tests for MNS. The consequences of an accident previously
based failure mechanisms are defined as proposed surveillance interval extension is evaluated?
degradation due to system and/or component bounded by the 15-year ILRT interval Response: No.
modifications or maintenance. Local leak rate currently authorized within NEI 94–01, The proposed activity involves the
test requirements and administrative controls Revisions 2–A and 3–A. Industry experience extension of the Clinton Power Station (CPS),
such as configuration management and supports the conclusion that Type B and C Unit 1, Type A containment test interval to
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![28896 Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices
15 years, and the extension of the Type C test that is detectable only by a Type A test. interval currently authorized within NEI
interval to 75 months. The current Type A Based on the above, the proposed extensions [Nuclear Energy Institute] 94–01, Revision 3–
test interval of 120 months (10 years) would do not significantly increase the A. Industry experience supports the
be extended on a permanent basis to no consequences of an accident previously conclusion that Type B and C testing detects
longer than 15 years from the last Type A evaluated. a large percentage of containment leakage
test. The current Type C test interval of 60 The proposed amendment also deletes an paths and that the percentage of containment
months for selected components would be exception previously granted to allow one- leakage paths that are detected only by Type
extended on a performance basis to no longer time extension of the ILRT test frequency for A testing is small. The containment
than 75 months. Extensions of up to nine CPS. This exception was for an activity that inspections performed in accordance with
months (total maximum interval of 84 has already taken place; therefore, this ASME Section Xl, and TS serve to provide
months for Type C tests) are permissible only deletion is solely an administrative action a high degree of assurance that the
for non-routine emergent conditions. The that does not result in any change in how containment would not degrade in a manner
proposed extension does not involve either a CPS is operated. that is detectable only by Type A testing. The
physical change to the plant or a change in Therefore, the proposed change does not combination of these factors ensures that the
the manner in which the plant is operated or result in a significant increase in the margin of safety in the plant safety analysis
controlled. The containment is designed to probability or consequences of an accident is maintained. The design, operation, testing
provide an essentially leak tight barrier previously evaluated. methods and acceptance criteria for Type A,
against the uncontrolled release of 2. Does the proposed change create the B, and C containment leakage tests specified
radioactivity to the environment for possibility of a new or different kind of in applicable codes and standards would
postulated accidents. As such, the accident from any accident previously continue to be met, with the acceptance of
containment and the testing requirements evaluated? this proposed change, since these are not
invoked to periodically demonstrate the Response: No. affected by changes to the Type A and Type
integrity of the containment exist to ensure The proposed amendment to the TS 5.5.13, C test intervals.
the plant’s ability to mitigate the ‘‘Primary Containment Leakage Rate Testing The proposed amendment also deletes
consequences of an accident, and do not Program,’’ involves the extension of the CPS, exceptions previously granted to allow one
involve the prevention or identification of Unit 1 Type A containment test interval to time extensions of the ILRT test frequency for
any precursors of an accident. 15 years and the extension of the Type C test CPS, Unit 1. This exception was for an
The change in dose risk for changing the interval to 75 months. The containment and activity that has taken place; therefore, the
Type A Integrated Leak Rate Test (ILRT) the testing requirements to periodically deletion is solely an administrative action
interval from three-per-ten years to once-per- demonstrate the integrity of the containment and does not change how CPS is operated
fifteen-years, measured as an increase to the exist to ensure the plant’s ability to mitigate and maintained. Thus, there is no reduction
total integrated dose risk for all accident the consequences of an accident. in any margin of safety.
sequences, is 3.80E–03 person-rem/yr using The proposed change does not involve a Therefore, the proposed changes do not
the EPRI [Electric Power Research Institute] physical change to the plant (i.e., no new or involve a significant reduction in a margin of
guidance with the base case corrosion different type of equipment will be installed) safety.
included. This change meets both of the nor does it alter the design, configuration, or
related acceptance criteria for change in change the manner in which the plant is The NRC staff has reviewed the
population dose of less than 1.0 person-rem/ operated or controlled beyond the standard licensee’s analysis and, based on this
yr or less than 1% person-rem/yr. The change functional capabilities of the equipment. review, it appears that the three
in dose risk drops to 9.37E–04 person-rem/ The proposed amendment also deletes an standards of 10 CFR 50.92(c) are
yr when using the EPRI Expert Elicitation exception previously granted to allow one- satisfied. Therefore, the NRC staff
methodology. The change in dose risk meets time extension of the ILRT test frequency for proposes to determine that the
both of the related acceptance for change in CPS. This exception was for an activity that amendment request involves no
population dose of less than 1.0 person-rem/ has already taken place; therefore, this significant hazards consideration.
yr or less than 1% person-rem/yr. Therefore, deletion is solely an administrative action Attorney for licensee: Bradley J.
this proposed extension does not involve a that does not result in any change in how
Fewell, Associate General Counsel,
significant increase in the probability of an CPS is operated.
accident previously evaluated. Therefore, the proposed change does not Exelon Nuclear, 4300 Winfield Road,
In addition, as documented in NUREG– create the possibility of a new or different Warrenville, IL 60555.
1493, Types B and C tests have identified a kind of accident from any previously NRC Acting Branch Chief: Justin C.
very large percentage of containment leakage evaluated. Poole.
paths, and the percentage of containment 3. Does the proposed change involve a
significant reduction in a margin of safety?
Exelon Generation Company, LLC,
leakage paths that are detected only by Type
A testing is very small. The CPS, Unit 1 Type Response: No. Docket No. 50–410, Nine Mile Point
A test history supports this conclusion. The proposed amendment to TS 5.5.13 Nuclear Station, Unit 2, Oswego County,
The integrity of the containment is subject involves the extension of the CPS, Unit 1 New York
to two types of failure mechanisms that can Type A containment test interval to 15 years Date of amendment request: February
be categorized as: (1) Activity based, and, (2) and the extension of the Type C test interval 23, 2016. A publicly-available version is
time based. Activity based failure to 75 months for selected components. This
mechanisms are defined as degradation due amendment does not alter the manner in
in ADAMS under Accession No.
to system and/or component modifications or which safety limits, limiting safety system set ML16054A359.
maintenance. Local leak rate test points, or limiting conditions for operation Description of amendment request:
requirements and administrative controls are determined. The specific requirements The amendment would revise the
such as configuration management and and conditions of the TS Containment Leak Technical Specifications to incorporate
procedural requirements for system Rate Testing Program exist to ensure that the previously NRC-approved Industry/
restoration ensure that containment integrity degree of containment structural integrity Technical Specification Task Force 439
is not degraded by plant modifications or and leaktightness that is considered in the (TSTF–439), Revision 2, ‘‘Eliminate
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maintenance activities. The design and plant safety analysis is maintained. The Second Completion Times Limiting
construction requirements of the overall containment leak rate limit specified Time From Discovery of Failure To
containment combined with the containment by TS is maintained.
inspections performed in accordance with The proposed change involves the
Meet an LCO.’’
American Society of Mechanical Engineers extension of the interval between Type A Basis for proposed no significant
(ASME) Section XI, and Technical containment leak rate tests and Type C tests hazards consideration determination:
Specifications (TS) requirements serve to for CPS, Unit 1. The proposed surveillance As required by 10 CFR 50.91(a), the
provide a high degree of assurance that the interval extension is bounded by the 15-year licensee has provided its analysis of the
containment would not degrade in a manner ILRT interval and the 75-month Type C test issue of no significant hazards
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consideration, which is presented satisfied. Therefore, the NRC staff Reactivity Anomalies surveillance will only
below: proposes to determine that the provide a new, more efficient method of
amendment request involves no detecting an unexpected change in core
1. Does the proposed amendment involve reactivity.
a significant increase in the probability or significant hazards consideration. Since all systems continue to be operated
consequences of an accident previously Attorney for licensee: Tamra Domeyer, within their design bases, no new failure
evaluated? Associate General Counsel, Exelon modes are introduced and the possibility of
Response: No. Generation Company, LLC, 4300 a new or different kind of accident is not
The proposed change eliminates certain Winfield Road, Warrenville, IL 60555. created.
Completion Times from the Technical NRC Branch Chief: Travis L. Tate. 3. Does the proposed amendment involve
Specifications. Completion Times are not an a significant reduction in a margin of safety?
initiator to any accident previously Exelon Generation Company, LLC, Response: No.
evaluated. As a result, the probability of an Docket Nos. 50–220 and 50–410, Nine This proposed TS amendment proposes to
accident previously evaluated is not affected. Mile Point Nuclear Station (NMPNS), change the method for performing the
The consequences of an accident during the Units 1 and 2, Oswego County, New Reactivity Anomalies surveillance from a
revised Completion Time are no different comparison of predicted to actual control rod
York
than the consequences of the same accident density to a comparison of predicted to
during the existing Completion Times. As a Date of amendment request: March monitored keff. The direct comparison of keff
result, the consequences of an accident 18, 2016. A publicly-available version is provides a technically superior method of
previously evaluated are not affected by this in ADAMS under Accession No. calculating any differences in the expected
change. The proposed change does not alter ML16078A065. core reactivity. The Reactivity Anomalies
or prevent the ability of SSCs [systems, Description of amendment request: surveillance will continue to be performed at
structures, and components] from performing the same frequency as is currently required
their intended function to mitigate the
The amendment would revise the
by the TS, only the method of performing the
consequences of an initiating event within Technical Specifications (TS) surveillance will be changed. Consequently,
the assumed acceptance limits. The proposed concerning a change to the method of core reactivity assumptions made in safety
change does not affect the source term, calculating core reactivity for the analyses will continue to be adequately
containment isolation, or radiological release purpose of performing the Reactivity verified.
assumptions used in evaluating the Anomalies surveillance at NMPNS,
radiological consequences of an accident
The NRC staff has reviewed the
Units 1 and 2. licensee’s analysis and, based on this
previously evaluated. Further, the proposed Basis for proposed no significant
change does not increase the types or review, it appears that the three
hazards consideration determination: standards of 10 CFR 50.92(c) are
amounts of radioactive effluent that may be
released offsite, nor significantly increase As required by 10 CFR 50.91(a), the satisfied. Therefore, the NRC staff
individual or cumulative occupational/ licensee has provided its analysis of the proposes to determine that the
public radiation exposures. The proposed issue of no significant hazards amendment request involves no
change is consistent with the safety analysis consideration, which is presented significant hazards consideration.
assumptions and resultant consequences. below: Attorney for licensee: Tamra Domeyer,
Therefore, the proposed change does not Associate General Counsel, Exelon
1. Does the proposed amendment involve
involve a significant increase in the
probability or consequences of an accident
a significant increase in the probability or Generation Company, LLC, 4300
previously evaluated. consequences of an accident previously Winfield Road, Warrenville, IL 60555.
2. Does the proposed amendment create evaluated? NRC Branch Chief: Travis L. Tate.
the possibility of a new or different kind of Response: No.
The proposed TS changes do not affect any Exelon Generation Company, LLC,
accident from any accident previously
plant systems, structures, or components Docket Nos. STN 50–456 and STN 50–
evaluated?
Response: No. designed for the prevention or mitigation of 457, Braidwood Station, Units 1 and 2,
The proposed change does not involve a previously evaluated accidents. The Will County, Illinois and Docket Nos.
physical alteration of the plant (i.e., no new amendment would only change how the STN 50–454 and STN 50–455, Byron
or different type of equipment will be Reactivity Anomalies surveillance is Station, Unit Nos. 1 and 2, Ogle County,
installed) or a change in the methods performed. Verifying that the core reactivity Illinois
governing normal plant operation. The is consistent with predicted values ensures
proposed change does not alter any that accident and transient safety analyses Date of amendment request: February
assumptions made in the safety analysis. remain valid. This amendment changes the 23, 2016. A publicly-available version is
Therefore, the proposed change does not TS requirements such that, rather than in ADAMS under Accession No.
create the possibility of anew or different performing the surveillance by comparing ML16055A149.
kind of accident from any accident predicted to actual control rod density, the Description of amendment request:
previously evaluated. surveillance is performed by a direct The amendment would (1) revise
3. Does the proposed amendment involve comparison of keff. Technical Specification (TS) 4.2.1,
a significant reduction in a margin of safety? Therefore, since the Reactivity Anomalies
‘‘Reactor Core, Fuel Assemblies,’’ to add
Response: No. surveillance will continue to be performed by
a viable method, the proposed amendment Optimized ZIRLOTM, as an approved
The proposed change to delete the second
does not involve a significant increase in the fuel rod cladding material, (2) revise TS
Completion Time does not alter the manner
in which safety limits, limiting safety system probability or consequence of a previously 5.6.5.b to add the Westinghouse topical
settings or limiting conditions for operation evaluated accident. reports for Optimized ZIRLOTM and
are determined. The safety analysis 2. Does the proposed amendment create ZIRLO®, and (3) revise TS 5.6.5.b with
acceptance criteria are not affected by this the possibility of a new or different kind of a non-technical change to the Reference
accident from any accident previously 11 title (replace a semicolon with a
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change. The proposed change will not result
in plant operation in a configuration outside evaluated? period).
of the design basis. Therefore, the proposed Response: No. Basis for proposed no significant
change does not involve a significant This TS amendment request does not hazards consideration determination:
reduction in a margin of safety. involve any changes to the operation, testing,
or maintenance of any safety-related, or
As required by 10 CFR 50.91(a), the
The NRC staff has reviewed the otherwise important to safety systems. All licensee has provided its analysis of the
licensee’s analysis and, based on this systems important to safety will continue to issue of no significant hazards
review, it appears that the three be operated and maintained within their consideration, which is presented
standards of 10 CFR 50.92(c) are design bases. The proposed changes to the below:
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![28898 Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices
EGC [Exelon Generation Company] has LOCA analyses using Optimized ZIRLOTM installation in MODES 1, 2, and 3, and drain-
evaluated the proposed changes for properties will demonstrate that the down of the reactor well in MODE 3. The
Braidwood and Byron, using the criteria in acceptance criteria of 10 CFR 50.46 have probability of an accident previously
10 CFR 50.92, and has determined that the been satisfied. Therefore, the proposed evaluated is unrelated to the water level in
proposed changes do not involve a change does not involve a significant these pools, since they are mitigating
significant hazards consideration. The reduction in a margin of safety. systems. The operation or failure of a
following information is provided to support Based on the above, EGC concludes that mitigating system does not contribute to the
a finding of no significant hazards the proposed amendment to allow the use of occurrence of an accident. No active or
consideration. Optimized ZIRLOTM fuel cladding material passive failure mechanisms that could lead to
Criteria does not involve a significant hazards an accident are affected by these proposed
1. Does the proposed change involve a consideration under the standards set forth in changes.
significant increase in the probability or 10 CFR 50.92(c), and, accordingly, a finding Suppression pool water levels are
consequences of an accident previously of ‘‘no significant hazards consideration’’ is increased during upper pool gate installation
evaluated? justified. in MODES 1, 2, and 3 and during reactor well
Response: No. drain-down in MODE 3, with a potential for
The proposed change would allow the use The NRC staff has reviewed the an increased probability of drywell flooding
of Optimized ZIRLOTM clad nuclear fuel in licensee’s analysis and, based on this during an inadvertent dump of the upper
the reactors. The NRC approved topical review, it appears that the three containment pool. An inadvertent dump of
report WCAP–12610–P–A & CENPD–404–P– standards of 10 CFR 50.92(c) are the upper pool during any period of
A, Addendum 1–A, ‘‘Optimized ZIRLOTM satisfied. Therefore, the NRC staff operation with a pressurized vessel does not
prepared by Westinghouse Electric Company proposes to determine that the represent, in and of itself, any significant
LLC (Westinghouse), addresses Optimized amendment request involves no hazard to the public, the plant operating
ZIRLOTM and demonstrates that Optimized personnel, or any plant equipment. The
ZIRLOTM has essentially the same properties
significant hazards consideration. piping components which would be affected
as currently licensed ZIRLO®. The fuel Attorney for licensee: Bradley J. in this event have been analyzed for the
cladding itself is not an accident initiator and Fewell, Associate General Counsel, flooding effect, and it has been determined
does not affect accident probability. With the Exelon Nuclear, 4300 Winfield Road, that this event could not initiate a loss of
approved exemption, use of Optimized Warrenville, IL 60555. coolant accident (LOCA).
ZIRLOTM fuel cladding will continue to meet NRC Acting Branch Chief: Justin C. The changes have no impact on the ability
all 10 CFR 50.46 acceptance criteria and, Poole. of any of the emergency core cooling systems
therefore, will not increase the consequences (ECCS) to function adequately, since
of an accident. Therefore, the proposed FirstEnergy Nuclear Operating adequate net positive suction head (NPSH) is
change does not involve a significant Company, Docket No. 50–440, Perry maintained. The increase in suppression pool
increase in the probability or consequences Nuclear Power Plant, Unit No. 1, Lake water level to compensate for the reduction
of an accident previously evaluated. County, Ohio in UCP volume will provide reasonable
2. Does the proposed change create the assurance that the minimum post-accident
possibility of a new or different kind of Date of amendment request: March vent coverage is adequate to assure the
accident from any accident previously 15, 2016. A publicly-available version is pressure suppression function of the
evaluated? in ADAMS under Accession No. suppression pool is accomplished. The
Response: No. ML16075A411. suppression pool water level will be raised
Use of Optimized ZIRLOTM clad fuel will Description of amendment request: above the current high water level for the
not result in changes in the operation or The proposed amendment would revise proposed reactor well drain-down activity
configuration of the facility. Topical Report Technical Specification (TS) 3.6.2.2, only after the reactor pressure has been
WCAP–12610–P–A & CENPD–404–P–A, ‘‘Suppression Pool Water Level,’’ as reduced sufficiently to assure that the
Addendum 1–A, demonstrated that the hydrodynamic loads from a loss of coolant
well as TS surveillance requirements
material properties of Optimized ZIRLOTM accident will not exceed the design values.
are similar to those of standard ZIRLO®. 3.6.2.4.1 and 3.6.2.4.4 associated with The reduced reactor pressure will also ensure
Therefore, Optimized ZIRLOTM fuel rod TS 3.6.2.4, ‘‘Suppression Pool Makeup that the loads due to main steam safety relief
cladding will perform similarly to those System (SPMU),’’ to allow installation valve actuation with an elevated pool level
fabricated from standard ZIRLO® thus of the reactor well to steam dryer storage are within the design loads.
precluding the possibility of the fuel pool gate in the upper containment pool Relative to dose rates on the refuel floor,
cladding becoming an accident initiator and (UCP) in MODES 1, 2, and 3. The the resultant dose rates from the reactor in
causing a new or different type of accident. proposed amendment would also create MODES 3 and 4 are the same regardless of
Therefore, the proposed change does not new special operations TS 3.10.9, a drain-down of the upper pool reactor well.
create the possibility of a new or different Relative to a low pressure LOCA in MODE
‘‘Suppression Pool Makeup—MODE 3
kind of accident from any previously 3, the reduced post-LOCA containment
evaluated. Upper Containment Pool Drain-Down,’’ pressure and the decay time to reach MODE
3. Does the proposed change involve a to allow draining of the reactor well 3 conditions ensures that post-accident dose
significant reduction in a margin of safety? portion of the UCP in MODE 3. consequences are bounded by the design-
Response: No. Basis for proposed no significant basis accident LOCA.
The proposed change will not involve a hazards consideration determination: Therefore, the proposed amendment does
significant reduction in the margin of safety. As required by 10 CFR 50.91(a), the not significantly increase the probability or
Topical Report WCAP–12610–P–A & licensee has provided its analysis of the consequences of an accident previously
CENPD–404–P–A, Addendum 1–A, issue of no significant hazards evaluated.
demonstrated that the material properties of 2. Does the proposed amendment create
consideration, which is presented
the Optimized ZIRLOTM are not significantly the possibility of a new or different kind of
different from those of standard ZIRLO®.
below: accident from an accident previously
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Optimized ZIRLOTM is expected to perform 1. Does the proposed amendment involve evaluated?
similarly to standard ZIRLO® for all normal a significant increase in the probability or Response: No.
operating and accident scenarios, including consequences of an accident previously The proposed changes specify different
both loss of coolant accident (LOCA) and evaluated? water level requirements in the upper
non-LOCA scenarios. For LOCA scenarios, Response: No. containment pool and suppression pool to
where the slight difference is Optimized The changes proposed in the license permit gate installation in MODES 1, 2, and
ZIRLOTM material properties relative to amendment request specify different water 3, and drain-down of the reactor well in
standard ZIRLO® could have some impact on level requirements in the upper containment MODE 3. These changes do not affect or alter
the overall accident scenario, plant-specific pool and suppression pool to permit gate the ability of the suppression pool makeup
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(SPMU) system to perform its design ‘‘Low Temperature Overpressure orifice to be used during LTOP operation to
function. The proposed change in the pool Protection (LTOP) System,’’ to reflect ensure that the total maximum mass injection
water levels will maintain the design the mass input transient analysis that capability with the NCP remained bounded
function of mitigating the pressure and assumes an emergency core cooling by the LTOP mass injection analysis. These
temperature increase generated by a LOCA, changes were implemented under 10 CFR
system (ECCS) centrifugal charging
and will maintain the required drywell vent 50.59. However, no physical changes are
coverage during post-accident ECCS draw pump (CCP) and the normal charging being made to the plant as a result of the
down. pump (NCP) capable of simultaneously proposed license amendment.
The altered water levels in the pools do not injecting into the reactor coolant system Therefore, the proposed change does not
create a different type of accident than (RCS) during TS 3.4.12 applicability. involve a significant increase in the
presently evaluated. With the reduced Basis for proposed no significant probability or consequences of an accident
pressure in the reactor coolant system, the hazards consideration determination: previously evaluated.
GOTHIC computer program simulations As required by 10 CFR 50.91(a), the 2. Does the proposed change create the
demonstrate that the accident responses at licensee has provided its analysis of the possibility of a new or different accident
defined conditions with the reactor well issue of no significant hazards from any accident previously evaluated?
drained in MODE 3 are bounded by the Response: No.
consideration, which is presented
current design basis accidents. The proposed change revises TS 3.4.12 to
Therefore, the proposed amendment does
below:
allow an ECCS CCP and the NCP aligned to
not create the possibility of a new or different 1. Does the proposed change involve a LTOP orifice to be capable of simultaneously
kind of accident from any accident significant increase in the probability or injecting into the RCS during low RCS
previously evaluated. consequences of an accident previously pressures and temperatures. The LCO
3. Does the proposed amendment involve evaluated? provides RCS overpressure protection by
a significant reduction in a margin of safety? Response: No. having a minimum coolant input capability
Response: No. The proposed change revises TS 3.4.12 to and have adequate pressure relief capability.
The proposed changes to the UCP and the allow an ECCS CCP and the NCP aligned to Analyses have demonstrated that one PORV
suppression pool water levels do not LTOP orifice to be capable of injecting into or an RCS vent of at least 2.07 square inches
introduce any new setpoints at which the RCS during low RCS pressures and is capable of limiting the RCS pressure
protective or mitigating actions are initiated. temperatures. The LCO [Limiting Condition excursions below the 10 CFR 50, Appendix
Current instrument setpoints remain for Operation] provides RCS overpressure G limits for the design basis LTOP limits.
unaltered by this change. Although the water protection by having a minimum coolant The proposed change will not physically
levels are adjusted for the UCP gate input capability and have adequate pressure alter the plant (no new or different type of
installation and the reactor well drain-down relief capability. Analyses have demonstrated equipment will be installed) or change the
activity, the design and functioning of the that one power operated relief valve (PORV) methods governing normal plant operation.
containment pressure suppression system or an RCS vent of at least 2.07 square inches The proposed change does not introduce new
remains unchanged. The proposed total is capable of limiting the RCS pressure accident initiators or impact assumptions
water volume is sufficient to provide high excursions below the 10 CFR 50, Appendix
made in the safety analysis. Testing
confidence that the pressure suppression and G limits for the design basis LTOP limits.
requirements continue to demonstrate that
The proposed change does not adversely
containment systems will be capable of the LCOs are met and the system components
affect accident initiators or precursors, nor
mitigating large and small break accidents. are functional.
alter the design assumptions, conditions, and
All analyzed accident results remain within Therefore, the proposed change does not
configuration of the facility or the manner in
the design values for the structures and create the possibility of a new or different
which the plant is operated and maintained.
equipment. accident from any accident previously
The proposed change does not adversely
Therefore, the proposed amendment does evaluated.
affect the ability of structures, systems, and
not involve a significant reduction in a 3. Does the proposed change involve a
components to perform their intended safety
margin of safety. significant reduction in a margin of safety?
function to mitigate the consequences of an
initiating event within the assumed Response: No.
The NRC staff has reviewed the The proposed change does not alter the
licensee’s analysis and, based on this acceptance limits. The proposed change does
not affect the source term, containment manner in which safety limits, limiting safety
review, it appears that the three system settings, or limiting conditions for
isolation, or radiological release assumptions
standards of 10 CFR 50.92(c) are operation are determined. The safety analysis
used in evaluating the radiological
satisfied. Therefore, the NRC staff consequences of any accident previously acceptance criteria are not impacted by this
proposes to determine that the evaluated. Further, the proposed change does change. The proposed change will not result
amendment request involves no not increase the types and amounts of in plant operation in a configuration outside
significant hazards consideration. radioactive effluent that may be released the design basis.
Attorney for licensee: David W. offsite, nor significantly increase individual Therefore, the proposed change does not
or cumulative occupational/public radiation involve a significant reduction in a margin of
Jenkins, Attorney, FirstEnergy
exposure. safety.
Corporation, Mail Stop A–GO–15, 76
South Main Street, Akron, OH 44308. The NRC has previously evaluated a
similar LAR [license amendment request] The NRC staff has reviewed the
NRC Branch Chief: David J. Wrona. related to Wolf Creek Generating Station. In licensee’s analysis and, based on this
Pacific Gas and Electric Company Amendment No. 207, the NRC concluded review, it appears that the three
(PG&E), Docket Nos. 50–275 and 50– that the proposed change does not involve a standards of 10 CFR 50.92(c) are
significant increase in the probability or satisfied. Therefore, the NRC staff
323, Diablo Canyon Nuclear Power
consequences of an accident previously proposes to determine that the
Plant, Units 1 and 2, San Luis Obispo evaluated [ADAMS Accession No.
asabaliauskas on DSK3SPTVN1PROD with NOTICES
County, California amendment requests involve no
ML13282A534].
significant hazards consideration.
Date of amendment request: March In 2007, PG&E replaced the Unit 1 non-
23, 2016. A publicly-available version is safety-related PDP [positive displacement Attorney for licensee: Jennifer Post,
pump] with a non-safety-related CCP, called Esq., Pacific Gas and Electric Company,
in ADAMS under Accession No.
the NCP, in order to alleviate operational P.O. Box 7442, San Francisco, CA
ML16084A588. issues associated with the PDP. In 2008,
Description of amendment request: 94120.
PG&E performed the replacement on Unit 2.
The proposed amendment would revise PG&E also designed, tested, and installed an NRC Branch Chief: Robert J.
Technical Specification (TS) 3.4.12, FCO [flow choking orifice] called the LTOP Pascarelli.
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![Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices 28901
Response: No. in a significant reduction in margin for any one train of safe shutdown equipment will
The proposed changes will revise the safety function, and were evaluated in the remain functional in the event of an
licensing basis documents related to the fuel Safety Evaluation Report for WCAP–17524– Appendix R fire, even though a radiant
system design, nuclear design, thermal P–A Revision 1 and found to be acceptable. energy shield will not be provided for two
hydraulic design, and accident analyses. Therefore, the proposed amendment does raceway containing safe shutdown circuits.
The UFSAR [Updated Final Safety not involve a significant reduction in a
Based on this, the proposed amendment will
Analysis Report] Chapter 15 accident margin of safety.
analyses describe the analyses of various not alter the requirements or function for
design basis transients and accidents to The NRC staff has reviewed the systems required during accident conditions.
demonstrate compliance of the AP1000 licensee’s analysis and, based on this Therefore, the proposed change does not
design with the acceptance criteria for these review, it appears that the three create the possibility of a new or different
events. The acceptance criteria for the standards of 10 CFR 50.92(c) are kind of accident from any accident
various events are based on meeting the satisfied. Therefore, the NRC staff previously evaluated.
relevant regulations, general design criteria, proposes to determine that the 3. Does the proposed amendment involve
the Standard Review Plan, and are a function amendment request involves no a significant reduction in a margin of safety?
of the anticipated frequency of occurrence of
significant hazards consideration. Response: No.
the event and potential radiological
consequences to the public. As such, each
Attorney for licensee: Kathryn M. A fire hazards analysis was performed for
design-basis event is categorized accordingly Sutton, Morgan, Lewis & Bockius LLC, the areas under the scope of this amendment.
based on these considerations. As discussed 1111 Pennsylvania Avenue NW, This fire hazards analysis demonstrates that
in Section 5.3 of WCAP–17524–P–A Revision Washington, DC 20004–2514. one train of safe shutdown equipment will
1, the revised accident analyses maintain NRC Acting Branch Chief: John remain functional in the event of an
their plant conditions, and thus their McKirgan. Appendix R fire, even though a radiant
frequency designation and consequence level energy shield will not be provided for two
as previously evaluated. As confirmed in the Tennessee Valley Authority, Docket No. raceway containing safe shutdown circuits.
Safety Evaluation Report (SER), the revised 50–390, Watts Bar Nuclear Plant (WBN), Therefore, the proposed change does not
analyses meet the applicable guidelines in Unit 1, Rhea County, Tennessee involve a significant reduction in a margin of
the Standard Review Plan.
Date of amendment request: February safety.
Therefore, the proposed amendment does
not involve a significant increase in the 23, 2016. A publicly-available version is
in ADAMS under Accession No. The NRC staff has reviewed the
probability or consequences of an accident
previously evaluated. ML16054A585. licensee’s analysis and, based on this
2. Does the proposed amendment create Description of amendment request: review, it appears that the three
the possibility of a new or different kind of The amendment would revise the WBN standards of 10 CFR 50.92(c) are
accident from any accident previously Dual Unit Fire Protection Report and satisfied. Therefore, the NRC staff
evaluated? would revise the associated License proposes to determine that the
Response: No. Condition regarding the WBN fire amendment request involves no
The proposed changes will revise the protection program. Specifically, the significant hazards consideration.
licensing basis documents related to the fuel
system design, nuclear design, thermal
amendment requests approval of a Attorney for licensee: Sherry A. Quirk,
hydraulic design, and accident analyses. deviation from the physical separation Executive Vice President and General
The proposed changes would not introduce requirements of 10 CFR part 50, Counsel, Tennessee Valley Authority,
a new failure mode, fault, or sequence of appendix R, section III.G.2.d. 400 West Summit Hill Drive, Knoxville,
events that could result in a radioactive Basis for proposed no significant
TN 37902.
material release. The proposed changes do hazards consideration determination:
not alter the design, configuration, or method As required by 10 CFR 50.91(a), the NRC Branch Chief: Benjamin G.
of operation of the plant beyond standard licensee has provided its analysis of the Beasley.
functional capabilities of the equipment. issue of no significant hazards
Therefore, the proposed amendment does III. Previously Published Notices of
consideration, which is presented Consideration of Issuance of
not create the possibility of a new or different
below:
kind of accident from any accident Amendments to Facility Operating
previously evaluated. 1. Does the proposed amendment involve Licenses and Combined Licenses,
3. Does the proposed amendment involve a significant increase in the probability or Proposed No Significant Hazards
a significant reduction in a margin of safety? consequences of an accident previously
Consideration Determination, and
Response: No. evaluated?
The proposed changes will revise the Response: No. Opportunity for a Hearing
licensing basis documents related to the fuel A fire hazards analysis was performed for
system design, nuclear design, thermal the areas under the scope of this amendment.
The following notices were previously
hydraulic design, and accident analyses. This fire hazards analysis demonstrates that published as separate individual
Safety margins are applied at many levels one train of safe shutdown equipment will notices. The notice content was the
to the design and licensing basis functions remain functional in the event of an same as above. They were published as
and to the controlling values of parameters to Appendix R fire, even though a radiant individual notices either because time
account for various uncertainties and to energy shield will not be provided for two did not allow the Commission to wait
avoid exceeding regulatory or licensing raceway containing safe shutdown circuits. for this biweekly notice or because the
limits. UFSAR Subsection 4.1.1 presents the Therefore, the proposed change does not
action involved exigent circumstances.
Principle Design Requirements imposed on involve a significant increase in the
the fuel and control rod mechanism design probability or consequences of an accident They are repeated here because the
asabaliauskas on DSK3SPTVN1PROD with NOTICES
to ensure that the performance and safety previously evaluated. biweekly notice lists all amendments
criteria described in UFSAR Chapter 4 and 2. Does the proposed amendment create issued or proposed to be issued
Chapter 15 are met. The revised fuel system the possibility of a new or different kind of involving no significant hazards
design, nuclear design, thermal hydraulic accident from any accident previously consideration.
design, and accident analyses maintain the evaluated?
same Principle Design Requirements, and Response: No. For details, see the individual notice
further, satisfy the applicable regulations, A fire hazards analysis was performed for in the Federal Register on the day and
general design criteria, and Standard Review the areas under the scope of this amendment. page cited. This notice does not extend
Plan. The effects of the changes do not result This fire hazards analysis demonstrates that the notice period of the original notice.
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![28904 Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices
Pacific Gas and Electric Company Southern Nuclear Operating Company, Amendment Nos.: 278 and 222. A
(PG&E), Docket Nos. 50–275 and 50– Docket Nos. 52–025 and 52–026, Vogtle publicly-available version is in ADAMS
323, Diablo Canyon Nuclear Power Electric Generating Plant (VEGP), Units under Accession No. ML16090A174;
Plant, Unit Nos. 1 and 2, San Luis 3 and 4, Burke County, Georgia documents related to these amendments
Obispo County, California Date of amendment request: are listed in the Safety Evaluation
September 1, 2015. enclosed with the amendments.
Date of application for amendments: Facility Operating License Nos. DPR–
June 26, 2013, as supplemented by Brief description of amendment: The
amendment authorized changes to the 57 and NPF–5: Amendments revised the
letters dated September 29, October 27, Facility Operating Licenses and
October 29, November 26, and VEGP Units 3 and 4 plant specific
emergency planning inspections, tests, Technical Specifications.
December 31, 2014; February 25 (two Date of initial notice in Federal
letters), May 7, October 15, and analyses, and acceptance criteria
(ITAAC) in Appendix C of VEGP Units Register: March 17, 2015 (80 FR
December 31, 2015; and January 28, 13911). The supplemental letters dated
2016. 3 and 4 Combined Operating Licenses
(COLs). The changes authorize the June 16 and November 24, 2015,
Brief description of amendments: The removal of the copy of Updated Final provided additional information that
amendments permit the PG&E (the Safety Analysis Report Table 7.5–1, clarified the application, did not expand
licensee) to adopt a new fire protection ‘‘Post-Accident Monitoring System’’ the scope of the application as originally
licensing basis based on National Fire from ITAAC in Appendix C of the VEGP noticed, and did not change the staff’s
Protection Association (NFPA) Standard Units 3 and 4 COLs. original proposed no significant hazards
805, ‘‘Performance-Based Standard for Date of issuance: March 30, 2016. consideration determination as
Fire Protection for Light Water Reactor Effective date: As of the date of published in the Federal Register.
Generating Plants (2001 Edition),’’ at issuance and shall be implemented The Commission’s related evaluation
Diablo Canyon Power Plant, Units 1 and within 30 days of issuance. of the amendment is contained in a
2, that complies with the requirements Amendment No.: 47. A publicly- Safety Evaluation dated April 14, 2016.
of 10 CFR 50.48(a) and (c) and the available version is in ADAMS under No significant hazards consideration
guidance in Revision 1 of Regulatory Accession No. ML16061A220; comments received: No.
Guide 1.205, ‘‘Risk Informed documents related to these amendments Wolf Creek Nuclear Operating
Performance-Based Fire Protection for are listed in the Safety Evaluation Corporation, Docket No. 50–482, Wolf
Existing Light-Water Nuclear Power enclosed with the amendments. Creek Generating Station, Coffey
Plants,’’ December 2009. Facility Combined Licenses Nos. NPF– County, Kansas
Date of issuance: April 14, 2016. 91 and NPF–92: Amendment revised the
Facility Combined Licenses. Date of amendment request:
Effective date: As of its date of September 23, 2015.
issuance and shall be implemented as Date of initial notice in Federal
Register: October 27, 2015 (80 FR Brief description of amendment: The
described in the transition license amendment revised the diesel generator
conditions. 65807).
The Commission’s related evaluation (DG) full load rejection test and
Amendment Nos.: Unit 1—225; Unit of the amendment is contained in a endurance and margin test specified by
2—227. A publicly-available version is Safety Evaluation dated March 30, 2015. Technical Specification (TS) 3.8.1, ‘‘AC
in ADAMS under Accession No. No significant hazards consideration [Alternating Current] Sources—
ML16035A441; documents related to comments received: No. Operating,’’ Surveillance Requirements
these amendments are listed in the (SR) 3.8.1.10 and 3.8.1.14, respectively.
Safety Evaluation enclosed with the Southern Nuclear Operating Company, The change adds a new Note to SR
amendments. Inc., Georgia Power Company, 3.8.1.10 and SR 3.8.1.14, consistent with
Oglethorpe Power Corporation, Technical Specification Task Force
Facility Operating License Nos. DPR–
Municipal Electric Authority of Georgia, (TSTF) traveler TSTF–276–A, Revision
80 and DPR–82: The amendments
City of Dalton, Georgia, Docket Nos. 50– 2, ‘‘Revise DG full load rejection test.’’
revised the Facility Operating Licenses
321 and 50–366, Edwin I. Hatch Nuclear The Note allows the full load rejection
and Technical Specifications.
Plant, Unit Nos. 1 and 2, Appling test and endurance and margin test to be
Date of initial notice in Federal County, Georgia
Register: December 26, 2013 (78 FR performed at the specified power factor
78408). The supplemental letters dated Date of amendment request: January with clarifications addressing situations
October 3, 2013; September 29, October 13, 2015, as supplemented by letters when the power factor cannot be
27, October 29, November 26, and dated June 16 and November 24, 2015. achieved.
December 31, 2014; February 25 (two Brief description of amendments: The Date of issuance: April 15, 2016.
letters), May 7, October 15, and amendments adopt Technical Effective date: As of its date of
December 31, 2015; and January 28, Specification Task Force change number issuance and shall be implemented
2016, provided additional information 523, Revision 2, ‘‘Generic Letter 2008– within 90 days of issuance.
01, Managing Gas Accumulation,’’ for Amendment No.: 215. A publicly-
that clarified the application, did not
the Hatch Nuclear Plant, Unit Nos 1 and available version is in ADAMS under
expand the scope of the application as
2, technical specifications. The change Accession No. ML16081A194;
originally noticed, and did not change
revised or added surveillance documents related to this amendment
the staff’s original proposed no
asabaliauskas on DSK3SPTVN1PROD with NOTICES
requirements to verify that the system are listed in the Safety Evaluation
significant hazards consideration
locations susceptible to gas enclosed with the amendment.
determination as published in the Renewed Facility Operating License
Federal Register. accumulation are sufficiently filled with
water and to provide allowances which No. NPF–42. The amendment revised
The Commission’s related evaluation permit performance of the verification. the Operating License and Technical
of the amendments is contained in a Date of issuance: April 14, 2016. Specifications.
Safety Evaluation dated April 14, 2016. Effective date: As of the date of Date of initial notice in Federal
No significant hazards consideration issuance and shall be implemented Register: November 24, 2015 (80 FR
comments received: No. within 120 days of issuance. 73242).
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![Federal Register / Vol. 81, No. 90 / Tuesday, May 10, 2016 / Notices 28905
The Commission’s related evaluation individual listed in the FOR FURTHER March 4, 2009, complies with the
of the amendment is contained in a INFORMATION CONTACT section of this requirements of the Atomic Energy Act
Safety Evaluation dated April 15, 2016. document. of 1954, as amended, and the NRC’s
No significant hazards consideration • NRC’s Agencywide Documents regulations.
comments received: No. Access and Management System The NRC originally intended for this
Dated at Rockville, Maryland, this 2nd day
(ADAMS): You may obtain publicly- notice to be published in the Federal
of May 2016. available documents online in the Registerimmediately following issuance
ADAMS Public Documents collection at
For the Nuclear Regulatory Commission. of the WBN, Unit 2, operating license on
http://www.nrc.gov/reading-rm/
Anne T. Boland, October 22, 2015; however, during
adams.html. To begin the search, select
Director, Division of Operating Reactor recent verification of operating license
‘‘ADAMS Public Documents’’ and then
Licensing, Office of Nuclear Reactor documentation the NRC identified that
select ‘‘Begin Web-based ADAMS
Regulation. the notice had not been forwarded to the
Search.’’ For problems with ADAMS,
[FR Doc. 2016–10949 Filed 5–9–16; 8:45 am] Office of the Federal Register for
please contact the NRC’s Public
publication as intended.
BILLING CODE 7590–01–P Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by II. Further Information
email to pdr.resource@nrc.gov. For the
NUCLEAR REGULATORY convenience of the reader, the ADAMS The NRC prepared a ‘‘Safety
COMMISSION accession numbers are provided in a Evaluation Report Related to the
[Docket No. 50–391; NRC–2008–0369] table in the ‘‘Availability of Documents’’ Operation of Watts Bar Nuclear Plant,
section of this document. Units 1 and 2’’ (NUREG–0847), that was
Issuance of Operating License and • NRC’s PDR: You may examine and published in June 1982, and
Record of Decision; Tennessee Valley purchase copies of public documents at Supplements 1 through 29 that were
Authority; Watts Bar Nuclear Plant, the NRC’s PDR, Room O1–F21, One published between September 1982 and
Unit 2 White Flint North, 11555 Rockville October 2015. In Supplements 1 through
Pike, Rockville, Maryland 20852. 20 the NRC staff concluded that WBN,
AGENCY: Nuclear Regulatory Unit 1, met all applicable regulations
FOR FURTHER INFORMATION CONTACT:
Commission. and regulatory guidance. In Supplement
Robert Schaaf, Office of Nuclear Reactor
ACTION: Operating license and record of Regulation, U.S. Nuclear Regulatory 21, the NRC staff reported on the WBN,
decision; issuance. Commission, Washington, DC 20555– Unit 2, open items remaining to be
0001; telephone: 301–415–6020; email: resolved, which were outstanding at the
SUMMARY: The U.S. Nuclear Regulatory Robert.Schaaf@nrc.gov. time that TVA deferred construction of
Commission (NRC) has issued operating WBN, Unit 2. In Supplements 22
SUPPLEMENTARY INFORMATION:
license No. NPF–96 to Tennessee Valley through 29, the NRC staff documented
Authority (TVA), the operator of Watts I. Introduction its evaluation and closure of the open
Bar Nuclear Plant (WBN), Unit 2. items in response to TVA’s updated
Notice is hereby given that the NRC
Operating license No. NPF–96 application for a license to operate WBN
issued operating license No. NPF–96 to
authorizes full power operation of WBN, Unit 2, filed on March 4, 2009. The NRC
TVA, the operator of WBN, Unit 2.
Unit 2. In addition, the NRC has staff also prepared a ‘‘Final
Operating license No. NPF–96
prepared a Record of Decision (ROD) Environmental Statement Related to the
authorizes full power operation of WBN,
that supports the NRC’s decision to Operation of Watts Bar Nuclear Plant,
Unit 2. The NRC’s ROD that supports its
issue operating license No. NPF–96. Unit 2’’ (NUREG–0498), Supplement 2,
decision to issue operating license No.
DATES: Operating license No. NPF–96 NPF–96 is available in ADAMS. The dated May 2013. NUREG–0847 and its
was effective on October 22, 2015. NRC staff’s safety analysis of TVA’s supplements and NUREG–0498,
ADDRESSES: Please refer to Docket ID application for the operating license is Supplement 2, document the
NRC–2008–0369 when contacting the documented in NUREG–0847, ‘‘Safety information reviewed and the NRC’s
NRC about the availability of Evaluation Report Related to the conclusions. The NRC also prepared a
information regarding this document. Operation of Watts Bar Nuclear Plant, ROD in accordance with the
You may obtain publicly-available Units 1 and 2’’, as supplemented Commission’s regulations to accompany
information related to this document through Supplement 29. The NRC staff’s its action on the operating license
using any of the following methods: updated assessment of the application. The ROD incorporates by
• Federal Rulemaking Web site: Go to environmental impacts of operation is reference the materials contained in
http://www.regulations.gov and search documented in NUREG–0498, ‘‘Final NUREG–0498, Supplement 2.
for Docket ID NRC–2008–0369. Address Environmental Statement Related to the III. Availability of Documents
questions about NRC dockets to Carol Operation of Watts Bar Nuclear Plant,
Gallagher; telephone: 301–415–3463; Unit 2,’’ Supplement 2. The NRC finds The documents identified in the
email: Carol.Gallagher@nrc.gov. For that the updated application for the following table are available to
technical questions, contact the operating license filed by TVA on interested persons, as indicated.
asabaliauskas on DSK3SPTVN1PROD with NOTICES
Document ADAMS accession No.
‘‘Watts Bar Nuclear Plant (WBN) Unit 2—Operating License Application ML090700378.
Update’’.
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Document Created: 2016-05-10 05:18:47
Document Modified: 2016-05-10 05:18:47
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by June 9, 2016. A request for a hearing must be filed by July 11, 2016. | |
| Contact | Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-5411, email: [email protected] | |
| FR Citation | 81 FR 28891 |
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