81 FR 13837 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 81, Issue 50 (March 15, 2016)
Federal Register Volume 81, Issue 50 (March 15, 2016)
| Page Range | 13837-13849 | |
| FR Document | 2016-05470 |
[Federal Register Volume 81, Number 50 (Tuesday, March 15, 2016)] [Notices] [Pages 13837-13849] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2016-05470] [[Page 13837]] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2016-0050] Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission. ACTION: Biweekly notice. ----------------------------------------------------------------------- SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued, from February 13, 2016, to February 29, 2016. The last biweekly notice was published on March 1, 2016. DATES: Comments must be filed by April 14, 2016. A request for a hearing must be filed by May 16, 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-2016-0050. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0050 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-2016-0050. 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 of this document. NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2016-0050, facility name, unit number(s), application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov, as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a 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 [[Page 13838]] 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 proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/ petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. 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 May 16, 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 May 16, 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 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [[Page 13839]] [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 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. 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). 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. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station (CNS), Units 1 and 2, York County, South Carolina Date of amendment request: January 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16026A048. Description of amendment request: The proposed amendments would modify the Renewed Facility Operating Licenses and Technical Specifications (TS) for CNS, Units 1 and 2. Specifically, the proposed amendments request to revise TS 5.5.2, ``Containment Leakage Rate Testing Program,'' to allow an increase in the existing Type A Integrated Leakage Rate Test (ILRT) program test interval from 10 years to 15 [[Page 13840]] years in accordance with Nuclear Energy Institute (NEI) Topical Report NE1 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J,'' and the conditions and limitations specified in NEI 94-01, Revision 2-A; adoption of an extension of the containment isolation valve leakage testing (Type C) frequency from the 60 months currently permitted by 10 CFR part 50, appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94- 01, Revision 3-A; adoption of the use of ANSI/ANS 56.8-2002, ``Containment System Leakage Testing Requirements''; and adoption of a more conservative grace interval of 9 months for Type A, Type B, and Type C leakage tests in accordance with NEI 94-01, Revision 3-A. The proposed amendments also request the following administrative changes: Deletion of the information regarding the performance of containment visual inspections as required by Regulatory Position C.3, as the containment inspections are addressed in TS Surveillance Requirement 3.6.1.1, deletion of the information regarding the performance of the next CNS, Unit 1, Type A test no later than November 13, 2015, and the next CNS, Unit 2, Type A test no later than February 6, 2008, as both Type A tests have already occurred. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below with NRC edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment to the Technical Specifications (TS) involves the extension of the Catawba Nuclear Station (CNS) Type A containment integrated leak rate test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. 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. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured, as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.026 person-rem/year. EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of [less than or equal to] 1.0 person-rem per year, or [less than or equal to] 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CNS Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and; (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes an exception previously granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT test frequency for CNS. This exception was for activities that have already taken place; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the units are operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the CNS Type A containment integrated leak rate test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. 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. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. The proposed amendment also deletes an exception previously granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT test frequency for CNS. This exception was for activities that have already taken; therefore, their deletion is solely an administrative action that does not result in any change in how the units are 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 the margin of safety? Response: No. The proposed amendment to TS 5.5.2 involves the extension of the CNS Type A containment integrated leak rate test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. 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. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests, and Type C tests for CNS. The proposed surveillance interval extension is bounded by the 15-year ILRT interval, and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, TS and the Maintenance Rule serve to provide a high [[Page 13841]] 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 an exception previously granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT test frequency for CNS. This exception was for activities that have already taken place; therefore, their deletion is solely an administrative action and does not change how the units are operated and maintained. Thus, there is no reduction in any margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC 28202. NRC Branch Chief: Michael T. Markley. Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of amendment request: November 19, 2015. A publicly-available version is in ADAMS under Accession No. ML15323A085. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to allow the extension of the Type A containment test interval to 15 years and the extension of the Type B and Type C test intervals for selected components to 120 months and 75 months, respectively. The proposed amendment also deletes from the TSs an already implemented one-time extension of the Type A test frequency. 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 H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Type A containment test interval to 15 years, the extension of the Type B test intervals to 120 months for selected components, and the extension of the Type C test interval to 75 months for selected components. 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 B test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years would be extended on a performance basis to no longer than 120 months. The current Type C test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years 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 extensions do 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. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured, as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.020 person-rem [roentgen equivalent man]/year. The Electric Power Research Institute (EPRI) 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 integrated leak rate test (ILRT) intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The HBRSEP2 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 the American Society of Mechanical Engineers (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. The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components and the extension of the Type C test interval to 75 months for selected components. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not 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 5.5.16 involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components 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 [[Page 13842]] 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, Type B tests and Type C tests for HBRSEP2. The proposed surveillance interval extension is bounded by the 15-year ILRT interval, the 120-month Type B interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Types 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, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types 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, Type B and Type C test intervals. The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not 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, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202. NRC Branch Chief: Benjamin G. Beasley. Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station (LGS), Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: January 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16015A316. Description of amendment request: The amendments would reduce the reactor vessel steam dome pressure associated with the Technical Specification (TS) Safety Limits (SLs) specified in TS 2.1.1 and TS 2.1.2. The amendments would also revise the setpoint and allowable value for the main steam line low pressure isolation function in TS Table 3.3.2-2. The proposed changes address a 10 CFR part 21 issue concerning the potential to violate the SLs limits during a pressure regulator failure maximum demand (open) transient. 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 do not involve a significant increase in the probability or consequences of an accident previously evaluated because decreasing the reactor vessel steam dome pressure in TS Safety Limits 2.1.1 and 2.1.2 for reactor thermal power ranges and increasing the trip setpoint and allowable value for the main steam line low pressure isolation effectively expands the validity range for GEXL critical power correlation and the calculation of the minimum critical power ratio. The critical power ratio rises during the pressure reduction following the scram that terminates the Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient. The reduction in the reactor vessel steam dome pressure value in the SL and the increase in the trip setpoint and the allowable value for the main steam line low pressure isolation provides adequate margin to accommodate the pressure reduction during the PRFO transient within the revised TS limit. The proposed changes do not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed changes are in accordance with an NRC approved critical power correlation methodology and do not adversely affect accident initiators or precursors. The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the applicable acceptance limits. The proposed changes are consistent with the safety analysis and resultant consequences. 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 do not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed reduction in the reactor vessel steam dome pressure value in the safety limit in conjunction with the increase in the trip setpoint and the allowable value for the main steam line low pressure isolation reflects a wider range of applicability for the GEXL critical power correlation which is approved by the NRC for both GE14 and GNF2 fuel types in [the] LGS reactor cores. In addition, no new failure modes are being introduced. There are no changes in the method by which any plant systems perform a safety function. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed changes do not introduce any new accident precursors, nor do they involve any changes in the methods governing normal plant operation. The proposed changes do not alter the outcome of the safety analysis. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. Evaluation of the 10 CFR part 21 condition by General Electric determined that, since the critical power ratio improves during the PRFO transient, there is no impact on the fuel safety margin, and therefore, there is no challenge to fuel cladding integrity. The proposed changes do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The proposed changes are consistent with the applicable NRC approved critical power correlation for the fuel designs in use at LGS. The proposed changes do not alter the manner in which the safety limits are determined. The reduction in value of the reactor vessel steam dome pressure safety limit and the increase in the trip setpoint and allowable value for the main steam line low pressure isolation provides adequate margin to accommodate the pressure reduction during the PRFO transient within the revised TS limit. Therefore, the proposed changes do not involve a significant reduction in any 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. [[Page 13843]] Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Douglas A. Broaddus. FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, Ohio Date of amendment request: February 17, 2016. A publicly-available version is in ADAMS under Accession No. ML16049A513. Description of amendment request: The licensee proposes to change the emergency plan for DBNPS, Unit No. 1, by revising the emergency action level (EAL) scheme based on the Nuclear Energy institute's (NEl's) guidance in NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors.'' The NEI 99-01, Revision 6, was endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession No. ML12346A463). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to DBNPS's EAL scheme to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. The proposed changes do not alter any of the requirements of the technical specifications. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes do not impact the ability of structures, systems, or components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits. 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 changes to DBNPS's EAL scheme to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. DBNPS functions will continue to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes to DBNPS's EAL scheme to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. Margins of safety are unaffected by the proposed changes. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed EAL scheme change. The proposed change does not affect the technical specifications. There are no changes to environmental conditions of any of the SSC or the manner in which any SSC is operated. The applicable requirements of 10 CFR 50.47 and 10 CFR part 50, appendix E will continue to be met. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. Acting NRC Branch Chief: Justin C. Poole. Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: January 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16034A032. Description of amendment request: The proposed amendment would modify technical specification (TS) requirements to address Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'' as described in the Technical Specification Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.'' 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 or adds SRs [Surveillance Requirements] that require verification that the ECCS [Emergency Core Cooling System], RHR [Residual Heat Removal] System, and the Containment Spray (CTS) System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the RHR System, and the CTS System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. 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. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the RHR System, and the CTS System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect [[Page 13844]] any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. 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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: David J. Wrona. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of amendment request: January 21, 2106. A publicly-available version is in ADAMS under Accession No. ML16021A067. Description of amendment request: The amendments would revise or add 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. The amendments would revise Technical Specification (TS) 3.4.6, ``RCS [Reactor Coolant System] Loops--MODE 4''; TS 3.4.7, ``RCS Loops--MODE 5, Loops Filled''; TS 3.4.8, ``RCS Loops--MODE 5, Loops Not Filled''; TS 3.5.2, ``ECCS [Emergency Core Cooling System]--Operating''; TS 3.6.6, ``Containment Spray and Cooling Systems''; TS 3.9.5, ``RHR [Residual Heat Removal] and Coolant Circulation--High Water Level''; and TS 3.9.6, ``RHR and Containment Circulation--Low Water Level.'' The proposed amendments would modify TS requirements to address Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'' as described in Technical Specification Task Force TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.'' 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 or adds Surveillance Requirement(s) (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Residual Heat Removal (RHR) System, and the Containment Spray (CS) System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased. 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 or adds SRs that require verification that the ECCS, RHR System, and CS System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. 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. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the RHR System, and the CS System are not rendered inoperable due to accumulated gas, and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs, and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. 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. South Carolina Electric & 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: January 19, 2016. A publicly-available version is in ADAMS under Accession No. ML16019A403. Description of amendment request: The requested amendment proposes to depart from Tier 2* information in the Updated Final Safety Analysis Report (which includes the plant-specific design control document Tier 2 information) related to the construction methods used for the composite floors and roof of the auxiliary building. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The design functions of the nuclear island structures are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the nuclear island. The nuclear island structures are structurally designed to meet seismic Category I requirements as defined in Regulatory Guide 1.29. The use of ACI 349 and AISC N690 provides criteria for the design, qualification, fabrication, and inspection of composite steel beam floors and roof in the auxiliary building. These structures continue to meet the applicable portions of ACI 349 and AISC N690. The proposed change does not have an adverse impact on the response of the nuclear island structures to safe shutdown earthquake ground motions or loads due to [[Page 13845]] anticipated transients or postulated accident conditions. The change does not impact the support, design, or operation of mechanical and fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor does the change described create any new accident precursors. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises the description of the construction of composite steel beam floors and roof in the auxiliary building. The proposed change does not change the design function, support, design, or operation of mechanical and fluid systems. The proposed change does not result in a new failure mechanism for the pertinent structures or new accident precursors. As a result, the design function of the structures is not adversely affected by the proposed change. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change is consistent with ACI 349 and AISC N690. The design and construction of the auxiliary building floors and roof remain in conformance with the requirements in ACI 349 and AISC N690. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. Acting NRC Branch Chief: John McKirgan. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50- 529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona Date of amendment request: February 27, 2015, as supplemented by letter dated January 19, 2016. Brief description of amendments: The amendments revised Technical Specification (TS) 1.3, ``Completion Times''; TS 3.7.5, ``Auxiliary Feedwater (AFW) System''; TS 3.8.1, ``AC [Alternating Current] Sources--Operating''; and TS 3.8.9, ``Distribution Systems-- Operating''; to remove the second Completion Times. The amendment also revised Example 1.3-3 in TS 1.3, ``Completion Times,'' by adding a discussion of administrative controls to combinations of conditions to ensure that the Completion Times for those conditions are not inappropriately extended. The changes are consistent with the NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-439-A, Revision 2, ``Eliminate Second Completion Times Limiting Time From Discovery of Failure to Meet an LCO [Limiting Condition of Operation],'' dated June 20, 2005. Date of issuance: February 19, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment Nos.: Unit 1--197; Unit 2--197; Unit 3--197. A publicly- available version is in ADAMS under Accession No. ML16004A013; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Operating Licenses and TSs. Date of initial notice in Federal Register: May 12, 2015 (80 FR 27195). The supplement dated January 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 February 19, 2016. No significant hazards consideration comments received: No. Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendment request: February 19, 2015, as supplemented by letter dated November 5, 2015. Description of amendment request: The amendments revised (1) technical specifications (TSs) by replacing AREVA Topical Report ANP- 10298PA, ``ACE/ATRIUM 10XM Critical Power Correlation,'' Revision 0, March 2010, with Revision 1, March 2014, of the same topical report; and (2) Appendix B, ``Additional Conditions,'' by removing the license condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2, respectively. Date of issuance: February 9, 2016. Effective date: Once approved, the Unit 1 amendment shall be implemented prior to start-up. from the 2016 Unit 1 refueling outage, and the Unit 2 amendment shall be implemented prior to start-up from the 2017 Unit 2 refueling outage. Amendment Nos.: 269 and 297. A publicly-available version is in ADAMS under Accession No. ML16019A029; [[Page 13846]] documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Facility Operating License Nos. DPR-71, and DPR-62: Amendments revised the renewed facility operating licenses and TSs. Date of initial notice in Federal Register: April 28, 2015 (80 FR 23603). The supplemental letter dated November 5, 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 an SE dated February 9, 2016. No significant hazards consideration comments received: No. Energy Northwest, Docket No. 50-397, Columbia Generating Station (CGS), Benton County, Washington Date of amendment request: September 2, 2015. Brief description of amendment: The amendment revised the Technical Specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9. The LCO allows a delay time for entering a supported system TS, when the inoperability is solely due to an unavailable barrier, if the risk is assessed and managed. The change is consistent with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification (STS) Change TSTF-427, Revision 2, ``Allowance for Non Technical Specification Barrier Degradation on Supported System OPERABILITY'' (ADAMS Accession No. ML061240055). The availability of this TS improvement was published in the Federal Register on October 3, 2006 (71 FR 58444), as part of the Consolidated Line Item Improvement Process. Additionally, LCO 3.0.8 has been revised to replace the term ``train'' with ``division'' to be consistent with CGS's TS definition of ``OPERABLE-OPERABILITY'' and the terminology used in Section 1.3, ``Completion Times,'' of the CGS TS. Date of issuance: February 16, 2016. Effective date: As of its date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 237. A publicly-available version is in ADAMS under Accession No. ML16020A031; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-21: The amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: October 27, 2015 (80 FR 65811). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 16, 2016. No significant hazards consideration comments received: No. Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas Nuclear One (ANO), Units 1 and 2, Pope County, Arkansas Date of amendment request: May 20, 2015. Brief description of amendments: The amendments revised the full implementation date (Milestone 8) of the ANO, Units 1 and 2, Cyber Security Plan, and revised the associated physical protection license conditions for each renewed facility operating license. Date of issuance: February 24, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1--255; Unit 2--303. A publicly-available version is in ADAMS under Accession No. ML16027A109; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-51 and NPF-6: The amendments revised the renewed facility operating licenses. Date of initial notice in Federal Register: June 23, 2015 (80 FR 35982). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 24, 2016. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York Date of amendment request: December 9, 2014, as supplemented by two letters dated May 20, 2015, and letters dated June 8, 2015, and June 29, 2015. Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.14, ``Containment Leakage Rate Testing Program,'' to extend the frequency of the containment integrated leak rate test from once every 10 years to once every 15 years on a permanent basis. Date of issuance: February 23, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 283. A publicly-available version is in ADAMS under Accession No. ML15349A794; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. DPR-26: The amendment revised the Facility Operating License and the Technical Specifications. Date of initial notice in Federal Register: March 17, 2015 (80 FR 13905). The supplemental letters dated May 20, 2015; June 8, 2015; and June 29, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 23, 2016. No significant hazards consideration comments received: Yes. The comments submitted by the State of New York on November 20, 2015, are addressed in the NRC staff's Safety Evaluation dated February 23, 2016. Entergy Operations, Inc.; System Energy Resources, Inc.; South Mississippi Electric Power Association; and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: May 27, 2015, as supplemented by letters dated October 28, 2015, and December 10, 2015. Brief description of amendment: The amendment revised the GGNS Technical Specifications (TSs) to allow for a permanent extension of the Type C leakage rate testing frequency and reduction of the Type B and Type C grace intervals that are required by GGNS TS 5.5.12, ``10 CFR part 50, appendix J, Testing Program,'' by including a reference to Nuclear Energy Institute (NEI) Topical Report, NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J,'' dated July 2012. In addition, the amendment changed Surveillance Requirement (SR) 3.6.5.1.1 by deleting the information regarding the performance of the last Type A test that has already occurred. This amendment [[Page 13847]] does not alter the Type A testing frequencies nor any other requirements as specified in the existing GGNS TS. Date of issuance: February 17, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No: 209. A publicly-available version is in ADAMS under Accession No. ML16011A247; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-29: The amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: September 29, 2015 (80 FR 58516). The supplemental letters dated October 28, 2015, and December 10, 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 February 17, 2016. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: March 25, 2014, as supplemented by letters dated October 7, 2014, and August 24, 2015. Brief description of amendment: The amendment modifies the Technical Specifications (TSs) by relocating certain surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program, using probabilistic risk guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.'' The changes are consistent with the approved Technical Specification Task Force (TSTF) Traveler TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b.'' Date of issuance: February 23, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment No.: 171. A publicly-available version is in ADAMS under Accession No. ML15307A349; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-58: Amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: September 16, 2014 (79 FR 55512). The supplemental letters dated October 7, 2014, and August 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 February 23, 2016. No significant hazards consideration comments received: No. Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: October 12, 2015. Brief description of amendments: The amendments revised the Technical Specifications (TSs) related to facility staff qualifications for licensed operators. Date of issuance: February 25, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos: 268 and 263. A publicly-available version is in ADAMS under Accession No. ML16008B072; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: December 22, 2015 (80 FR 79620). The Commission's related evaluation of the amendments is contained in an SE dated February 25, 2016. No significant hazards consideration comments received: No. Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: August 20, 2015, as supplemented by letter dated January 27, 2016. Brief description of amendment: The amendment made administrative changes to update personnel and committee titles in the Technical Specifications (TSs), deleted outdated or completed additional actions contained in Appendix B, Additional Conditions, of the license, and relocated the definition of Process Control Program from the TSs to the Updated Safety Analysis Report. Date of issuance: February 23, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 286. A publicly-available version is in ADAMS under Accession No. ML15307A013; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-40: The amendment revised the license, TSs, and Appendix B to the license. Date of initial notice in Federal Register: October 13, 2015 (80 FR 61486). The supplemental letter dated January 27, 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 February 23, 2016. No significant hazards consideration comments received: No. 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 amendment request: February 25, 2015, as supplemented by letter dated July 8, 2015. Brief description of amendments: The amendments incorporated into the licensing basis an analysis of pressurizer reaching a water-solid (filled) condition associated with the main feedwater pipe rupture accident summarized in the Updated Final Safety Analysis Report (UFSAR), Section 15.4.2.2. Further, the amendments involved the addition of time critical operator actions and modifications of the PG&E Design Class I backup nitrogen accumulators, which are credited in the new pressurizer filling analysis. Date of issuance: February 19, 2016. Effective date: As of its date of issuance and shall be implemented within 90 days following PG&E implementation of Design Class 1 backup nitrogen accumulator modifications, planned for the nineteenth refueling outage 2R19 for Unit No. 2. [[Page 13848]] Amendment Nos.: Unit 1--223; Unit 2--225. A publicly-available version is in ADAMS under Accession No. ML16032A006; 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 UFSAR. Date of initial notice in Federal Register: April 28, 2015 (80 FR 23605). The supplemental letter dated July 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 amendments is contained in a Safety Evaluation dated February 19, 2016 No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50- 364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request: May 12, 2015, as supplemented by letters dated September 15, 2015; November 25, 2015; and January 28, 2016. Brief description of amendments: The amendments revised and added Surveillance Requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances that permit performance of the verification. The changes are consistent with Technical Specification Trask Force Traveler (TSTF)-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.'' Date of issuance: February 26, 2016. Effective date: As of its date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: Unit 1--200, Unit 2--196. A publicly-available version is in ADAMS under Accession No. ML15345A131, documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: June 23, 2015 (80 FR 35982). The supplemental letters dated September 15, 2015; November 25, 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 February 26, 2016. No significant hazards consideration comments received: No. South Carolina Electric & 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: May 18, 2015. Description of amendment: The amendment authorizes changes to the VCSNS, Units 2 and 3 Updated Final Safety Analysis Report by revising the Radiation Emergency Plan to expand the plume exposure pathway emergency planning zone (EPZ) boundary. The Evacuation Time Estimates Study and Alert and Notification System Design Report have also been revised to encompass the expanded EPZ boundary. Date of issuance: February 5, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 41. A publicly-available version is in ADAMS under Accession No. ML15292A404; documents related to this amendment are listed in a Safety Evaluation enclosed with the amendment. Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses. Date of initial notice in Federal Register: September 29, 2015 (80 FR 585120). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 5, 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: August 21, 2015, as supplemented by letters dated September 17, 2015, and September 22, 2015. Brief description of amendment: The amendment authorized changes to the VEGP, Units 3 and 4, Updated Final Safety Analysis Report in the form of departures from the incorporated plant-specific Design Control Document Tier 2* and associated Tier 2 information. The changes are to demonstrate that the capacity of mechanical couplers welded to structural steel embed plates required by American Concrete Institute (ACI) 349-01, ``Code Requirements for Nuclear Safety Related Concrete Structures,'' is satisfied using American Institute of Steel Construction (AISC) N690-1994, ``Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities,'' analysis and testing provisions. Date of issuance: November 5, 2015. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 40. A publicly-available version is in ADAMS under Accession No. ML15287A031; 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: September 3, 2015 (80 FR 53340). The supplemental letters dated September 17, 2015, and September 22, 2015, provided additional information that did not change the scope or the conclusions of the no significant hazards determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 5, 2015. No significant hazards consideration comments received: No. Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri Date of amendment request: May 8, 2015, as supplemented by letter dated November 9, 2015. Brief description of amendment: The amendment revised Technical Specifications (TSs) 2.1.1.1 and 5.6.5 to adopt the NRC-approved methodologies of Westinghouse Commercial Atomic Power reports (WCAP)- 14483-A, ``Generic Method for Expanded Core Operating Limits Report,'' and WCAP-14565-P-A, Addendum 2-P-A, ``VIPRE-1 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,'' respectively. The change in TS 2.1.1.1 would provide the departure from nucleate boiling ratio in a form that reduces the need for cycle-specific license amendments, and the change in TS 5.6.5 adds an NRC-approved methodology for determining core operating limits. [[Page 13849]] Date of issuance: February 29, 2016. Effective date: As of its date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 216. A publicly-available version is in ADAMS under Accession No. ML16020A516; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-30: The amendment revised the operating license and TSs. Date of initial notice in Federal Register: July 7, 2015 (80 FR 38763). The supplemental letter dated November 9, 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 February 29, 2016. No significant hazards consideration comments received: No. Virginia Electric and Power Company, Docket No. 50-339, North Anna Power Station, Unit No. 2, Louisa County, Virginia Date of amendment request: May 22, 2015. As supplemented by letter dated October 13, 2015. Brief description of amendment: The amendment revised the Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' to remove the limitation in Note 1 that the surveillance is only applicable to Unit 1. Revised Surveillance Requirement (SR) 3.8.1.8 is applicable to both units. Date of issuance: February 22, 2016. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment No.: 260. A publicly-available version is in ADAMS under Accession No. ML16013A444. Documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-7: Amendment revised the Facility Operating License and Technical Specification. Date of initial notice in Federal Register: July 21, 2015 (80 FR 43131). The supplement letter dated October 13, 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. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 22, 2016. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 2nd day of March 2016. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2016-05470 Filed 3-14-16; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices 13837
NUCLEAR REGULATORY SUPPLEMENTARY INFORMATION: II. Notice of Consideration of Issuance
COMMISSION of Amendments to Facility Operating
I. Obtaining Information and
Licenses and Combined Licenses and
[NRC–2016–0050] Submitting Comments
Proposed No Significant Hazards
Biweekly Notice; Applications and A. Obtaining Information Consideration Determination
Amendments to Facility Operating Please refer to Docket ID NRC–2016– The Commission has made a
Licenses and Combined Licenses 0050 when contacting the NRC about proposed determination that the
Involving No Significant Hazards the availability of information for this following amendment requests involve
Considerations action. You may obtain publicly- no significant hazards consideration.
available information related to this Under the Commission’s regulations in
AGENCY: Nuclear Regulatory action by any of the following methods:
Commission. § 50.92 of title 10 of the Code of Federal
• Federal Rulemaking Web site: Go to Regulations (10 CFR), this means that
ACTION: Biweekly notice. http://www.regulations.gov and search operation of the facility in accordance
SUMMARY: Pursuant to Section 189a.(2) for Docket ID NRC–2016–0050. with the proposed amendment would
• NRC’s Agencywide Documents not (1) involve a significant increase in
of the Atomic Energy Act of 1954, as
Access and Management System the probability or consequences of an
amended (the Act), the U.S. Nuclear
(ADAMS): You may obtain publicly- accident previously evaluated, or (2)
Regulatory Commission (NRC) is
available documents online in the create the possibility of a new or
publishing this regular biweekly notice.
ADAMS Public Documents collection at different kind of accident from any
The Act requires the Commission to
http://www.nrc.gov/reading-rm/ accident previously evaluated; or (3)
publish notice of any amendments
adams.html. To begin the search, select involve a significant reduction in a
issued, or proposed to be issued, and
‘‘ADAMS Public Documents’’ and then margin of safety. The basis for this
grants the Commission the authority to
select ‘‘Begin Web-based ADAMS proposed determination for each
issue and make immediately effective
Search.’’ For problems with ADAMS, amendment request is shown below.
any amendment to an operating license
please contact the NRC’s Public The Commission is seeking public
or combined license, as applicable,
Document Room (PDR) reference staff at comments on this proposed
upon a determination by the
1–800–397–4209, 301–415–4737, or by determination. Any comments received
Commission that such amendment
email to pdr.resource@nrc.gov. The within 30 days after the date of
involves no significant hazards
ADAMS accession number for each publication of this notice will be
consideration, notwithstanding the
document referenced (if it is available in considered in making any final
pendency before the Commission of a
ADAMS) is provided the first time that
request for a hearing from any person. determination.
This biweekly notice includes all it is mentioned in the SUPPLEMENTARY Normally, the Commission will not
INFORMATION section of this document.
notices of amendments issued, or issue the amendment until the
• NRC’s PDR: You may examine and
proposed to be issued, from February expiration of 60 days after the date of
purchase copies of public documents at
13, 2016, to February 29, 2016. The last publication of this notice. The
the NRC’s PDR, Room O1–F21, One
biweekly notice was published on Commission may issue the license
White Flint North, 11555 Rockville
March 1, 2016. amendment before expiration of the 60-
Pike, Rockville, Maryland 20852.
DATES: Comments must be filed by April day period provided that its final
14, 2016. A request for a hearing must B. Submitting Comments determination is that the amendment
be filed by May 16, 2016. Please include Docket ID NRC–2016– involves no significant hazards
ADDRESSES: You may submit comments 0050, facility name, unit number(s), consideration. In addition, the
by any of the following methods (unless application date, and subject in your Commission may issue the amendment
this document describes a different comment submission. prior to the expiration of the 30-day
method for submitting comments on a The NRC cautions you not to include comment period should circumstances
specific subject): identifying or contact information that change during the 30-day comment
• Federal Rulemaking Web site: Go to you do not want to be publicly period such that failure to act in a
http://www.regulations.gov and search disclosed in your comment submission. timely way would result, for example in
for Docket ID NRC–2016–0050. Address The NRC posts all comment derating or shutdown of the facility.
questions about NRC dockets to Carol submissions at http:// Should the Commission take action
Gallagher; telephone: 301–415–3463; www.regulations.gov, as well as entering prior to the expiration of either the
email: Carol.Gallagher@nrc.gov. the comment submissions into ADAMS. comment period or the notice period, it
• Mail comments to: Cindy Bladey, The NRC does not routinely edit will publish in the Federal Register a
Office of Administration, Mail Stop: comment submissions to remove notice of issuance. Should the
OWFN–12–H08, U.S. Nuclear identifying or contact information. Commission make a final No Significant
Regulatory Commission, Washington, If you are requesting or aggregating Hazards Consideration Determination,
DC 20555–0001. comments from other persons for any hearing will take place after
For additional direction on obtaining submission to the NRC, then you should issuance. The Commission expects that
information and submitting comments, inform those persons not to include the need to take this action will occur
see ‘‘Obtaining Information and identifying or contact information that very infrequently.
Submitting Comments’’ in the they do not want to be publicly
asabaliauskas on DSK3SPTVN1PROD with NOTICES
SUPPLEMENTARY INFORMATION section of disclosed in their comment submission. A. Opportunity To Request a Hearing
this document. Your request should state that the NRC and Petition for Leave To Intervene
FOR FURTHER INFORMATION CONTACT: does not routinely edit comment Within 60 days after the date of
Lynn Ronewicz, Office of Nuclear submissions to remove such information publication of this notice, any person(s)
Reactor Regulation, U.S. Nuclear before making the comment whose interest may be affected by this
Regulatory Commission, Washington DC submissions available to the public or action may file a request for a hearing
20555–0001; telephone: 301–415–1927, entering the comment submissions into and a petition to intervene with respect
email: Lynn.Ronewicz@nrc.gov. ADAMS. to issuance of the amendment to the
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![13840 Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices
years in accordance with Nuclear uncontrolled release of radioactivity to the containment integrated leak rate test interval
Energy Institute (NEI) Topical Report environment for postulated accidents. The to 15 years and the extension of the Type C
NE1 94–01, Revision 3–A, ‘‘Industry containment and the testing requirements test interval to 75 months for selected
invoked to periodically demonstrate the components.
Guideline for Implementing The current Type A test interval of 120
integrity of the containment exist to ensure
Performance-Based Option of 10 CFR the plant’s ability to mitigate the months (10 years) would be extended on a
part 50, appendix J,’’ and the conditions consequences of an accident, and do not permanent basis to no longer than 15 years
and limitations specified in NEI 94–01, involve the prevention or identification of from the last Type A test. The current Type
Revision 2–A; adoption of an extension any precursors of an accident. The change in C test interval of 60 months for selected
of the containment isolation valve dose risk for changing the Type A test components would be extended on a
leakage testing (Type C) frequency from frequency from three-per-ten years to once- performance basis to no longer than 75
the 60 months currently permitted by 10 per-fifteen years, measured, as an increase to months. The containment and the testing
the total integrated plant risk for those requirements to periodically demonstrate the
CFR part 50, appendix J, Option B, to a
accident sequences influenced by Type A integrity of the containment exist to ensure
75-month frequency for Type C leakage testing, is 0.026 person-rem/year. EPRI the plant’s ability to mitigate the
rate testing of selected components, in Report No. 1009325, Revision 2–A states that consequences of an accident do not involve
accordance with NEI 94–01, Revision 3– a very small population dose is defined as an any accident precursors or initiators. The
A; adoption of the use of ANSI/ANS increase of [less than or equal to] 1.0 person- proposed change does not involve a physical
56.8–2002, ‘‘Containment System rem per year, or [less than or equal to] 1% change to the plant (i.e., no new or different
Leakage Testing Requirements’’; and of the total population dose, whichever is type of equipment will be installed) or a
adoption of a more conservative grace less restrictive for the risk impact assessment change to the manner in which the plant is
of the extended ILRT intervals. Therefore, operated or controlled.
interval of 9 months for Type A, Type
this proposed extension does not involve a The proposed amendment also deletes an
B, and Type C leakage tests in significant increase in the probability of an exception previously granted to allow one-
accordance with NEI 94–01, Revision 3– accident previously evaluated. time extensions of the Unit 1 and Unit 2 ILRT
A. The proposed amendments also As documented in NUREG–1493, Type B test frequency for CNS. This exception was
request the following administrative and C tests have identified a very large for activities that have already taken;
changes: Deletion of the information percentage of containment leakage paths, and therefore, their deletion is solely an
regarding the performance of the percentage of containment leakage paths administrative action that does not result in
containment visual inspections as that are detected only by Type A testing is any change in how the units are operated.
required by Regulatory Position C.3, as very small. The CNS Type A test history Therefore, the proposed change does not
supports this conclusion. create the possibility of a new or different
the containment inspections are
The integrity of the containment is subject kind of accident from any previously
addressed in TS Surveillance to two types of failure mechanisms that can evaluated.
Requirement 3.6.1.1, deletion of the be categorized as: (1) Activity based, and; (2) 3. Does the proposed change involve a
information regarding the performance time based. Activity based failure significant reduction in the margin of safety?
of the next CNS, Unit 1, Type A test no mechanisms are defined as degradation due Response: No.
later than November 13, 2015, and the to system and/or component modifications or The proposed amendment to TS 5.5.2
next CNS, Unit 2, Type A test no later maintenance. Local leak rate test involves the extension of the CNS Type A
than February 6, 2008, as both Type A requirements and administrative controls containment integrated leak rate test interval
tests have already occurred. such as configuration management and to 15 years and the extension of the Type C
Basis for proposed no significant procedural requirements for system test interval to 75 months for selected
restoration ensure that containment integrity components. The current Type A test interval
hazards consideration determination:
is not degraded by plant modifications or of 120 months (10 years) would be extended
As required by 10 CFR 50.91(a), the maintenance activities. The design and on a permanent basis to no longer than 15
licensee has provided its analysis of the construction requirements of the years from the last Type A test. The current
issue of no significant hazards containment combined with the containment Type C test interval of 60 months for selected
consideration, which is presented below inspections performed in accordance with components would be extended on a
with NRC edits in square brackets: ASME Section Xl, the Maintenance Rule, and performance basis to no longer than 75
TS requirements serve to provide a high months. This amendment does not alter the
1. Does the proposed amendment involve
degree of assurance that the containment manner in which safety limits, limiting safety
a significant increase in the probability or
consequences of an accident previously would not degrade in a manner that is system set points, or limiting conditions for
evaluated? detectable only by a Type A test. Based on operation are determined. The specific
Response: No. the above, the proposed extensions do not requirements and conditions of the TS
The proposed amendment to the Technical significantly increase the consequences of an Containment Leak Rate Testing Program exist
Specifications (TS) involves the extension of accident previously evaluated. to ensure that the degree of containment
the Catawba Nuclear Station (CNS) Type A The proposed amendment also deletes an structural integrity and leak tightness that is
containment integrated leak rate test interval exception previously granted to allow one- considered in the plant safety analysis is
to 15 years and the extension of the Type C time extensions of the Unit 1 and Unit 2 ILRT maintained. The overall containment leak
test interval to 75 months for selected test frequency for CNS. This exception was rate limit specified by TS is maintained.
components. The current Type A test interval for activities that have already taken place; The proposed change involves only the
of 120 months (10 years) would be extended therefore, their deletion is solely an extension of the interval between Type A
on a permanent basis to no longer than 15 administrative action that has no effect on containment leak rate tests, and Type C tests
years from the last Type A test. The current any component and no impact on how the for CNS. The proposed surveillance interval
Type C test interval of 60 months for selected units are operated. extension is bounded by the 15-year ILRT
components would be extended on a Therefore, the proposed change does not interval, and the 75-month Type C test
performance basis to no longer than 75 result in a significant increase in the interval currently authorized within NEI 94–
asabaliauskas on DSK3SPTVN1PROD with NOTICES
months. Extensions of up to nine months probability or consequences of an accident 01, Revision 3–A. Industry experience
(total maximum interval of 84 months for previously evaluated. supports the conclusion that Type B and C
Type C tests) are permissible only for non- 2. Does the proposed change create the testing detects a large percentage of
routine emergent conditions. The proposed possibility of a new or different kind of containment leakage paths and that the
extension does not involve either a physical accident from any accident previously percentage of containment leakage paths that
change to the plant or a change in the manner evaluated? are detected only by Type A testing is small.
in which the plant is operated or controlled. Response: No. The containment inspections performed in
The containment is designed to provide an The proposed amendment to the TS accordance with ASME Section Xl, TS and
essentially leak tight barrier against the involves the extension of the CNS Type A the Maintenance Rule serve to provide a high
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![Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices 13841
degree of assurance that the containment consequences of an accident previously restoration ensure that containment integrity
would not degrade in a manner that is evaluated? is not degraded by plant modifications or
detectable only by Type A testing. The Response: No. maintenance activities. The design and
combination of these factors ensures that the The proposed amendment to the Technical construction requirements of the
margin of safety in the plant safety analysis Specifications (TS) involves the extension of containment combined with the containment
is maintained. The design, operation, testing the H. B. Robinson Steam Electric Plant Unit inspections performed in accordance with
methods and acceptance criteria for Type A, No. 2 (HBRSEP2) Type A containment test the American Society of Mechanical
B, and C containment leakage tests specified interval to 15 years, the extension of the Type Engineers (ASME) Section XI, the
in applicable codes and standards would B test intervals to 120 months for selected Maintenance Rule, and TS requirements
continue to be met, with the acceptance of components, and the extension of the Type serve to provide a high degree of assurance
this proposed change, since these are not C test interval to 75 months for selected that the containment would not degrade in a
affected by changes to the Type A, and Type components. The current Type A test interval manner that is detectable only by a Type A
C test intervals. of 120 months (10 years) would be extended test. Based on the above, the proposed
The proposed amendment also deletes an on a permanent basis to no longer than 15 extensions do not significantly increase the
exception previously granted to allow one- years from the last Type A test. The current consequences of an accident previously
time extensions of the Unit 1 and Unit 2 ILRT Type B test interval of each reactor shutdown evaluated.
test frequency for CNS. This exception was for refueling but in no case at intervals The proposed amendment also deletes an
for activities that have already taken place; greater than 2 years would be extended on a exception previously granted to allow one-
therefore, their deletion is solely an performance basis to no longer than 120 time extension of the ILRT test frequency for
administrative action and does not change months. The current Type C test interval of HBRSEP2. This exception was for an activity
how the units are operated and maintained. each reactor shutdown for refueling but in no that has already taken place so the deletion
Thus, there is no reduction in any margin of case at intervals greater than 2 years would is solely an administrative action that has no
safety. be extended on a performance basis to no effect on any component and no impact on
Therefore, the proposed change does not longer than 75 months. Extensions of up to how the unit is operated.
involve a significant reduction in a margin of nine months (total maximum interval of 84 Therefore, the proposed change does not
safety. months for Type C tests) are permissible only result in a significant increase in the
for non-routine emergent conditions. The probability or consequences of an accident
The NRC staff has reviewed the proposed extensions do not involve either a previously evaluated.
licensee’s analysis and, based on this physical change to the plant or a change in 2. Does the proposed change create the
review, it appears that the three the manner in which the plant is operated or possibility of a new or different kind of
standards of 10 CFR 50.92(c) are controlled. The containment is designed to accident from any accident previously
satisfied. Therefore, the NRC staff provide an essentially leak tight barrier evaluated?
proposes to determine that the against the uncontrolled release of Response: No.
radioactivity to the environment for The proposed amendment to the TS
amendment request involves no postulated accidents. The containment and involves the extension of the HBRSEP2 Type
significant hazards consideration. the testing requirements invoked to A containment test interval to 15 years, the
Attorney for licensee: Lara S. Nichols, periodically demonstrate the integrity of the Type B test interval to 120 months for
Associate General Counsel, Duke Energy containment exist to ensure the plant’s selected components and the extension of the
Corporation, 526 South Church Street– ability to mitigate the consequences of an Type C test interval to 75 months for selected
EC07H, Charlotte, NC 28202. accident, and do not involve the prevention components. The containment and the
NRC Branch Chief: Michael T. or identification of any precursors of an testing requirements to periodically
Markley. accident. The change in dose risk for demonstrate the integrity of the containment
changing the Type A test frequency from exist to ensure the plant’s ability to mitigate
Duke Energy Progress, Inc., Docket No. three-per-ten years to once-per-fifteen years, the consequences of an accident do not
50–261, H. B. Robinson Steam Electric measured, as an increase to the total involve any accident precursors or initiators.
Plant, Unit No. 2, Darlington County, integrated plant risk for those accident The proposed change does not involve a
South Carolina sequences influenced by Type A testing, is physical change to the plant (i.e., no new or
0.020 person-rem [roentgen equivalent man]/ different type of equipment will be installed)
Date of amendment request: year. The Electric Power Research Institute or a change to the manner in which the plant
November 19, 2015. A publicly- (EPRI) Report No. 1009325, Revision 2–A, is operated or controlled.
available version is in ADAMS under states that a very small population dose is The proposed amendment also deletes an
Accession No. ML15323A085. defined as an increase of ≤1.0 person-rem per exception previously granted to allow one-
year, or ≤1% of the total population dose, time extension of the ILRT test frequency for
Description of amendment request: whichever is less restrictive for the risk HBRSEP2. This exception was for an activity
The proposed amendment would revise impact assessment of the extended integrated that has already taken place so the deletion
the Technical Specifications (TSs) to leak rate test (ILRT) intervals. Therefore, this is solely an administrative action that has no
allow the extension of the Type A proposed extension does not involve a effect on any component and no impact on
containment test interval to 15 years significant increase in the probability of an how the unit is operated.
and the extension of the Type B and accident previously evaluated. Therefore, the proposed change does not
Type C test intervals for selected As documented in NUREG–1493, Type B create the possibility of a new or different
components to 120 months and 75 and C tests have identified a very large kind of accident from any previously
months, respectively. The proposed percentage of containment leakage paths, and evaluated.
the percentage of containment leakage paths 3. Does the proposed change involve a
amendment also deletes from the TSs an that are detected only by Type A testing is significant reduction in a margin of safety?
already implemented one-time very small. The HBRSEP2 Type A test history Response: No.
extension of the Type A test frequency. supports this conclusion. The proposed amendment to TS 5.5.16
Basis for proposed no significant The integrity of the containment is subject involves the extension of the HBRSEP2 Type
asabaliauskas on DSK3SPTVN1PROD with NOTICES
hazards consideration determination: to two types of failure mechanisms that can A containment test interval to 15 years, the
As required by 10 CFR 50.91(a), the be categorized as: (1) Activity based, and (2) Type B test interval to 120 months for
licensee has provided its analysis of the time based. Activity based failure selected components and the extension of the
issue of no significant hazards mechanisms are defined as degradation due Type C test interval to 75 months for selected
to system and/or component modifications or components. This amendment does not alter
consideration, which is presented
maintenance. Local leak rate test the manner in which safety limits, limiting
below: requirements and administrative controls safety system set points, or limiting
1. Does the proposed amendment involve such as configuration management and conditions for operation are determined. The
a significant increase in the probability or procedural requirements for system specific requirements and conditions of the
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TS Containment Leak Rate Testing Program reactor vessel steam dome pressure The proposed changes do not create the
exist to ensure that the degree of containment associated with the Technical possibility of a new or different kind of
structural integrity and leak tightness that is Specification (TS) Safety Limits (SLs) accident from any accident previously
considered in the plant safety analysis is evaluated because the proposed reduction in
maintained. The overall containment leak
specified in TS 2.1.1 and TS 2.1.2. The
the reactor vessel steam dome pressure value
rate limit specified by TS is maintained. amendments would also revise the
in the safety limit in conjunction with the
The proposed change involves only the setpoint and allowable value for the increase in the trip setpoint and the
extension of the interval between Type A main steam line low pressure isolation allowable value for the main steam line low
containment leak rate tests, Type B tests and function in TS Table 3.3.2–2. The pressure isolation reflects a wider range of
Type C tests for HBRSEP2. The proposed proposed changes address a 10 CFR part applicability for the GEXL critical power
surveillance interval extension is bounded by 21 issue concerning the potential to correlation which is approved by the NRC for
the 15-year ILRT interval, the 120-month both GE14 and GNF2 fuel types in [the] LGS
violate the SLs limits during a pressure
Type B interval and the 75-month Type C test reactor cores.
interval currently authorized within NEI 94– regulator failure maximum demand
(open) transient. In addition, no new failure modes are
01, Revision 3–A. Industry experience being introduced. There are no changes in
supports the conclusion that Types B and C Basis for proposed no significant the method by which any plant systems
testing detects a large percentage of hazards consideration determination: perform a safety function. No new accident
containment leakage paths and that the As required by 10 CFR 50.91(a), the scenarios, failure mechanisms, or limiting
percentage of containment leakage paths that licensee has provided its analysis of the single failures are introduced as a result of
are detected only by Type A testing is small.
The containment inspections performed in
issue of no significant hazards the proposed changes.
consideration, which is presented The proposed changes do not introduce
accordance with ASME Section XI, TS and
the Maintenance Rule serve to provide a high below: any new accident precursors, nor do they
degree of assurance that the containment involve any changes in the methods
1. Do the proposed changes involve a governing normal plant operation. The
would not degrade in a manner that is significant increase in the probability or
detectable only by Type A testing. The proposed changes do not alter the outcome
consequences of an accident previously of the safety analysis.
combination of these factors ensures that the evaluated?
margin of safety in the plant safety analysis Therefore, the proposed changes do not
Response: No. create the possibility of a new or different
is maintained. The design, operation, testing The proposed changes do not involve a
methods and acceptance criteria for Types A, kind of accident from any accident
significant increase in the probability or previously evaluated.
B, and C containment leakage tests specified consequences of an accident previously
in applicable codes and standards would 3. Do the proposed changes involve a
evaluated because decreasing the reactor significant reduction in a margin of safety?
continue to be met, with the acceptance of
vessel steam dome pressure in TS Safety Response: No.
this proposed change, since these are not
Limits 2.1.1 and 2.1.2 for reactor thermal The margin of safety is established through
affected by changes to the Type A, Type B
power ranges and increasing the trip setpoint the design of the plant structures, systems,
and Type C test intervals.
The proposed amendment also deletes an and allowable value for the main steam line and components, and through the parameters
exception previously granted to allow one- low pressure isolation effectively expands for safe operation and setpoints for the
time extension of the ILRT test frequency for the validity range for GEXL critical power actuation of equipment relied upon to
HBRSEP2. This exception was for an activity correlation and the calculation of the respond to transients and design basis
that has already taken place so the deletion minimum critical power ratio. The critical accidents. Evaluation of the 10 CFR part 21
is solely an administrative action that has no power ratio rises during the pressure condition by General Electric determined
effect on any component and no impact on reduction following the scram that terminates that, since the critical power ratio improves
how the unit is operated. the Pressure Regulator Failure Maximum during the PRFO transient, there is no impact
Therefore, the proposed change does not Demand (Open) (PRFO) transient. The on the fuel safety margin, and therefore, there
involve a significant reduction in a margin of reduction in the reactor vessel steam dome is no challenge to fuel cladding integrity. The
safety. pressure value in the SL and the increase in proposed changes do not change the
the trip setpoint and the allowable value for requirements governing operation or
The NRC staff has reviewed the the main steam line low pressure isolation availability of safety equipment assumed to
licensee’s analysis and, based on this provides adequate margin to accommodate operate to preserve the margin of safety.
review, it appears that the three the pressure reduction during the PRFO The proposed changes are consistent with
standards of 10 CFR 50.92(c) are transient within the revised TS limit. the applicable NRC approved critical power
satisfied. Therefore, the NRC staff The proposed changes do not alter the use correlation for the fuel designs in use at LGS.
proposes to determine that the of the analytical methods used to determine The proposed changes do not alter the
amendment request involves no the safety limits that have been previously manner in which the safety limits are
significant hazards consideration. reviewed and approved by the NRC. The determined.
Attorney for licensee: Lara S. Nichols, proposed changes are in accordance with an The reduction in value of the reactor vessel
NRC approved critical power correlation steam dome pressure safety limit and the
Deputy General Counsel, Duke Energy methodology and do not adversely affect
Corporation, 550 South Tyron Street, increase in the trip setpoint and allowable
accident initiators or precursors. value for the main steam line low pressure
Mail Code DEC45A, Charlotte, NC The proposed changes do not alter or isolation provides adequate margin to
28202. prevent the ability of structures, systems, and accommodate the pressure reduction during
NRC Branch Chief: Benjamin G. components from performing their intended the PRFO transient within the revised TS
Beasley. function to mitigate the consequences of an limit.
initiating event within the applicable Therefore, the proposed changes do not
Exelon Generation Company, LLC, acceptance limits. The proposed changes are
Docket Nos. 50–352 and 50–353, involve a significant reduction in any margin
consistent with the safety analysis and of safety.
Limerick Generating Station (LGS), resultant consequences.
asabaliauskas on DSK3SPTVN1PROD with NOTICES
Units 1 and 2, Montgomery County, Therefore, the proposed changes do not The NRC staff has reviewed the
Pennsylvania involve a significant increase in the licensee’s analysis and, based on this
probability or consequences of an accident review, it appears that the three
Date of amendment request: January previously evaluated.
15, 2016. A publicly-available version is 2. Do the proposed changes create the
standards of 10 CFR 50.92(c) are
in ADAMS under Accession No. possibility of a new or different kind of satisfied. Therefore, the NRC staff
ML16015A316. accident from any accident previously proposes to determine that the
Description of amendment request: evaluated? amendment request involves no
The amendments would reduce the Response: No. significant hazards consideration.
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![Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices 13843
Attorney for licensee: Tamra Domeyer, operation of plant equipment beyond its issue of no significant hazards
Associate General Counsel, Exelon normal functional capabilities. DBNPS consideration, which is presented
Generation Company, LLC, 4300 functions will continue to be performed as below:
required. The proposed changes do not create
Winfield Road, Warrenville, IL 60555. 1. Does the proposed change involve a
any new credible failure mechanisms,
NRC Branch Chief: Douglas A. malfunctions, or accident initiators. significant increase in the probability or
Broaddus. Therefore, the proposed change does not consequences of an accident previously
create the possibility of a new or different evaluated?
FirstEnergy Nuclear Operating
kind of accident from any previously Response: No.
Company, et al., Docket No. 50–346, The proposed change revises or adds SRs
evaluated.
Davis-Besse Nuclear Power Station 3. Does the proposed amendment involve [Surveillance Requirements] that require
(DBNPS), Unit No. 1, Ottawa County, a significant reduction in a margin of safety? verification that the ECCS [Emergency Core
Ohio Response: No. Cooling System], RHR [Residual Heat
Date of amendment request: February The proposed changes to DBNPS’s EAL Removal] System, and the Containment
scheme to adopt the NRC-endorsed guidance Spray (CTS) System are not rendered
17, 2016. A publicly-available version is
in NEI 99–01, Revision 6, do not involve any inoperable due to accumulated gas and to
in ADAMS under Accession No. physical changes to plant systems or provide allowances which permit
ML16049A513. equipment. Margins of safety are unaffected performance of the revised verification. Gas
Description of amendment request: by the proposed changes. There are no accumulation in the subject systems is not an
The licensee proposes to change the changes being made to safety analysis initiator of any accident previously
emergency plan for DBNPS, Unit No. 1, assumptions, safety limits, or limiting safety evaluated. As a result, the probability of any
by revising the emergency action level system settings that would adversely affect accident previously evaluated is not
(EAL) scheme based on the Nuclear plant safety as a result of the proposed EAL significantly increased. The proposed SRs
Energy institute’s (NEl’s) guidance in scheme change. The proposed change does ensure that the subject systems continue to
NEI 99–01, Revision 6, ‘‘Development of not affect the technical specifications. There be capable to perform their assumed safety
are no changes to environmental conditions function and are not rendered inoperable due
Emergency Action Levels for Non- of any of the SSC or the manner in which any to gas accumulation. Thus, the consequences
Passive Reactors.’’ The NEI 99–01, SSC is operated. The applicable requirements of any accident previously evaluated are not
Revision 6, was endorsed by the NRC by of 10 CFR 50.47 and 10 CFR part 50, significantly increased.
letter dated March 28, 2013 (ADAMS appendix E will continue to be met. Therefore, the proposed change does not
Accession No. ML12346A463). Therefore, the proposed change does not involve a significant increase in the
Basis for proposed no significant involve a significant reduction in a margin of probability or consequences of an accident
hazards consideration determination: safety. previously evaluated.
As required by 10 CFR 50.91(a), the The NRC staff has reviewed the 2. Does the proposed change create the
licensee has provided its analysis of the licensee’s analysis and, based on this possibility of a new or different kind of
issue of no significant hazards accident from any accident previously
review, it appears that the three evaluated?
consideration, which is presented standards of 10 CFR 50.92(c) are Response: No.
below: satisfied. Therefore, the NRC staff The proposed change revises or adds SRs
1. Does the proposed amendment involve proposes to determine that the that require verification that the ECCS, the
a significant increase in the probability or amendment request involves no RHR System, and the CTS System are not
consequences of an accident previously significant hazards consideration. rendered inoperable due to accumulated gas
evaluated? Attorney for licensee: David W. and to provide allowances which permit
Response: No. Jenkins, Attorney, FirstEnergy performance of the revised verification. The
The proposed changes to DBNPS’s EAL Corporation, Mail Stop A–GO–15, 76 proposed change does not involve a physical
scheme to adopt the NRC-endorsed guidance alteration of the plant (i.e., no new or
South Main Street, Akron, OH 44308.
in NEI 99–01, Revision 6, do not involve any different type of equipment will be installed)
Acting NRC Branch Chief: Justin C.
physical changes to plant systems or or a change in the methods governing normal
equipment. The proposed changes do not Poole. plant operation. In addition, the proposed
alter any of the requirements of the technical Indiana Michigan Power Company, change does not impose any new or different
specifications. The proposed changes do not Docket Nos. 50–315 and 50–316, Donald requirements that could initiate an accident.
modify any plant equipment and do not The proposed change does not alter
C. Cook Nuclear Plant, Units 1 and 2,
impact any failure modes that could lead to assumptions made in the safety analysis and
an accident. Additionally, the proposed Berrien County, Michigan is consistent with the safety analysis
changes do not impact the ability of Date of amendment request: January assumptions.
structures, systems, or components (SSCs) to 29, 2016. A publicly-available version is Therefore, the proposed change does not
perform their intended safety functions in in ADAMS under Accession No. create the possibility of a new or different
mitigating the consequences of an initiating ML16034A032. kind of accident from any accident
event within the assumed acceptance limits. Description of amendment request: previously evaluated.
Therefore, the proposed change does not 3. Does the proposed change involve a
involve a significant increase in the
The proposed amendment would significant reduction in a margin of safety?
probability or consequences of an accident modify technical specification (TS) Response: No.
previously evaluated. requirements to address Generic Letter The proposed change revises or adds SRs
2. Does the proposed amendment create 2008–01, ‘‘Managing Gas Accumulation that require verification that the ECCS, the
the possibility of a new or different kind of in Emergency Core Cooling, Decay Heat RHR System, and the CTS System are not
accident from any accident previously Removal, and Containment Spray rendered inoperable due to accumulated gas
evaluated? Systems,’’ as described in the Technical and to provide allowances which permit
asabaliauskas on DSK3SPTVN1PROD with NOTICES
Response: No. Specification Task Force (TSTF) performance of the revised verification. The
The proposed changes to DBNPS’s EAL Traveler TSTF–523, Revision 2, proposed change adds new requirements to
scheme to adopt the NRC-endorsed guidance manage gas accumulation in order to ensure
in NEI 99–01, Revision 6, do not involve any
‘‘Generic Letter 2008–01, Managing Gas the subject systems are capable of performing
physical changes to plant systems or Accumulation.’’ their assumed safety functions. The proposed
equipment. The proposed changes do not Basis for proposed no significant SRs are more comprehensive than the current
involve the addition of any new plant hazards consideration determination: SRs and will ensure that the assumptions of
equipment. The proposed changes will not As required by 10 CFR 50.91(a), the the safety analysis are protected. The
alter the design configuration, or method of licensee has provided its analysis of the proposed change does not adversely affect
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![13844 Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices
any current plant safety margins or the consideration, which is presented changes being made to any safety analysis
reliability of the equipment assumed in the below: assumptions, safety limits, or limiting safety
safety analysis. Therefore, there are no system settings that would adversely affect
changes being made to any safety analysis 1. Does the proposed change involve a plant safety as a result of the proposed
assumptions, safety limits or limiting safety significant increase in the probability or change.
system settings that would adversely affect consequences of an accident previously Therefore, the proposed change does not
plant safety as a result of the proposed evaluated? involve a significant reduction in a margin of
Response: No. safety.
change.
The proposed change revises or adds
Therefore, the proposed change does not
Surveillance Requirement(s) (SRs) that The NRC staff has reviewed the
involve a significant reduction in a margin of
safety.
require verification that the Emergency Core licensee’s analysis and, based on this
Cooling System (ECCS), the Residual Heat review, it appears that the three
The NRC staff has reviewed the Removal (RHR) System, and the Containment standards of 10 CFR 50.92(c) are
licensee’s analysis and, based on this Spray (CS) System are not rendered
satisfied. Therefore, the NRC staff
review, it appears that the three inoperable due to accumulated gas and to
provide allowances which permit proposes to determine that the
standards of 10 CFR 50.92(c) are amendment requests involve no
performance of the revised verification. Gas
satisfied. Therefore, the NRC staff accumulation in the subject systems is not an significant hazards consideration.
proposes to determine that the initiator of any accident previously Attorney for licensee: Jennifer Post,
amendment request involves no evaluated. As a result, the probability of any Esq., Pacific Gas and Electric Company,
significant hazards consideration. accident previously evaluated is not P.O. Box 7442, San Francisco, CA
Attorney for licensee: Robert B. significantly increased. The proposed SRs 94120.
Haemer, Senior Nuclear Counsel, One ensure that the subject systems continue to NRC Branch Chief: Robert J.
Cook Place, Bridgman, MI 49106. be capable to perform their assumed safety Pascarelli.
NRC Branch Chief: David J. Wrona. function and are not rendered inoperable due
to gas accumulation. Thus, the consequences South Carolina Electric & Gas Company,
Pacific Gas and Electric Company, of any accident previously evaluated are not Docket Nos. 52–027 and 52–028, Virgil
Docket Nos. 50–275 and 50–323, Diablo significantly increased. C. Summer Nuclear Station (VCSNS)
Canyon Nuclear Power Plant, Unit Nos. Therefore, the proposed change does not
Units 2 and 3, Fairfield County, South
involve a significant increase in the
1 and 2, San Luis Obispo County, Carolina
probability or consequences of an accident
California previously evaluated. Date of amendment request: January
Date of amendment request: January 2. Does the proposed change create the 19, 2016. A publicly-available version is
21, 2106. A publicly-available version is possibility of a new or different accident in ADAMS under Accession No.
in ADAMS under Accession No. from any accident previously evaluated?
ML16019A403.
Response: No.
ML16021A067. The proposed change revises or adds SRs
Description of amendment request:
Description of amendment request: that require verification that the ECCS, RHR The requested amendment proposes to
The amendments would revise or add System, and CS System are not rendered depart from Tier 2* information in the
Surveillance Requirements to verify that inoperable due to accumulated gas and to Updated Final Safety Analysis Report
the system locations susceptible to gas provide allowances which permit (which includes the plant-specific
accumulation are sufficiently filled with performance of the revised verification. The design control document Tier 2
water and to provide allowances, which proposed change does not involve a physical information) related to the construction
permit performance of the verification. alteration of the plant (i.e., no new or methods used for the composite floors
different type of equipment will be installed)
The amendments would revise or a change in the methods governing normal
and roof of the auxiliary building.
Technical Specification (TS) 3.4.6, ‘‘RCS plant operation. In addition, the proposed Basis for proposed no significant
[Reactor Coolant System] Loops—MODE change does not impose any new or different hazards consideration determination:
4’’; TS 3.4.7, ‘‘RCS Loops—MODE 5, requirements that could initiate an accident. As required by 10 CFR 50.91(a), the
Loops Filled’’; TS 3.4.8, ‘‘RCS Loops— The proposed change does not alter licensee has provided its analysis of the
MODE 5, Loops Not Filled’’; TS 3.5.2, assumptions made in the safety analysis and issue of no significant hazards
‘‘ECCS [Emergency Core Cooling is consistent with the safety analysis consideration, which is presented
System]—Operating’’; TS 3.6.6, assumptions. below:
Therefore, the proposed change does not
‘‘Containment Spray and Cooling create the possibility of a new or different 1. Does the proposed amendment involve
Systems’’; TS 3.9.5, ‘‘RHR [Residual kind of accident from any accident a significant increase in the probability or
Heat Removal] and Coolant previously evaluated. consequences of an accident previously
Circulation—High Water Level’’; and TS 3. Does the proposed change involve a evaluated?
3.9.6, ‘‘RHR and Containment significant reduction in a margin of safety? Response: No.
Circulation—Low Water Level.’’ The Response: No. The design functions of the nuclear island
proposed amendments would modify The proposed change revises or adds SRs structures are to provide support, protection,
that require verification that the ECCS, the and separation for the seismic Category I
TS requirements to address Generic mechanical and electrical equipment located
RHR System, and the CS System are not
Letter 2008–01, ‘‘Managing Gas rendered inoperable due to accumulated gas, in the nuclear island. The nuclear island
Accumulation in Emergency Core and to provide allowances which permit structures are structurally designed to meet
Cooling, Decay Heat Removal, and performance of the revised verification. The seismic Category I requirements as defined in
Containment Spray Systems,’’ as proposed change adds new requirements to Regulatory Guide 1.29.
described in Technical Specification manage gas accumulation in order to ensure The use of ACI 349 and AISC N690
asabaliauskas on DSK3SPTVN1PROD with NOTICES
Task Force TSTF–523, Revision 2, the subject systems are capable of performing provides criteria for the design, qualification,
‘‘Generic Letter 2008–01, Managing Gas their assumed safety functions. The proposed fabrication, and inspection of composite steel
Accumulation.’’ SRs are more comprehensive than the current beam floors and roof in the auxiliary
SRs, and will ensure that the assumptions of building. These structures continue to meet
Basis for proposed no significant the safety analysis are protected. The the applicable portions of ACI 349 and AISC
hazards consideration determination: proposed change does not adversely affect N690. The proposed change does not have an
As required by 10 CFR 50.91(a), the any current plant safety margins or the adverse impact on the response of the
licensee has provided its analysis of the reliability of the equipment assumed in the nuclear island structures to safe shutdown
issue of no significant hazards safety analysis. Therefore, there are no earthquake ground motions or loads due to
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![Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices 13845
anticipated transients or postulated accident complies with the standards and Task Force (TSTF) Traveler TSTF–439–
conditions. The change does not impact the requirements of the Atomic Energy Act A, Revision 2, ‘‘Eliminate Second
support, design, or operation of mechanical of 1954, as amended (the Act), and the Completion Times Limiting Time From
and fluid systems. There is no change to Commission’s rules and regulations. Discovery of Failure to Meet an LCO
plant systems or the response of systems to
postulated accident conditions. There is no
The Commission has made appropriate [Limiting Condition of Operation],’’
change to the predicted radioactive releases findings as required by the Act and the dated June 20, 2005.
due to normal operation or postulated Commission’s rules and regulations in Date of issuance: February 19, 2016.
accident conditions. The plant response to 10 CFR Chapter I, which are set forth in Effective date: As of the date of
previously evaluated accidents or external the license amendment. issuance and shall be implemented
events is not adversely affected, nor does the A notice of consideration of issuance within 90 days from the date of
change described create any new accident of amendment to facility operating issuance.
precursors. license or combined license, as Amendment Nos.: Unit 1—197; Unit
Therefore, the proposed amendment does applicable, proposed no significant 2—197; Unit 3—197. A publicly-
not involve a significant increase in the hazards consideration determination, available version is in ADAMS under
probability or consequences of an accident and opportunity for a hearing in Accession No. ML16004A013;
previously evaluated.
2. Does the proposed amendment create
connection with these actions, was documents related to these amendments
the possibility of a new or different kind of published in the Federal Register as are listed in the Safety Evaluation
accident from any accident previously indicated. enclosed with the amendments.
evaluated? Unless otherwise indicated, the Renewed Facility Operating License
Response: No. Commission has determined that these Nos. NPF–41, NPF–51, and NPF–74: The
The proposed change revises the amendments satisfy the criteria for amendments revised the Operating
description of the construction of composite categorical exclusion in accordance Licenses and TSs.
steel beam floors and roof in the auxiliary with 10 CFR 51.22. Therefore, pursuant Date of initial notice in Federal
building. The proposed change does not to 10 CFR 51.22(b), no environmental Register: May 12, 2015 (80 FR 27195).
change the design function, support, design, impact statement or environmental The supplement dated January 19, 2016,
or operation of mechanical and fluid systems.
assessment need be prepared for these provided additional information that
The proposed change does not result in a
new failure mechanism for the pertinent amendments. If the Commission has clarified the application, did not expand
structures or new accident precursors. As a prepared an environmental assessment the scope of the application as originally
result, the design function of the structures under the special circumstances noticed, and did not change the staff’s
is not adversely affected by the proposed provision in 10 CFR 51.22(b) and has original proposed no significant hazards
change. made a determination based on that consideration determination as
Therefore, the proposed amendment does assessment, it is so indicated. published in the Federal Register.
not create the possibility of a new or different For further details with respect to the The Commission’s related evaluation
kind of accident from any accident action see (1) the applications for of the amendments is contained in a
previously evaluated. amendment, (2) the amendment, and (3) Safety Evaluation dated February 19,
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
the Commission’s related letter, Safety 2016.
Response: No. Evaluation and/or Environmental No significant hazards consideration
The proposed change is consistent with Assessment as indicated. All of these comments received: No.
ACI 349 and AISC N690. The design and items can be accessed as described in
Duke Energy Progress, Inc., Docket Nos.
construction of the auxiliary building floors the ‘‘Obtaining Information and
50–325 and 50–324, Brunswick Steam
and roof remain in conformance with the Submitting Comments’’ section of this
requirements in ACI 349 and AISC N690. Electric Plant, Units 1 and 2, Brunswick
document.
Therefore, the proposed amendment does County, North Carolina
not involve a significant reduction in a Arizona Public Service Company, et al., Date of amendment request: February
margin of safety. Docket Nos. STN 50–528, STN 50–529, 19, 2015, as supplemented by letter
The NRC staff has reviewed the and STN 50–530, Palo Verde Nuclear dated November 5, 2015.
licensee’s analysis and, based on this Generating Station, Unit Nos. 1, 2, and Description of amendment request:
review, it appears that the three 3, Maricopa County, Arizona The amendments revised (1) technical
standards of 10 CFR 50.92(c) are Date of amendment request: February specifications (TSs) by replacing
satisfied. Therefore, the NRC staff 27, 2015, as supplemented by letter AREVA Topical Report ANP–10298PA,
proposes to determine that the dated January 19, 2016. ‘‘ACE/ATRIUM 10XM Critical Power
amendment request involves no Brief description of amendments: The Correlation,’’ Revision 0, March 2010,
significant hazards consideration. amendments revised Technical with Revision 1, March 2014, of the
Attorney for licensee: Ms. Kathryn M. Specification (TS) 1.3, ‘‘Completion same topical report; and (2) Appendix
Sutton, Morgan, Lewis & Bockius LLC, Times’’; TS 3.7.5, ‘‘Auxiliary Feedwater B, ‘‘Additional Conditions,’’ by
1111 Pennsylvania Avenue NW., (AFW) System’’; TS 3.8.1, ‘‘AC removing the license condition issued
Washington, DC 20004–2514. [Alternating Current] Sources— by Amendment Nos. 262 and 290 for
Acting NRC Branch Chief: John Operating’’; and TS 3.8.9, ‘‘Distribution Units 1 and Unit 2, respectively.
McKirgan. Systems—Operating’’; to remove the Date of issuance: February 9, 2016.
second Completion Times. The Effective date: Once approved, the
III. Notice of Issuance of Amendments amendment also revised Example 1.3–3 Unit 1 amendment shall be
asabaliauskas on DSK3SPTVN1PROD with NOTICES
to Facility Operating Licenses and in TS 1.3, ‘‘Completion Times,’’ by implemented prior to start-up. from the
Combined Licenses adding a discussion of administrative 2016 Unit 1 refueling outage, and the
During the period since publication of controls to combinations of conditions Unit 2 amendment shall be
the last biweekly notice, the to ensure that the Completion Times for implemented prior to start-up from the
Commission has issued the following those conditions are not inappropriately 2017 Unit 2 refueling outage.
amendments. The Commission has extended. Amendment Nos.: 269 and 297. A
determined for each of these The changes are consistent with the publicly-available version is in ADAMS
amendments that the application NRC-approved Technical Specification under Accession No. ML16019A029;
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![Federal Register / Vol. 81, No. 50 / Tuesday, March 15, 2016 / Notices 13849
Date of issuance: February 29, 2016. the staff’s original proposed no For additional direction on obtaining
Effective date: As of its date of significant hazards consideration information and submitting comments,
issuance and shall be implemented determination. see ‘‘Obtaining Information and
within 90 days from the date of The Commission’s related evaluation Submitting Comments’’ in the
issuance. of the amendments is contained in a SUPPLEMENTARY INFORMATION section of
Amendment No.: 216. A publicly- Safety Evaluation dated February 22, this document.
available version is in ADAMS under 2016. FOR FURTHER INFORMATION CONTACT:
Accession No. ML16020A516; No significant hazards consideration Todd Keene, Office of Nuclear Reactor
documents related to this amendment comments received: No. Regulation, U.S. Nuclear Regulatory
are listed in the Safety Evaluation Dated at Rockville, Maryland, this 2nd day Commission, Washington DC 20555–
enclosed with the amendment. of March 2016. 0001; telephone: 301–415–1994, email:
Renewed Facility Operating License For the Nuclear Regulatory Commission. Todd.Keene@nrc.gov.
No. NPF–30: The amendment revised
Anne T. Boland, SUPPLEMENTARY INFORMATION:
the operating license and TSs.
Date of initial notice in Federal Director, Division of Operating Reactor I. Obtaining Information and
Licensing, Office of Nuclear Reactor
Register: July 7, 2015 (80 FR 38763). Submitting Comments
Regulation.
The supplemental letter dated
[FR Doc. 2016–05470 Filed 3–14–16; 8:45 am] A. Obtaining Information
November 9, 2015, provided additional
information that clarified the BILLING CODE 7590–01–P Please refer to Docket ID NRC–2016–
application, did not expand the scope of 0054 when contacting the NRC about
the application as originally noticed, the availability of information regarding
NUCLEAR REGULATORY this document. You may obtain
and did not change the staff’s original
COMMISSION publicly-available information related to
proposed no significant hazards
consideration determination as [NRC–2016–0054] this action by the following methods:
published in the Federal Register. • Federal Rulemaking Web site: Go to
The Commission’s related evaluation License Amendment Requests for http://www.regulations.gov and search
of the amendment is contained in a Changes to Emergency Response for Docket ID NRC–2016–0054.
Safety Evaluation dated February 29, Organization Staffing and • NRC’s Agencywide Documents
2016. Augmentation Access and Management System
No significant hazards consideration (ADAMS): You may obtain publicly
AGENCY: Nuclear Regulatory available documents online in the
comments received: No. Commission. ADAMS Public Documents collection at
Virginia Electric and Power Company, ACTION: Draft regulatory issue summary; http://www.nrc.gov/reading-rm/
Docket No. 50–339, North Anna Power request for comment. adams.html. To begin the search, select
Station, Unit No. 2, Louisa County, ‘‘ADAMS Public Documents’’ and then
Virginia SUMMARY: The U.S. Nuclear Regulatory
Commission (NRC) is seeking public select ‘‘Begin Web-based ADAMS
Date of amendment request: May 22, comment on a draft regulatory issue Search.’’ For problems with ADAMS,
2015. As supplemented by letter dated summary (RIS) to inform certain nuclear please contact the NRC’s Public
October 13, 2015. power reactor licensees of the use of Document Room (PDR) reference staff at
Brief description of amendment: The 1–800–397–4209, 301–415–4737, or by
guidance documents to support license
amendment revised the Technical email to pdr.resource@nrc.gov. The draft
amendment requests (LAR) to change
Specification (TS) 3.8.1, ‘‘AC Sources- RIS, ‘‘License Amendment Requests for
augmenting emergency response
Operating,’’ to remove the limitation in Changes to Emergency Response
organization (ERO) staffing and arrival
Note 1 that the surveillance is only Organization Staffing and
times. The RIS will clarify the scope and
applicable to Unit 1. Revised Augmentation,’’ is available in ADAMS
level of detail that should be provided
Surveillance Requirement (SR) 3.8.1.8 is under Accession No. ML15338A291.
to facilitate NRC review of the LARs.
applicable to both units. • NRC’s PDR: You may examine and
Date of issuance: February 22, 2016. DATES: Submit comments by April 14, purchase copies of public documents at
Effective date: As of the date of 2016. Comments received after this date the NRC’s PDR, Room O1–F21, One
issuance and shall be implemented will be considered if it is practical to do White Flint North, 11555 Rockville
within 30 days from the date of so, but the Commission is able to ensure Pike, Rockville, Maryland 20852.
issuance. consideration only for comments
Amendment No.: 260. A publicly- received before this date. B. Submitting Comments
available version is in ADAMS under ADDRESSES: You may submit comments Please include Docket ID NRC–2016–
Accession No. ML16013A444. by any of the following methods (unless 0054 in the subject line of your
Documents related to this amendment this document describes a different comment submission.
are listed in the Safety Evaluation method for submitting comments on a The NRC cautions you not to include
enclosed with the amendment. specific subject): identifying or contact information that
Renewed Facility Operating License • Federal Rulemaking Web site: Go to you do not want to be publicly
No. NPF–7: Amendment revised the http://www.regulations.gov and search disclosed in your comment submission.
Facility Operating License and for Docket ID NRC–2016–0054. Address The NRC will post all comment
asabaliauskas on DSK3SPTVN1PROD with NOTICES
Technical Specification. questions about NRC dockets to Carol submissions at http://
Date of initial notice in Federal Gallagher; telephone: 301–415–3463; www.regulations.gov as well as enter the
Register: July 21, 2015 (80 FR 43131). email: Carol.Gallagher@nrc.gov. comment submissions into ADAMS.
The supplement letter dated October 13, • Mail comments to: Cindy K. Bladey, The NRC does not routinely edit
2015, provided additional information Office of Administration, Mail Stop: comment submissions to remove
that clarified the application, did not OWFN–12–H08, U.S. Nuclear identifying or contact information.
expand the scope of the application as Regulatory Commission, Washington, If you are requesting or aggregating
originally noticed, and did not change DC 20555–0001. comments from other persons for
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Document Created: 2016-03-15 04:07:26
Document Modified: 2016-03-15 04:07:26
| 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 April 14, 2016. A request for a hearing must be filed by May 16, 2016. | |
| Contact | Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected] | |
| FR Citation | 81 FR 13837 |
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