80 FR 38756 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 80, Issue 129 (July 7, 2015)
Federal Register Volume 80, Issue 129 (July 7, 2015)
| Page Range | 38756-38770 | |
| FR Document | 2015-16539 |
[Federal Register Volume 80, Number 129 (Tuesday, July 7, 2015)] [Notices] [Pages 38756-38770] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2015-16539] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2015-0163] 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 June 11, 2015, to June 24, 2015. The last biweekly notice was published on June 23, 2015. DATES: Comments must be filed by August 6, 2015. A request for a hearing must be filed by September 8, 2015. 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-2015-0163. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear [[Page 38757]] 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: Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2015-0163 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-2015-0163. NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [email protected]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section. NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2015-0163, facility name, unit number(s), application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment 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 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 by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the requestor/ [[Page 38758]] 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. If a hearing is requested, 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. 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 [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, [[Page 38759]] 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. 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: December 19, 2014. A publicly-available version is in ADAMS under Accession No. ML14353A349. Description of amendment request: The proposed amendment would revise the technical specifications (TS) to adopt performance-based Type C testing for the reactor containment, which would allow for extended test intervals for Type C valves, and corrects an editorial issue in the TS. 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 adopts the NRC-accepted guidelines of [Nuclear Energy Institute] NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' for DBNPS performance-based Type C containment isolation valve testing. Revision 3-A of NEI 94-01 allows, based on previous valve leak test performance, an extension of Type C containment isolation valve leak test intervals. Since the change involves only performance-based Type C testing, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the components of the primary containment system will limit leakage rates to less than the values assumed in the plant safety analyses. The proposed amendment will not change the leakage rate acceptance requirements. As such, the containment will continue to perform its design function as a barrier to fission product releases. Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to revise the extended frequency performance-based Type C testing program does not change the design or operation of structures, systems, or components of the plant. The proposed amendment would continue to ensure containment operability and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by the proposed amendment. Therefore, the proposed amendment 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 amendment to revise the extended frequency performance-based Type C testing program does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation. The specific requirements and conditions of the Technical Specification Containment Leakage Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by Technical Specifications is maintained, thus ensuring the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods, and acceptance criteria for Type A, Type B, and Type 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 this revision to the performance-based containment testing program. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Travis L. Tate. FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio Date of amendment request: April 1, 2015. A publicly-available version is in ADAMS under Accession No. ML15091A143. Description of amendment request: The proposed amendment requests changes to certain technical specification minimum voltage and frequency acceptance criteria for emergency diesel generator (EDG) surveillance testing. [[Page 38760]] 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 would provide more restrictive acceptance criteria for certain EDG technical specification surveillance tests. The proposed acceptance criteria changes would help to ensure the EDGs are capable of carrying the electrical loading assumed in the safety analyses that take credit for the operation of the EDGs, would not affect the capability of other structures, systems, and components to perform their design function, and would not increase the likelihood of a malfunction. Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes would provide more restrictive acceptance criteria to be applied to existing technical specification surveillance tests that demonstrate the capability of the facility EDGs to perform their design function. The proposed acceptance criteria changes would not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. Therefore, the proposed amendment 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 EDG surveillance requirement changes involve increased minimum voltage and frequency test acceptance criteria. The conduct of surveillance tests on safety related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed amendment does not affect EDG performance as described in the design basis analyses, including the capability for the EDG to attain and maintain required voltage and frequency for accepting and supporting plant safety loads should an EDG start signal be received. The proposed amendment does not introduce changes to limits established in the accident analysis. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Travis L. Tate. National Institute of Standards and Technology (NIST), Docket No. 50- 184, Center for Neutron Research (NCNR), Montgomery County, Maryland Date of amendment request: June 23, 2014, as supplemented by letter dated August 20, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14196A043 and ML14241A384, respectively. Description of amendment request: The proposed amendment would revise the NIST NCNR's Technical Specifications, sections 3.6 and 4.6, pertaining to the replacement of NCNR's Uninterruptable Power Supplies (UPS) which supplies emergency alternating current power to reactor critical loads. 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 would not increase the probability or consequences of an accident previously evaluated. The proposed amendment modifies maintenance requirements for emergency power systems due to a change in battery technology used in commercially available UPS. The proposed amendment will assure the reliability of the emergency power systems utilizing valve-regulated lead acid (VRLA) batteries by increasing the frequency of performance testing as recommended by the battery manufacturer and the IEEE (Institute of Electrical and Electronics Engineers). The IEEE recommends the performance test Interval for VRLA batteries (IEEE-I 188) should not be greater than 25% of the expected service life or two years, whichever is less. The expected lifespan of a VRLA battery is ten years so a two year testing interval was selected. More frequent performance testing will ensure all the station batteries used for emergency power remain capable of supplying emergency electrical loads for a minimum of four hours as required. The proposed amendment will also correct a typographical error and add the requirement in the Limiting Conditions for Operations (LCO) for at least one of the two replacement UPS system batteries to be available to operate the reactor. Each UPS battery system is capable of independently supplying the designated emergency electrical loads for a minimum of four hours. Power for larger electrical loads such as primary cooling backup pumps (shutdown pumps) and emergency ventilation fans comes from other sources of emergency electrical power (diesel generators, critical power bus, or 125 VDC [volt direct current] station battery). 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The replacement UPS utilizes a different battery type (VRLA) which has shorter lifespan than traditional Vented Lead Acid (VLA) batteries. Increasing the frequency of performance monitoring as recommended by the IEEE accounts for the shorter lifespan of VRLA batteries and will enable the facility to identify a loss of battery capacity early to permit scheduled replacement of individual system components. Two identical but redundant UPS systems will each provide for a minimum of four hours at fully rated emergency power loading (20 kVA [kilovolt ampere]). The actual emergency electrical loads on the UPS will be significantly less because the larger electrical loads will continue to be powered from the 125 VDC station battery directly or from one of two emergency diesel generators. The new system will have higher reliability and capacity than the existing emergency power system. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment would not involve a significant reduction in a margin of safety. More frequent monitoring of the capacity or performance of the VRLA batteries utilized in the replacement UPS supplying power to critical reactor loads will ensure the UPS performs its design function and loss of battery capacity is detected early before safety margins are reduced. 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: Melissa J. Lieberman, Deputy Chief Counsel, National Institute of Standards and Technology, 100 Bureau Drive, Gaithersburg, MD 20899. NRC Branch Chief: Alexander Adams, Jr. Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2, Appling County, Georgia Date of amendment request: April 2, 2015. A publicly-available version is in [[Page 38761]] ADAMS under Accession No. ML15092A856. Description of amendment request: The proposed amendment modifies Technical Specifications (TS) section 1.0 (``Definitions''), Limiting Conditions for Operation and Surveillance Requirement Applicability, section 3.4.9 (``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits''), and section 5.0 (``Administrative Controls'') to delete reference to the pressure and temperature curves, and to include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of Boiling Water Reactor Owners Group (BWROG)-TP-11-022-A Revision 1 (SIR-05-044, Revision 1-A), ``Pressure- Temperature Limits Report Methodology for Boiling Water Reactors,'' dated June 2013, and of BWROG-TP-11-023-A, Revision 0 (0900876.401, Revision 0-A), ``Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations,'' dated May 2013, for preparation of the pressure and temperature (P-T) curves, and incorporates the guidance of Technical Specification Task Force (TSTF) TSTF-419-A, ``Revise PTLR Definition and References in Improved Standard Technical Specification (ISTS) 5.6.6, RCS PTLR.'' The HNP PTLRs have been developed based on the methodologies provided in BWROG- TP-11-022-A, Revision 1, and BWROG-TP-11-023-A, Revision 0, and based on the template provided in BWROG-TP-11-022-A, Revision 1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change modifies Edwin I. Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) Section 1.0 (``Definitions''), Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 (``RCS Pressure and Temperature (P/T) Limits''), and Section 5.0 (``Administrative Controls''), to delete reference to the pressure and temperature curves, and to include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of BWROG- TP-11-022-A, Revision 1 (SIR-05-044, Revision 1-A), ``Pressure- Temperature Limits Report Methodology for Boiling Water Reactors,'' dated June 2013, and of BWROG-TP-11-023-A, Revision 0 (0900876.401, Revision 0-A), ``Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations,'' dated May 2013, for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A, ``Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR.'' The HNP PTLRs have been developed based on the methodologies provided in BWROG-TP-11-022-A, Revision 1 and BWROG-TP-11-023-A, Revision 0, and based on the template provided in BWROG-TP-11-022-A, Revision 1. The HNP PTLRs meet all Conditions specified in the Safety Evaluation Reports (SERs) for BWROG-TP-11-022-A, Revision 1 and for BWROG-TP-11-023-A, Revision 0. The NRC has established requirements in Appendix G to 10 CFR 50 in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Additionally, the regulation in 10 CFR part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of reactor pressure vessel (RPV) material surveillance programs for operating lightwater reactors. Implementing these NRC approved methodologies does not reduce the ability to protect the reactor coolant pressure boundary as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodologies for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR, provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not affect the assumed accident performance of the RCPB, nor any plant structure, system, or component previously evaluated. The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change in methodology ensures that the RCPB remains capable of performing its safety functions. No set points are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change does not affect the function of the RCPB or its response during plant transients. There are no changes proposed which alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based upon the above, SNC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35201. NRC Branch Chief: Robert J. Pascarelli. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: October 7, 2014. A publicly-available version is in ADAMS under Accession No. ML14280A391. Description of amendment request: The proposed change would amend Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4. The requested amendment proposes to modify the existing feedwater controller logic to allow the controller program to respond as required to various plant transients while minimizing the potential for false actuation. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes will modify the control logic for actuation of the startup feedwater (SFW) pumps to support their [[Page 38762]] defense-in-depth function of core decay heat removal. The instrumentation used for actuation of the SFW pumps in their defense-in-depth function are not initiators of any accident. The proposed control logic uses different instrument tag numbers than the current design. The instruments used for the actuation of this function exist as a part of the current design; therefore this proposed change does not require any additional instrumentation. These instruments, to be included as part of the Design Reliability Assurance Program (D-RAP), will be held to the same enhanced quality assurance (QA) requirements as the current instruments and therefore neither safety, performance, nor reliance will be reduced as a part of this change. Additionally, the proposed changes do not adversely affect any accident initiating event or component failure, thus accidents previously evaluated are not adversely affected. In the event of loss of offsite power that results in a loss of main feedwater (MFW) supply, the SFW pumps automatically supply feedwater to the steam generators to cool down the reactor under emergency shutdown conditions. The standby source motor control center circuit powers each of the two SFW pumps and their associated instruments and valves. The pump discharge isolation valves are motor-operated and are normally closed and interlocked with the SFW pumps. In the event of loss of offsite power, the onsite standby power supply diesel generators will power the SFW pumps. If both the normal ac power and the onsite standby ac power are unavailable, these valves will fail ``as-is.'' The pump suction header isolation valves are pneumatically actuated. The main and startup feedwater system (FWS) also has temperature instrumentation in the pump discharge that would permit monitoring of the SFW temperature. This proposed change therefore has no impact on the ability of the AP1000 plant to cool down under emergency shutdown conditions or during a loss of offsite power event. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not adversely affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes will modify the control logic for actuation of the startup feedwater (SFW) pumps to support their defense-in- depth function of core decay heat removal. The instrumentation used for actuation of the SFW pumps in their defense-in-depth function are not initiators of any accident. The proposed control logic uses different instrument tag numbers than the current design. However, the instruments used for the actuation of this function already exist as a part of the current design and so this change does not require any additional instrumentation. These instruments, to be included as part of the D-RAP, will be held to the same enhanced QA requirements as the current instruments and so neither safety, performance, nor reliance will be reduced as a part of this change. Furthermore, since the D-RAP ensures consistency with the Probabilistic Risk Assessment (PRA), the changes do not impact the PRA. The proposed changes would not introduce a new failure mode, fault, or sequence of events that could result in a radioactive material release. The proposed change does not alter the design, configuration, or method of operation of the plant beyond standard functional capabilities of the equipment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes will modify the control logic for actuation of the startup feedwater (SFW) pumps to support their defense-in- depth function of core decay heat removal. These changes will have no negative impacts on the safety margin associated with the design functions of the SFW pumps. The proposed logic changes will only resolve the current conditions associated with undesired start up signals for the SFW pumps. The changes set forth in this amendment correct the actuation logic of the SFW pumps, so that the feedwater controller logic is now aligned with the guidance provided in the Advanced Light Water Reactor Utility Requirements Document (ALWR URD). In addition, the operation of the startup feedwater system function is not credited to mitigate a design-basis accident. Since there is no change to an existing design basis limit/criterion, design function, or regulatory criterion no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. Acting NRC Branch Chief: Paul Kallan. Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station (SSES), Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: March 19, 2015. A publicly-available version is available in ADAMS under Package Accession No. ML15091A657. Description of amendment request: The amendment would revise the Emergency Plan for SSES to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors'' (ADAMS Accession No. ML12326A805), which was endorsed by the NRC, as documented in NRC letter dated March 28, 2013 (ADAMS Accession No. ML12346A463). On June 1, 2015, the NRC staff issued an amendment changing the name on the SSES license from PPL Susquehanna, LLC, to Susquehanna Nuclear, LLC. This amendment was issued subsequent to an order issued on April 10, 2015, to SSES, approving an indirect license transfer. 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 any accident previously evaluated? Response: No. The proposed changes to the EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors,'' do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of the ERO [Emergency Response Organization] to respond in mitigating the consequences of any design basis accident. The probability of a reactor accident requiring implementation of Emergency Plan EALs has no relevance in determining whether the proposed changes to the EALs reduce the effectiveness of the Emergency Plan. As discussed in Section I.D, ``Planning Basis,'' of NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants''; . . . The overall objective of emergency response plans is to provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that could produce offsite doses in excess of Protective Action Guides (PAGs). No single specific accident sequence should be isolated as the one for which to plan because each accident could have different consequences, both in nature and degree. Further, the range of possible selection for a planning basis is very large, starting with a zero point of requiring no planning at all because significant offsite radiological accident consequences are unlikely to occur, to planning for the worst possible accident, regardless of its extremely low likelihood . . . Therefore, risk insights are not considered for any specific accident initiation or [[Page 38763]] progression in evaluating the proposed changes. The proposed changes do not involve any physical changes to plant equipment or systems, nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors nor do they alter the design assumptions, conditions, and configuration or the manner in which the plants are operated and maintained. The proposed changes do not adversely affect 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 changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the 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. All ERO 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 changes do not create the possibility of a new or different kind of accident from those that have been previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes to the EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6 do not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limit, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated. Margins of safety are unaffected by the proposed changes to adopt the NEI 99-01, Revision 6 EAL scheme guidance. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met. Therefore, the proposed changes do not involve any 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: Damon D. Obie, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101. NRC Branch Chief: Douglas A. Broaddus. Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1 (Callaway), Callaway County, Missouri Date of amendment request: May 8, 2015. A publicly-available version is in ADAMS under Accession No. ML15132A137. Description of amendment request: The amendment would change Technical Specification (TS) 2.1.1.1 and 5.6.5 to adopt NRC-approved Westinghouse Electric Company LLC topical report WCAP-14565-P-A, Addendum 2-P-A, ``Extended Application of ABB-NV [Asea Brown Boveri N.V.] Correlation and Modified ABB-NV Correlation WLOP [Westinghouse Low Pressure] for PWR [Pressurized-Water Reactor] Low Pressure Applications,'' April 2008 (proprietary). 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. Overall protection system performance will remain within the bounds of the accident analyses since there are no design changes. The design of the reactor trip system (RTS) instrumentation will be unaffected, and thus, the protection system will continue to function in a manner consistent with the plant design basis. All applicable design, material, and construction standards will continue to be maintained. The proposed changes will not affect any assumptions regarding accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility or the intended manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions. TS 5.6.5.b continues to ensure that the analytical methods used to determine the core operating limits meet NRC reviewed and approved methodologies. TS 5.6.5.c, unchanged by this amendment application, will continue to ensure that applicable limits of the safety analyses are met. The proposed change to TS 2.1.1.1 to specify only the true DNBR [departure from nucleate boiling ratio] safety limit without the addition of analytical uncertainties does not alter the use of the analytical methods used to determine core operating limits that have been reviewed and approved by the NRC. Removing analytical uncertainties from the TS would allow the use of current topical reports to refine those uncertainties without having to submit an amendment to the operating license, consistent with the intent of WCAP-14483-A [``Generic Methodology for Expanded Core Operating Limits Report,'' dated January 19, 1999; ADAMS Accession No. ML020430092]. Implementation of revisions to topical reports for Callaway Plant applications would still be reviewed in accordance with 10 CFR 50.59(c)(2)(viii) and, where required, receive prior NRC review and approval. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the FSAR [Final Safety Analysis Report]. The applicable radiological dose acceptance criteria will continue to be met. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The ABB-NV correlation was originally developed for Combustion Engineering fuel designs, and has also been qualified and licensed for Westinghouse fuel applications for the fuel region below the first mixing vane grid where the W-3 correlation is currently applied. The WLOP correlation is developed for DNBR calculations at low pressure conditions. The W-3A correlations, which are based exclusively on DNB [departure from nucleate boiling] data from rod bundle tests, have a wider applicable range and are more accurate than the W-3 correlation, leading to increased DNB margin in the plant safety analyses. The NRC-approved ABB-NV and WLOP correlation 95/95 DNBR limits with the VIPRE-W code are 1.13 and 1.18, respectively. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated, as the change is simply allowing the use of more accurate correlations when evaluating DNBR. The change does not involve any physical changes to the facility. Likewise, revising TS 2.1.1.1 to present the DNBR safety limit calculated using the WRB-2 methodology, without uncertainties being applied, does [sic] not introduce any new or different failure mode from what has been previously been evaluated. The change does not involve any change to a methodology, including how uncertainties are calculated and accounted for, nor does it involve any physical change to the facility. [[Page 38764]] Collectively, and based on the above, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The ABB-NV correlation was originally developed for Combustion Engineering fuel designs, and has also been qualified and licensed for Westinghouse fuel applications for the fuel region below the first mixing vane grid where the W-3 correlation is currently applied. The WLOP correlation is developed for DNBR calculations at low pressure conditions. The W-3A correlations, which are based exclusively on DNB data from rod bundle tests, have a wider applicable range and are more accurate than the W-3 correlation, leading to increased DNB margin in the plant safety analyses. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The currently listed Safety Limit in TS 2.1.1.1 for DNBR of 1.22 is calculated with some uncertainties statistically combined into the 1.17 value calculated using the WRB-2 methodology. These uncertainties are combined using the RTDP [Revised Thermal Design Procedure] methodology described in WCAP-11397-P-A [``Revised Thermal Design Procedure,'' April 1989 (proprietary)]. Callaway FSAR Section 4.4.1.1 discusses which uncertainties are statistically combined into the correlation limit. Revising TS 2.1.1.1 to present the DNBR safety limit calculated using the WRB-2 methodology, without uncertainties being applied, does not represent a change in methodology, but rather allows for changes in calculated uncertainties using methodologies previously approved for Callaway without requiring a license amendment. The proposed TS 2.1.1.1 revision does not represent a change in methodology for performing analyses. The proposed changes do not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications. The nominal RTS and ESFAS [engineered safety features actuation system] trip setpoints (as well as the associated allowable values) will remain unchanged. None of the acceptance criteria for any accident analysis will be changed. As there is no change to the source term, radiological release, or [dose] mitigation functions assumed in the accident analysis, the proposed changes have no impact on the radiological consequences of a design basis accident. Based on the above, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Unit Nos. 1 and 2 (NAPS), Louisa County, Virginia Date of amendment request: May 4, 2015. A publicly-available version is in ADAMS under Accession No. ML15131A026. Description of amendment request: The proposed license amendment would revise an incorrect equipment mark number, due to an administrative error in the current Emergency Action Levels (EAL). The amendment would correct the equipment mark number from the ``GW-RI-178- 1 Process Vent Normal Range'' monitor to the ``VG-RI-180-1 Vent Stack B Normal Range'' monitor for Initiating Condition (IC) RA2, EAL RA2.1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This administrative change affects the NAPS EALs, but does not alter any of the requirements of the Operating License or the Technical Specifications. The proposed change does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change has no effect on the consequences of any analyzed accident since the change does not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase 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 change affects the NAPS EALs by correcting an incorrect radiation monitor reference, but does not alter any of the requirements of the Operating License or the Technical Specifications. It does not modify any plant equipment and there is no impact on the capability of the existing equipment to perform its intended functions. No system setpoints are being modified. No new failure modes are introduced by the proposed change. The proposed amendment does not introduce an accident initiator or any malfunctions that would cause a new or different kind of accident. 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 change involve a significant reduction in a margin of safety? Response: No. The change affects the NAPS EALs, but does not alter any of the requirements of the Operating License or the Technical Specifications. The proposed change does not affect any of the assumptions used in the accident analysis, nor does it affect any operability requirements for equipment important to plant safety. Therefore, the proposed change will not result in a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Robert J. Pascarelli. 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 [[Page 38765]] under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station (Catawba), Units 1 and 2, York County, South Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370 McGuire Nuclear Station (McGuire), Units 1 and 2, Mecklenburg County, North Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station (Oconee), Units 1, 2, and 3, Oconee County, South Carolina Date of application for amendments: November 6, 2014. Brief description of amendments: The amendments changed the completion date for implementing Milestone 8 of the Duke Cyber Security Plan. Specifically, the amendments revised the date from June 30, 2015, to December 31, 2017. Date of issuance: June 11, 2015. Effective date: The license amendments are effective as of their dates of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: Catawba Unit 1--276; Catawba Unit 2--272; McGuire Unit 1--279; McGuire Unit 2--259; Oconee Unit 1--391; Oconee Unit 2-- 393; and Oconee Unit 3--392. A publicly-available version is in ADAMS under Accession No. ML15133A453; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF- 17, DPR-38, DPR-47, and DPR-55: Amendments revised the licenses. Date of initial notice in Federal Register: February 3, 2015 (80 FR 5818). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 11, 2015. No significant hazards consideration comments received: No. Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: November 8, 2010, as supplemented by letters dated September 1, 2011; June 28, 2012; March 28, April 18, September 27, and November 29, 2013; March 20 (two letters), and April 23, 2014; and May 28, 2015. Brief description of amendments: The amendments approved revisions to the updated final safety analysis report to incorporate the licensee's reactor vessel internals inspection plan based on the Materials Reliability Program: ``Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A),'' published by the Electric Power Research Institute. Date of issuance: June 19, 2015. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment Nos.: 392 for Unit 1, 394 for Unit 2, and 393 for Unit 3. A publicly-available version is in ADAMS under Accession No. ML15050A671; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The amendments revised the Facility Operating Licenses. Date of initial notice in Federal Register: October 2, 2012 (77 FR 60149). The supplemental letters dated September 1, 2011; March 28, April 18, September 27, and November 29, 2013; March 20 (two letters), and April 23, 2014; and May 28, 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 June 19, 2015. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of application for amendments: July 14, 2014. Brief description of amendments: The amendments revised and added Technical Specification (TS) surveillance requirements to address the concerns discussed in Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'' dated January 11, 2008. The TS changes are based on TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013. Date of issuance: June 19, 2015. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: 183, 183, 189, 189, 204, 244, 237, 214, 200, 257, and 252. A publicly-available version is in ADAMS under Accession No. ML15114A188; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, and DPR-30: The amendments revised the TSs and Licenses. Date of initial notice in Federal Register: October 28, 2014 (79 FR 64224). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 19, 2015. No significant hazards consideration comments received: No. [[Page 38766]] Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of amendment request: November 13, 2013, as supplemented by letter dated June 13, 2014. Brief description of amendments: The amendments revised Technical Specification (TS) 3.4.9, ``Pressurizer,'' TS 3.6.6, ``Containment Spray and Cooling Systems,'' TS 3.6.8, ``Iodine Removal System,'' TS 3.7.8, ``Control Room Emergency Ventilation System,'' and TS 3.7.12, ``Penetration Room Exhaust Ventilation System'' to provide a short completion time to restore an inoperable system for conditions under which existing TSs require a plant shutdown. Date of issuance: January 14, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 309 and 287. A publicly-available version is in ADAMS under Accession No. ML14307A842; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-53 and DPR-69: The amendments revised the Licenses and TSs. Date of initial notice in Federal Register: July 22, 2014 (79 FR 42548). The supplement dated June 13, 2014, 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. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 14, 2015. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of application for amendment: June 23, 2014, as supplemented by a letter dated February 27, 2015. Brief description of amendment: The amendment updated the technical specification (TS) pressure and temperature (P/T) figures using an NRC approved methodology to adjust the P/T limit curves for the previously missing data, addresses the reactor coolant system (RCS) vacuum condition that can occur under certain conditions, and aligns the heatup/cooldown requirements of the TS with the limits in the associated P/T figures. Additionally editorial changes are proposed related to the P/T figures including clarifications and updates to the associated titles, labeling, and notes. Date of issuance: June 12, 2015. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment No.: 168. A publicly-available version is in ADAMS under Accession No. ML15141A482; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-58: This amendment revised the TSs and License. Date of initial notice in Federal Register: September 30, 2014 (79 FR 58817). The February 27, 2015, supplement contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 12, 2015. 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: September 12, 2014. Brief description of amendment: The amendment modified the Technical Specification (TS) definition of SHUTDOWN MARGIN (SDM) to require calculation of SDM at the reactor moderator temperature corresponding to the most reactive state throughout the operating cycle. The changes are consistent with the approved Technical Specification Task Force (TSTF) Traveler, TSTF-535, Revision 0. Date of issuance: June 23, 2015. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 169. A publicly-available version is in ADAMS under Accession No. ML15160A028; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-58: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: November 25, 2014 (79 FR 70216). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 23, 2015. No significant hazards consideration comments received: No. Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: February 20, 2014, as supplemented by letters dated December 11, 2014, and January 13, January 28, April 18, and May 19, 2015. Brief description of amendments: The amendments revised the Technical Specifications (TSs) by relocating specific surveillance frequency requirements to a licensee-controlled program with implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The NEI 04-10 methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Guide 1.177, ``An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications'' (ADAMS Accession No. ML003740176). The changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specifications Change TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee Control--RITSTF [Risk Informed Technical Specifications Task Force] Initiative 5b'' (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of TSTF-425, Revision 3. The amendments also include editorial changes to the TSs, administrative deviations from TSTF-425, and other changes resulting from differences between the St. Lucie Plant TSs and the TSs on which TSTF-425 was based. Date of issuance: June 22, 2015. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 223 and 173. A publicly-available version is in ADAMS under Accession No. ML15127A066; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs. Date of initial notice in Federal Register: The NRC staff initially made a proposed determination that the amendment request dated February 20, [[Page 38767]] 2014, involved no significant hazards consideration (July 22, 2014, 79 FR 42550). By letters dated December 11, 2014, and January 13, 2015, the licensee provided clarifying information that did not expand the scope of the application and did not change the NRC staff's original proposed no significant hazards consideration (NSHC) determination, as published in the Federal Register on July 22, 2014 (79 FR 44550). Subsequently, by letter dated January 28, 2015, the licensee supplemented its amendment request with a proposed change that expanded the scope of the request. Therefore, the NRC published a second proposed NSHC determination in the Federal Register on March 3, 2015 (80 FR 11477), which superseded the notice dated July 22, 2014 (79 FR 44550). The licensee's supplements dated April 18, 2015, and May 19, 2015, provided clarifying information that did not expand the scope of the submittal and did not change the NRC staff's proposed NSHC determination, as published in the notice dated March 3, 2015 (80 FR 11477). The Commission's related evaluation of the amendments is contained in an SE dated June 22, 2015. No significant hazards consideration comments received: No. PSEG Nuclear LLC, Docket Nos. 50-354, 50-272 and 50-311, Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date amendment request: December 9, 2014, as supplemented by letter dated April 9, 2015. Brief description of amendments: The amendments revised the PSEG Nuclear LLC (PSEG) Environmental Protection Plans (non-radiological) (EPPs), contained in Appendix B to the renewed facility operating licenses for Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2, to clarify that PSEG must adhere to the currently applicable biological opinion issued by the National Marine Fisheries Service. The amendments also simplify the Aquatic Monitoring section of the EPPs, modify reporting requirements related to New Jersey Pollutant Discharge Elimination System permits, modify the criteria for reporting Unusual or Important Environmental Events, and remove the requirement for PSEG to submit an Annual Environmental Operating Report. Date of issuance: June 17, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 198, 308, 290. A publicly-available version is in ADAMS under Accession No. ML15141A271; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The amendments revised the Renewed Facility Operating Licenses and EPPs. Date of initial notice in Federal Register: April 14, 2015 (80 FR 20024). The supplemental letter dated April 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 amendments is contained in a Safety Evaluation dated June 17, 2015. No significant hazards consideration comments received: Yes. The Commission's related evaluation of the public comments is contained in the safety evaluation dated June 17, 2015. Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, 50-362, and 72-041, San Onofre Nuclear Generating Station (SONGS), Units 1, 2, and 3, and the Independent Spent Fuel Storage Installation, San Diego County, California Date of amendment request: March 31, 2014, as supplemented by letters dated October 21, 2014, March 17, 2015, and April 29, 2015. Brief description of amendment: The amendments revised the SONGS emergency plan to reflect the low likelihood of any credible accident at the facility in its permanently shutdown and defueled condition that could result in radiological releases requiring offsite protective measures and how, in the unlikely event of certain severe, beyond- design-basis accidents, sufficient time would be available to initiate appropriate mitigating actions, and if needed, for offsite authorities to implement protective actions to protect the public health and safety. Date of issuance: June 5, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: Unit 1--167; Unit 2--229; Unit 3--222. A publicly- available version is in ADAMS under Accession No. ML15126A461; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The amendments revised the emergency plan. Date of initial notice in Federal Register: December 23, 2014 (79 FR 77049). The supplemental letters dated March 17, 2015, and April 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 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 June 5, 2015. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear Plant, Unit 2, Limestone County, Alabama Date of amendment request: June 19, 2014, as supplemented by letter dated December 2, 2014. Brief description of amendment: The amendment revised the Technical Specification (TS) 3.4.9, ``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' and Figures 3.4.9-1 through 3.4.9-2. The P/ T limits are based on proprietary topical report NEDC-33178P-A, Revision 1, ``GE [General Electric] Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves.'' NEDO-33178-A, Revision 1, is the non-proprietary version of the NRC-approved topical report. Date of issuance: June 2, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 314. A publicly available version is in ADAMS under Accession No. ML15065A049; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-52: The amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: February 3, 2015 (80 FR 5819). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 2, 2015. No significant hazards consideration comments received: No. [[Page 38768]] Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia Date of application for amendment: June 30, 2014, as supplemented by letter dated July 28, 2015. Brief description of amendment: The license amendments approved the changes to the Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing Program,'' by replacing the reference to Regulatory Guide (RG) 1.163 with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, as the implementation document used to develop the North Anna performance- based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. Date of issuance: June 16, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 274 and 257. A publicly-available version is in ADAMS under Accession No. ML15133A381; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License Nos. NPF-4 and NPF-7: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: September 2, 2014 (79 FR 52070). The supplement dated January 28, 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 June 16, 2015. No significant hazards consideration comments received: No. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances) 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 for the amendment 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. Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual notice of consideration of issuance of amendment, proposed no significant hazards consideration determination, and opportunity for a hearing. For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments. In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible. Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved. The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 51.12(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 application for amendment, (2) the amendment to Facility Operating License or Combined License, as applicable, 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. A. Opportunity To Request a Hearing and Petition for Leave To Intervene The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, and electronically on the Internet at the NRC's Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or by email to [email protected]. If a request for a hearing or petition for leave to intervene is filed by the above date, the [[Page 38769]] Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) the name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. 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 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least ten 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [email protected], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC- issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's ``Guidance for Electronic Submission,'' which is available on the agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software. If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic 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 [[Page 38770]] 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. Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington Date of application for amendment: May 15, 2015, as supplemented by letter dated May 19, 2015. Brief description of amendment: The amendment extended the implementation period for Amendment No. 232, License Amendment Request for Changing Technical Specification Table 3.3.1.1-1 Function 7, ``Scram Discharge Volume Water Level--High,'' which was issued on March 27, 2015 (ADAMS Accession No. ML15063A010). Amendment No. 232 was effective as of the date of issuance (i.e., on March 27, 2015), and was required to be implemented prior to restarting from refueling outage R- 22, scheduled for spring 2015. Amendment No. 235 extends the implementation period for Amendment No. 232 to prior to restarting from refueling outage R-23, scheduled for spring 2017. Date of issuance: June 11, 2015. Effective date: As of its date of issuance and shall be implemented prior to restarting from refueling outage R-23, scheduled for spring 2017. Amendment No.: 235. A publicly-available version is in ADAMS under Accession No. ML15154A800; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-21: Amendment revised the Renewed Facility Operating License to extend the implementation date of Amendment No. 232, issued on March 27, 2015, to prior to restarting from refueling outage R-23, scheduled for spring 2017. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Public notice of the proposed amendment was published in the Tri-City Herald, located in in Kennewick, Washington, from June 2 through June 4, 2015. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments were received. The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, public comments, and final NSHC determination are contained in a Safety Evaluation dated June 11, 2015. Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817. NRC Branch Chief: Michael T. Markley. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1 (WBN-1), Rhea County, Tennessee Date of amendment request: May 29, 2015, as supplemented by letter dated June 5, 2015. Description of amendment request: The amendment provided a one-time change to Technical Specification Table 3.3.4-1, Function 4a, ``Reactor Coolant System (RCS) Hot Leg Temperature Indication,'' to permit the temperature indication for RCS Loop 4 to be inoperable for the remainder of WBN-1 Operating Cycle 13. Date of issuance: June 12, 2015. Effective date: June 12, 2015. Amendment No.: 100. A publicly-available version is in ADAMS under Accession No. ML15160A407; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. (NPF-90): The amendment revised the Technical Specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. (The Advocate & Democrat and The Herald-News on June 7 and June 10, 2015 as well as The Daily Post-Athenian on June 5 and June 8, 2015.) The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments were received. The supplemental letter dated June 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 Advocate & Democrat and The Herald-News on June 7, and June 10, 2015, as well as The Daily Post-Athenian, on June 5, and June 8, 2015. The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and NSHC determination are contained in a safety evaluation dated June 12, 2015. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, TN 37902. NRC Branch Chief: Jessie F. Quichocho. Dated at Rockville, Maryland, this 29th day of June 2015. For the Nuclear Regulatory Commission. George A. Wilson, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2015-16539 Filed 7-6-15; 8:45 am] BILLING CODE 7590-01-P
![38756 Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices
I. Obtaining Information and II. Discussion For the Nuclear Regulatory Commission.
Submitting Comments Mark Henry Salley,
Aspirated smoke detection systems Chief, Fire Research Branch, Division of Risk
A. Obtaining Information have been available on the commercial Analysis, Office of Nuclear Regulatory
Please refer to Docket ID NRC–2015– market for more than four decades as an Research.
0112 when contacting the NRC about alternative technology to spot-type [FR Doc. 2015–16547 Filed 7–6–15; 8:45 am]
the availability of information for this smoke detection for detecting products BILLING CODE 7590–01–P
action. You may obtain publicly- of combustion. In the United States,
available information related to this several nuclear power plants (NPPs)
action by any of the following methods: have installed these systems as early as NUCLEAR REGULATORY
• Federal Rulemaking Web site: Go to the mid-1990s as an alternative method COMMISSION
http://www.regulations.gov and search to conventional fire detection systems
for Docket ID NRC–2015–0112. [NRC–2015–0163]
with the idea to provide advanced
• NRC’s Agencywide Documents warning of potential fire threats.
Access and Management System Biweekly Notice; Applications and
Recently, there has been indication that Amendments to Facility Operating
(ADAMS): You may obtain publicly- numerous licensees of NPPs
available documents online in the Licenses and Combined Licenses
transitioning to a performance-based fire Involving No Significant Hazards
ADAMS Public Documents collection at protection program intend to install
http://www.nrc.gov/reading-rm/ Considerations
these types of systems configured as
adams.html. To begin the search, select very early warning fire detection. In AGENCY: Nuclear Regulatory
‘‘ADAMS Public Documents’’ and then many, but not all cases, the choice to Commission.
select ‘‘Begin Web-based ADAMS install these systems is based on the ACTION: Biweekly notice.
Search.’’ For problems with ADAMS,
expectation that these systems may
please contact the NRC’s Public SUMMARY: Pursuant to section 189a. (2)
reduce the estimated fire risk in a fire
Document Room (PDR) reference staff at of the Atomic Energy Act of 1954, as
probabilistic risk assessment (PRA).
1–800–397–4209, 301–415–4737, or by amended (the Act), the U.S. Nuclear
email to pdr.resource@nrc.gov. The In 2008, the NRC issued a staff Regulatory Commission (NRC) is
proposed draft NUREG–2180, interim position documented in a publishing this regular biweekly notice.
‘‘Determining the Effectiveness, National Fire Protection Association The Act requires the Commission to
Limitations, and Operator Response for Standard 805 Frequently Asked publish notice of any amendments
Very Early Warning Fire Detection Question 08–0046, ‘‘Incipient Fire issued, or proposed to be issued and
Systems in Nuclear Facilities Detection Systems.’’ This staff interim grants the Commission the authority to
(DELORES–VEWFIRE), Draft Report for position provides guidance on the use of issue and make immediately effective
Comment,’’ is available electronically these systems and the associated fire any amendment to an operating license
under ADAMS Accession No. PRA quantification for in-cabinet or combined license, as applicable,
ML15162A416. applications. At that time, there was upon a determination by the
• NRC’s PDR: You may examine and limited test data and PRA experience Commission that such amendment
purchase copies of public documents at available for those applications and as involves no significant hazards
the NRC’s PDR, Room O1–F21, One such a confirmatory research program consideration, notwithstanding the
White Flint North, 11555 Rockville was needed. Research was also needed pendency before the Commission of a
Pike, Rockville, Maryland 20852. to advance the state of knowledge request for a hearing from any person.
B. Submitting Comments related to the performance of these This biweekly notice includes all
systems. This report documents the notices of amendments issued, or
Please include Docket ID NRC–2015– results and findings from the proposed to be issued from June 11,
0112 in your comment submission. confirmatory research program. 2015, to June 24, 2015. The last
The NRC cautions you not to include
Specific areas of this draft report biweekly notice was published on June
identifying or contact information that
where comments and additional 23, 2015.
you do not want to be publicly
disclosed in your comment submission. relevant information or supporting data DATES: Comments must be filed by
The NRC will post all comment are sought include: August 6, 2015. A request for a hearing
submissions at http:// 1. System availability, including must be filed by September 8, 2015.
www.regulations.gov as well as enter the system down time and surveillance test ADDRESSES: You may submit comments
comment submissions into ADAMS. interval for the aspirated smoke by any of the following methods (unless
The NRC does not routinely edit detection systems used in nuclear and this document describes a different
comment submissions to remove non-nuclear facilities. method for submitting comments on a
identifying or contact information. specific subject):
2. Time duration between a very early
If you are requesting or aggregating • Federal Rulemaking Web site: Go to
warning fire detection system ‘‘alert’’
comments from other persons for http://www.regulations.gov and search
condition and the commencement of
submission to the NRC, then you should for Docket ID NRC–2015–0163. Address
inform those persons not to include flaming conditions. Alternatively, the questions about NRC dockets to Carol
identifying or contact information that time duration of the incipient stage, Gallagher; telephone: 301–415–3463;
they do not want to be publicly from start of component degradation to email: Carol.Gallagher@nrc.gov. For
flaming conditions. Include a
tkelley on DSK3SPTVN1PROD with NOTICES
disclosed in their comment submission. technical questions, contact the
Your request should state that the NRC description of the type of electrical individual listed in the FOR FURTHER
does not routinely edit comment enclosure (e.g., motor control center, INFORMATION CONTACT section of this
submissions to remove such information relay rack, control panel, etc.) and document.
before making the comment voltage level of initiation component) • Mail comments to: Cindy Bladey,
submissions available to the public or Dated at Rockville, Maryland, this 26th day Office of Administration, Mail Stop:
entering the comment into ADAMS. of June 2015. OWFN–12–H08, U.S. Nuclear
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![Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices 38759
express mail, or expedited delivery FirstEnergy Nuclear Operating ensure operation within the bounds of
service to the Office of the Secretary, Company, et al., Docket No. 50–346, existing accident analyses. There are no
Sixteenth Floor, One White Flint North, Davis-Besse Nuclear Power Station accident initiators created or affected by the
11555 Rockville Pike, Rockville, proposed amendment.
(DBNPS), Unit No. 1, Ottawa County,
Therefore, the proposed amendment does
Maryland, 20852, Attention: Ohio not create the possibility of a new or different
Rulemaking and Adjudications Staff. Date of amendment request: kind of accident from any previously
Participants filing a document in this December 19, 2014. A publicly-available evaluated.
manner are responsible for serving the version is in ADAMS under Accession 3. Does the proposed amendment involve
document on all other participants. No. ML14353A349. a significant reduction in a margin of safety?
Filing is considered complete by first- Description of amendment request: Response: No.
class mail as of the time of deposit in The proposed amendment to revise the
The proposed amendment would revise extended frequency performance-based Type
the mail, or by courier, express mail, or the technical specifications (TS) to
expedited delivery service upon C testing program does not affect plant
adopt performance-based Type C testing operations, design functions, or any analysis
depositing the document with the for the reactor containment, which that verifies the capability of a structure,
provider of the service. A presiding would allow for extended test intervals system, or component of the plant to perform
officer, having granted an exemption for Type C valves, and corrects an a design function. In addition, this change
request from using E-Filing, may require editorial issue in the TS. does not affect safety limits, limiting safety
a participant or party to use E-Filing if Basis for proposed no significant system setpoints, or limiting conditions for
the presiding officer subsequently hazards consideration determination: operation. The specific requirements and
determines that the reason for granting As required by 10 CFR 50.91(a), the conditions of the Technical Specification
the exemption from use of E-Filing no Containment Leakage Rate Testing Program
licensee has provided its analysis of the exist to ensure that the degree of containment
longer exists. issue of no significant hazards
Documents submitted in adjudicatory structural integrity and leak-tightness that is
consideration, which is presented considered in the plant safety analysis is
proceedings will appear in the NRC’s
below: maintained.
electronic hearing docket which is
1. Does the proposed amendment involve The overall containment leak rate limit
available to the public at http:// specified by Technical Specifications is
ehd1.nrc.gov/ehd/, unless excluded a significant increase in the probability or
consequences of an accident previously maintained, thus ensuring the margin of
pursuant to an order of the Commission, safety in the plant safety analysis is
evaluated?
or the presiding officer. Participants are Response: No. maintained. The design, operation, testing
requested not to include personal The proposed amendment adopts the NRC- methods, and acceptance criteria for Type A,
privacy information, such as social accepted guidelines of [Nuclear Energy Type B, and Type C containment leakage
security numbers, home addresses, or Institute] NEI 94–01, Revision 3–A, ‘‘Industry tests specified in applicable codes and
home phone numbers in their filings, Guideline for Implementing Performance- standards would continue to be met with the
unless an NRC regulation or other law Based Option of 10 CFR part 50, Appendix acceptance of this proposed change, since
J,’’ for DBNPS performance-based Type C these are not affected by this revision to the
requires submission of such
containment isolation valve testing. Revision performance-based containment testing
information. However, in some program.
3–A of NEI 94–01 allows, based on previous
instances, a request to intervene will valve leak test performance, an extension of Therefore, the proposed amendment does
require including information on local Type C containment isolation valve leak test not involve a significant reduction in a
residence in order to demonstrate a intervals. Since the change involves only margin of safety.
proximity assertion of interest in the performance-based Type C testing, the
proceeding. With respect to copyrighted proposed amendment does not involve either
The NRC staff has reviewed the
works, except for limited excerpts that a physical change to the plant or a change in licensee’s analysis and, based on this
serve the purpose of the adjudicatory the manner in which the plant is operated or review, it appears that the three
filings and would constitute a Fair Use controlled. standards of 10 CFR 50.92(c) are
application, participants are requested Implementation of these guidelines satisfied. Therefore, the NRC staff
continues to provide adequate assurance that proposes to determine that the
not to include copyrighted materials in during design basis accidents, the
their submission. amendment request involves no
components of the primary containment significant hazards consideration.
Petitions for leave to intervene must system will limit leakage rates to less than
be filed no later than 60 days from the the values assumed in the plant safety Attorney for licensee: David W.
date of publication of this notice. analyses. Jenkins, Attorney, FirstEnergy
Requests for hearing, petitions for leave The proposed amendment will not change Corporation, Mail Stop A–GO–15, 76
to intervene, and motions for leave to the leakage rate acceptance requirements. As South Main Street, Akron, OH 44308.
file new or amended contentions that such, the containment will continue to NRC Branch Chief: Travis L. Tate.
are filed after the 60-day deadline will perform its design function as a barrier to
fission product releases. FirstEnergy Nuclear Operating
not be entertained absent a Company, et al., Docket No. 50–346,
Therefore, the proposed amendment does
determination by the presiding officer not significantly increase the probability or Davis-Besse Nuclear Power Station, Unit
that the filing demonstrates good cause consequences of an accident previously No. 1, Ottawa County, Ohio
by satisfying the three factors in 10 CFR evaluated.
2.309(c)(1)(i)–(iii). 2. Does the proposed amendment create Date of amendment request: April 1,
For further details with respect to the possibility of a new or different kind of 2015. A publicly-available version is in
these license amendment applications, accident from any accident previously ADAMS under Accession No.
see the application for amendment evaluated? ML15091A143.
Response: No.
tkelley on DSK3SPTVN1PROD with NOTICES
which is available for public inspection Description of amendment request:
in ADAMS and at the NRC’s PDR. For The proposed amendment to revise the The proposed amendment requests
extended frequency performance-based Type
additional direction on accessing C testing program does not change the design changes to certain technical
information related to this document, or operation of structures, systems, or specification minimum voltage and
see the ‘‘Obtaining Information and components of the plant. frequency acceptance criteria for
Submitting Comments’’ section of this The proposed amendment would continue emergency diesel generator (EDG)
document. to ensure containment operability and would surveillance testing.
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Basis for proposed no significant standards of 10 CFR 50.92(c) are battery system is capable of independently
hazards consideration determination: satisfied. Therefore, the NRC staff supplying the designated emergency
As required by 10 CFR 50.91(a), the proposes to determine that the electrical loads for a minimum of four hours.
Power for larger electrical loads such as
licensee has provided its analysis of the amendment request involves no primary cooling backup pumps (shutdown
issue of no significant hazards significant hazards consideration. pumps) and emergency ventilation fans
consideration, which is presented Attorney for licensee: David W. comes from other sources of emergency
below: Jenkins, Attorney, FirstEnergy electrical power (diesel generators, critical
1. Does the proposed amendment involve Corporation, Mail Stop A–GO–15, 76 power bus, or 125 VDC [volt direct current]
a significant increase in the probability or South Main Street, Akron, OH 44308. station battery).
consequences of an accident previously NRC Branch Chief: Travis L. Tate. 2. Does the change create the possibility of
evaluated? a new or different kind of accident from any
National Institute of Standards and accident previously evaluated?
Response: No.
The proposed amendment would provide
Technology (NIST), Docket No. 50–184, Response: No.
Center for Neutron Research (NCNR), The proposed amendment would not
more restrictive acceptance criteria for
Montgomery County, Maryland create the possibility of a new or different
certain EDG technical specification
kind of accident from any accident
surveillance tests. The proposed acceptance Date of amendment request: June 23, previously evaluated. The replacement UPS
criteria changes would help to ensure the 2014, as supplemented by letter dated utilizes a different battery type (VRLA) which
EDGs are capable of carrying the electrical
loading assumed in the safety analyses that
August 20, 2014. Publicly-available has shorter lifespan than traditional Vented
versions are in ADAMS under Lead Acid (VLA) batteries. Increasing the
take credit for the operation of the EDGs,
Accession Nos. ML14196A043 and frequency of performance monitoring as
would not affect the capability of other
ML14241A384, respectively. recommended by the IEEE accounts for the
structures, systems, and components to
Description of amendment request: shorter lifespan of VRLA batteries and will
perform their design function, and would not
enable the facility to identify a loss of battery
increase the likelihood of a malfunction. The proposed amendment would revise capacity early to permit scheduled
Therefore, the proposed amendment does the NIST NCNR’s Technical replacement of individual system
not significantly increase the probability or Specifications, sections 3.6 and 4.6, components. Two identical but redundant
consequences of an accident previously pertaining to the replacement of NCNR’s UPS systems will each provide for a
evaluated. Uninterruptable Power Supplies (UPS) minimum of four hours at fully rated
2. Does the proposed amendment create emergency power loading (20 kVA [kilovolt
the possibility of a new or different kind of
which supplies emergency alternating
current power to reactor critical loads. ampere]). The actual emergency electrical
accident from any accident previously loads on the UPS will be significantly less
evaluated? Basis for proposed no significant
hazards consideration determination: because the larger electrical loads will
Response: No. continue to be powered from the 125 VDC
The proposed changes would provide more As required by 10 CFR 50.91(a), the station battery directly or from one of two
restrictive acceptance criteria to be applied to licensee has provided its analysis of the emergency diesel generators. The new system
existing technical specification surveillance issue of no significant hazards will have higher reliability and capacity than
tests that demonstrate the capability of the consideration, which is presented the existing emergency power system.
facility EDGs to perform their design below: 3. Does the proposed change involve a
function. The proposed acceptance criteria significant reduction in a margin of safety?
changes would not create any new failure 1. Does the proposed amendment involve
Response: No.
mechanisms, malfunctions, or accident a significant increase in the probability or
The proposed amendment would not
initiators not considered in the design and consequences of an accident previously
involve a significant reduction in a margin of
licensing bases. evaluated?
safety. More frequent monitoring of the
Therefore, the proposed amendment does Response: No.
capacity or performance of the VRLA
not create the possibility of a new or different The proposed amendment would not
batteries utilized in the replacement UPS
kind of accident from any previously increase the probability or consequences of
supplying power to critical reactor loads will
evaluated. an accident previously evaluated. The
ensure the UPS performs its design function
3. Does the proposed amendment involve proposed amendment modifies maintenance
and loss of battery capacity is detected early
a significant reduction in a margin of safety? requirements for emergency power systems
before safety margins are reduced.
Response: No. due to a change in battery technology used
The proposed EDG surveillance in commercially available UPS. The The NRC staff has reviewed the
requirement changes involve increased proposed amendment will assure the licensee’s analysis and, based on this
minimum voltage and frequency test reliability of the emergency power systems review, it appears that the three
acceptance criteria. The conduct of utilizing valve-regulated lead acid (VRLA) standards of 10 CFR 50.92(c) are
surveillance tests on safety related plant batteries by increasing the frequency of satisfied. Therefore, the NRC staff
equipment is a means of assuring that the performance testing as recommended by the
battery manufacturer and the IEEE (Institute
proposes to determine that the
equipment is capable of maintaining the amendment request involves no
margin of safety established in the safety of Electrical and Electronics Engineers). The
IEEE recommends the performance test significant hazards consideration.
analyses for the facility. The proposed Attorney for licensee: Melissa J.
amendment does not affect EDG performance Interval for VRLA batteries (IEEE–I 188)
as described in the design basis analyses, should not be greater than 25% of the Lieberman, Deputy Chief Counsel,
including the capability for the EDG to attain expected service life or two years, whichever National Institute of Standards and
and maintain required voltage and frequency is less. The expected lifespan of a VRLA Technology, 100 Bureau Drive,
for accepting and supporting plant safety battery is ten years so a two year testing Gaithersburg, MD 20899.
loads should an EDG start signal be received. interval was selected. More frequent NRC Branch Chief: Alexander Adams,
The proposed amendment does not introduce performance testing will ensure all the Jr.
changes to limits established in the accident station batteries used for emergency power
remain capable of supplying emergency Southern Nuclear Operating Company,
tkelley on DSK3SPTVN1PROD with NOTICES
analysis.
Therefore, the proposed amendment does electrical loads for a minimum of four hours Inc. (SNC), Docket Nos. 50–321 and 50–
not involve a significant reduction in a as required. The proposed amendment will 366, Edwin I. Hatch Nuclear Plant
margin of safety. also correct a typographical error and add the (HNP), Units 1 and 2, Appling County,
requirement in the Limiting Conditions for Georgia
The NRC staff has reviewed the Operations (LCO) for at least one of the two
licensee’s analysis and, based on this replacement UPS system batteries to be Date of amendment request: April 2,
review, it appears that the three available to operate the reactor. Each UPS 2015. A publicly-available version is in
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![Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices 38761
ADAMS under Accession No. Boiling Water Reactors,’’ dated June 2013, no change to the operability requirements for
ML15092A856. and of BWROG–TP–11–023–A, Revision 0 equipment assumed to operate for accident
Description of amendment request: (0900876.401, Revision 0–A), ‘‘Linear Elastic mitigation.
The proposed amendment modifies Fracture Mechanics Evaluation of General Therefore, the proposed change does not
Electric Boiling Water Reactor Water Level involve a significant reduction in a margin of
Technical Specifications (TS) section Instrument Nozzles for Pressure-Temperature safety.
1.0 (‘‘Definitions’’), Limiting Conditions Curve Evaluations,’’ dated May 2013, for Based upon the above, SNC concludes that
for Operation and Surveillance preparation of the pressure and temperature the proposed amendment presents no
Requirement Applicability, section 3.4.9 curves, and incorporates the guidance of significant hazards consideration under the
(‘‘RCS [Reactor Coolant System] TSTF–419–A, ‘‘Revise PTLR Definition and standards set forth in 10 CFR 50.92(c), and,
Pressure and Temperature (P/T) References in ISTS 5.6.6, RCS PTLR.’’ The accordingly, a finding of no significant
Limits’’), and section 5.0 HNP PTLRs have been developed based on hazards is justified.
(‘‘Administrative Controls’’) to delete the methodologies provided in BWROG–TP–
11–022–A, Revision 1 and BWROG–TP–11– The NRC staff has reviewed the
reference to the pressure and 023–A, Revision 0, and based on the template licensee’s analysis and, based on this
temperature curves, and to include provided in BWROG–TP–11–022–A, review, it appears that the three
reference to the Pressure and Revision 1. The HNP PTLRs meet all standards of 10 CFR 50.92(c) are
Temperature Limits Report (PTLR). This Conditions specified in the Safety Evaluation satisfied. Therefore, the NRC staff
change adopts the methodology of Reports (SERs) for BWROG–TP–11–022–A, proposes to determine that the
Boiling Water Reactor Owners Group Revision 1 and for BWROG–TP–11–023–A, amendment request involves no
(BWROG)–TP–11–022–A Revision 1 Revision 0.
significant hazards consideration.
(SIR–05–044, Revision 1–A), ‘‘Pressure- The NRC has established requirements in
Appendix G to 10 CFR 50 in order to protect
Attorney for licensee: Jennifer M.
Temperature Limits Report Buettner, Associate General Counsel,
the integrity of the reactor coolant pressure
Methodology for Boiling Water boundary (RCPB) in nuclear power plants. Southern Nuclear Operating Company,
Reactors,’’ dated June 2013, and of Additionally, the regulation in 10 CFR part 40 Inverness Center Parkway,
BWROG–TP–11–023–A, Revision 0 50, Appendix H, provides the NRC staff’s Birmingham, AL 35201.
(0900876.401, Revision 0–A), ‘‘Linear criteria for the design and implementation of NRC Branch Chief: Robert J.
Elastic Fracture Mechanics Evaluation reactor pressure vessel (RPV) material Pascarelli.
of General Electric Boiling Water surveillance programs for operating
Reactor Water Level Instrument Nozzles lightwater reactors. Implementing these NRC Southern Nuclear Operating Company,
for Pressure-Temperature Curve approved methodologies does not reduce the Inc., Docket Nos. 52–025 and 52–026,
Evaluations,’’ dated May 2013, for ability to protect the reactor coolant pressure Vogtle Electric
boundary as specified in Appendix G, nor
preparation of the pressure and Generating Plant (VEGP), Units 3 and
will this change increase the probability of
temperature (P–T) curves, and malfunction of plant equipment, or the 4, Burke County, Georgia
incorporates the guidance of Technical failure of plant structures, systems, or Date of amendment request: October
Specification Task Force (TSTF) TSTF– components. Incorporation of the new 7, 2014. A publicly-available version is
419–A, ‘‘Revise PTLR Definition and methodologies for calculating P–T curves, in ADAMS under Accession No.
References in Improved Standard and the relocation of the P–T curves from the ML14280A391.
Technical Specification (ISTS) 5.6.6, TS to the PTLR, provides an equivalent level Description of amendment request:
RCS PTLR.’’ The HNP PTLRs have been of assurance that the RCPB is capable of The proposed change would amend
performing its intended safety functions. Combined License Nos. NPF–91 and
developed based on the methodologies
Therefore, the proposed change does not NPF–92 for the VEGP, Units 3 and 4.
provided in BWROG–TP–11–022–A, involve a significant increase in the
Revision 1, and BWROG–TP–11–023–A, The requested amendment proposes to
probability or consequences of an accident
Revision 0, and based on the template previously evaluated.
modify the existing feedwater controller
provided in BWROG–TP–11–022–A, 2. Does the proposed change create the logic to allow the controller program to
Revision 1. possibility of a new or different kind of respond as required to various plant
Basis for proposed no significant accident from any accident previously transients while minimizing the
hazards consideration determination: evaluated? potential for false actuation. Because,
As required by 10 CFR 50.91(a), the Response: No. this proposed change requires a
licensee has provided its analysis of the The proposed change does not affect the departure from Tier 1 information in the
assumed accident performance of the RCPB, Westinghouse Advanced Passive 1000
issue of no significant hazards
nor any plant structure, system, or Design Control Document (DCD), the
consideration which is presented below: component previously evaluated. The
proposed change does not involve the
licensee also requested an exemption
1. Does the proposed change involve a
significant increase in the probability or installation of new equipment, and installed from the requirements of the Generic
consequences of an accident previously equipment is not being operated in a new or DCD Tier 1 in accordance with 10 CFR
evaluated? different manner. The change in 52.63(b)(1).
Response: No. methodology ensures that the RCPB remains Basis for proposed no significant
The proposed change modifies Edwin I. capable of performing its safety functions. No hazards consideration determination:
Hatch Nuclear Plant (HNP) Unit 1 and Unit set points are being changed which would As required by 10 CFR 50.91(a), the
2 Technical Specifications (TS) Section 1.0 alter the dynamic response of plant licensee has provided its analysis of the
(‘‘Definitions’’), Limiting Conditions for equipment. Accordingly, no new failure issue of no significant hazards
Operation and Surveillance Requirement modes are introduced which could introduce
consideration, which is presented
Applicability Section 3.4.9 (‘‘RCS Pressure the possibility of a new or different kind of
and Temperature (P/T) Limits’’), and Section accident from any previously evaluated. below:
5.0 (‘‘Administrative Controls’’), to delete 3. Does the proposed change involve a 1. Does the proposed amendment involve
tkelley on DSK3SPTVN1PROD with NOTICES
reference to the pressure and temperature significant reduction in a margin of safety? a significant increase in the probability or
curves, and to include reference to the Response: No. consequences of an accident previously
Pressure and Temperature Limits Report The proposed change does not affect the evaluated?
(PTLR). This change adopts the methodology function of the RCPB or its response during Response: No.
of BWROG–TP–11–022–A, Revision 1 (SIR– plant transients. There are no changes The proposed changes will modify the
05–044, Revision 1–A), ‘‘Pressure- proposed which alter the setpoints at which control logic for actuation of the startup
Temperature Limits Report Methodology for protective actions are initiated, and there is feedwater (SFW) pumps to support their
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![38762 Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices
defense-in-depth function of core decay heat instrumentation. These instruments, to be Emergency Plan for SSES to adopt the
removal. The instrumentation used for included as part of the D–RAP, will be held Nuclear Energy Institute’s (NEI’s)
actuation of the SFW pumps in their defense- to the same enhanced QA requirements as revised Emergency Action Level (EAL)
in-depth function are not initiators of any the current instruments and so neither safety,
performance, nor reliance will be reduced as
scheme described in NEI 99–01,
accident. The proposed control logic uses
different instrument tag numbers than the a part of this change. Furthermore, since the Revision 6, ‘‘Development of Emergency
current design. The instruments used for the D–RAP ensures consistency with the Action Levels for Non-Passive Reactors’’
actuation of this function exist as a part of Probabilistic Risk Assessment (PRA), the (ADAMS Accession No. ML12326A805),
the current design; therefore this proposed changes do not impact the PRA. The which was endorsed by the NRC, as
change does not require any additional proposed changes would not introduce a new documented in NRC letter dated March
instrumentation. These instruments, to be failure mode, fault, or sequence of events that 28, 2013 (ADAMS Accession No.
included as part of the Design Reliability could result in a radioactive material release.
ML12346A463).
Assurance Program (D–RAP), will be held to The proposed change does not alter the
design, configuration, or method of operation
On June 1, 2015, the NRC staff issued
the same enhanced quality assurance (QA)
requirements as the current instruments and of the plant beyond standard functional an amendment changing the name on
therefore neither safety, performance, nor capabilities of the equipment. the SSES license from PPL
reliance will be reduced as a part of this Therefore, the proposed amendment does Susquehanna, LLC, to Susquehanna
change. not create the possibility of a new or different Nuclear, LLC. This amendment was
Additionally, the proposed changes do not kind of accident from any accident issued subsequent to an order issued on
adversely affect any accident initiating event previously evaluated. April 10, 2015, to SSES, approving an
or component failure, thus accidents 3. Does the proposed amendment involve
indirect license transfer.
previously evaluated are not adversely a significant reduction in a margin of safety?
Response: No.
Basis for proposed no significant
affected. In the event of loss of offsite power hazards consideration determination:
that results in a loss of main feedwater The proposed changes will modify the
control logic for actuation of the startup As required by 10 CFR 50.91(a), the
(MFW) supply, the SFW pumps
automatically supply feedwater to the steam feedwater (SFW) pumps to support their licensee has provided its analysis of the
generators to cool down the reactor under defense-in-depth function of core decay heat issue of no significant hazards
emergency shutdown conditions. The removal. These changes will have no consideration, which is presented
standby source motor control center circuit negative impacts on the safety margin below:
powers each of the two SFW pumps and their associated with the design functions of the
SFW pumps. The proposed logic changes 1. Does the proposed change involve a
associated instruments and valves. The pump significant increase in the probability or
will only resolve the current conditions
discharge isolation valves are motor-operated consequences of any accident previously
associated with undesired start up signals for
and are normally closed and interlocked with evaluated?
the SFW pumps. The changes set forth in this
the SFW pumps. In the event of loss of offsite Response: No.
amendment correct the actuation logic of the
power, the onsite standby power supply The proposed changes to the EAL scheme
SFW pumps, so that the feedwater controller
diesel generators will power the SFW pumps. to adopt the NRC-endorsed guidance in NEI
logic is now aligned with the guidance
If both the normal ac power and the onsite 99–01, Revision 6, ‘‘Development of
provided in the Advanced Light Water
standby ac power are unavailable, these Emergency Action Levels for Non-Passive
Reactor Utility Requirements Document
valves will fail ‘‘as-is.’’ The pump suction (ALWR URD). In addition, the operation of Reactors,’’ do not reduce the capability to
header isolation valves are pneumatically the startup feedwater system function is not meet the emergency planning requirements
actuated. The main and startup feedwater credited to mitigate a design-basis accident. established in 10 CFR 50.47 and 10 CFR 50,
system (FWS) also has temperature Since there is no change to an existing design Appendix E. The proposed changes do not
instrumentation in the pump discharge that basis limit/criterion, design function, or reduce the functionality, performance, or
would permit monitoring of the SFW regulatory criterion no margin of safety is capability of the ERO [Emergency Response
temperature. This proposed change therefore reduced. Organization] to respond in mitigating the
has no impact on the ability of the AP1000 Therefore, the proposed amendment does consequences of any design basis accident.
plant to cool down under emergency not involve a significant reduction in a The probability of a reactor accident
shutdown conditions or during a loss of margin of safety. requiring implementation of Emergency Plan
offsite power event. EALs has no relevance in determining
No function used to mitigate a radioactive The NRC staff has reviewed the whether the proposed changes to the EALs
material release and no radioactive material licensee’s analysis and, based on this reduce the effectiveness of the Emergency
release source term is involved, thus the review, it appears that the three Plan. As discussed in Section I.D, ‘‘Planning
radiological releases in the accident analyses standards of 10 CFR 50.92(c) are Basis,’’ of NUREG–0654, Revision 1, ‘‘Criteria
are not adversely affected. satisfied. Therefore, the NRC staff for Preparation and Evaluation of
Therefore, the proposed amendment does proposes to determine that the Radiological Emergency Response Plans and
not involve a significant increase in the Preparedness in Support of Nuclear Power
amendment request involves no
probability or consequences of an accident Plants’’;
previously evaluated. significant hazards consideration. . . . The overall objective of emergency
2. Does the proposed amendment create Attorney for licensee: M. Stanford response plans is to provide dose savings
the possibility of a new or different kind of Blanton, Balch & Bingham LLP, 1710 (and in some cases immediate life saving) for
accident from any accident previously Sixth Avenue North, Birmingham, AL a spectrum of accidents that could produce
evaluated? 35203–2015. offsite doses in excess of Protective Action
Response: No. Acting NRC Branch Chief: Paul Guides (PAGs). No single specific accident
The proposed changes will modify the Kallan. sequence should be isolated as the one for
control logic for actuation of the startup which to plan because each accident could
feedwater (SFW) pumps to support their Susquehanna Nuclear, LLC, Docket Nos. have different consequences, both in nature
defense-in-depth function of core decay heat 50–387 and 50–388, Susquehanna and degree. Further, the range of possible
removal. The instrumentation used for Steam Electric Station (SSES), Units 1 selection for a planning basis is very large,
actuation of the SFW pumps in their defense- and 2, Luzerne County, Pennsylvania starting with a zero point of requiring no
tkelley on DSK3SPTVN1PROD with NOTICES
in-depth function are not initiators of any planning at all because significant offsite
accident. The proposed control logic uses
Date of amendment request: March radiological accident consequences are
different instrument tag numbers than the 19, 2015. A publicly-available version is unlikely to occur, to planning for the worst
current design. However, the instruments available in ADAMS under Package possible accident, regardless of its extremely
used for the actuation of this function already Accession No. ML15091A657. low likelihood . . .
exist as a part of the current design and so Description of amendment request: Therefore, risk insights are not considered
this change does not require any additional The amendment would revise the for any specific accident initiation or
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![Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices 38763
progression in evaluating the proposed Energy Supply, LLC, 835 Hamilton St., nucleate boiling ratio] safety limit without
changes. Suite 150, Allentown, PA 18101. the addition of analytical uncertainties does
The proposed changes do not involve any NRC Branch Chief: Douglas A. not alter the use of the analytical methods
physical changes to plant equipment or Broaddus. used to determine core operating limits that
systems, nor do they alter the assumptions of have been reviewed and approved by the
any accident analyses. The proposed changes Union Electric Company, Docket No. NRC. Removing analytical uncertainties from
do not adversely affect accident initiators or 50–483, Callaway Plant, Unit 1 the TS would allow the use of current topical
precursors nor do they alter the design (Callaway), Callaway County, Missouri reports to refine those uncertainties without
assumptions, conditions, and configuration having to submit an amendment to the
or the manner in which the plants are Date of amendment request: May 8, operating license, consistent with the intent
operated and maintained. The proposed 2015. A publicly-available version is in of WCAP–14483–A [‘‘Generic Methodology
changes do not adversely affect the ability of ADAMS under Accession No. for Expanded Core Operating Limits Report,’’
Structures, Systems, or Components (SSCs) ML15132A137. dated January 19, 1999; ADAMS Accession
to perform their intended safety functions in Description of amendment request: No. ML020430092]. Implementation of
mitigating the consequences of an initiating The amendment would change revisions to topical reports for Callaway Plant
event within the assumed acceptance limits. applications would still be reviewed in
Therefore, the proposed changes do not Technical Specification (TS) 2.1.1.1 and accordance with 10 CFR 50.59(c)(2)(viii) and,
involve a significant increase in the 5.6.5 to adopt NRC-approved where required, receive prior NRC review
probability or consequences of an accident Westinghouse Electric Company LLC and approval.
previously evaluated. topical report WCAP–14565–P–A, All accident analysis acceptance criteria
2. Does the proposed change create the Addendum 2–P–A, ‘‘Extended will continue to be met with the proposed
possibility of a new or different kind of Application of ABB–NV [Asea Brown changes. The proposed changes will not
accident from any accident previously Boveri N.V.] Correlation and Modified affect the source term, containment isolation,
evaluated? ABB–NV Correlation WLOP or radiological release assumptions used in
Response: No. evaluating the radiological consequences of
The proposed changes to the EAL scheme [Westinghouse Low Pressure] for PWR an accident previously evaluated. The
to adopt the NRC-endorsed guidance in NEI [Pressurized-Water Reactor] Low proposed changes will not alter any
99–01, Revision 6 do not involve any Pressure Applications,’’ April 2008 assumptions or change any mitigation actions
physical changes to plant systems or (proprietary). in the radiological consequence evaluations
equipment. The proposed changes do not Basis for proposed no significant in the FSAR [Final Safety Analysis Report].
involve the addition of any new plant hazards consideration determination: The applicable radiological dose acceptance
equipment. The proposed changes will not As required by 10 CFR 50.91(a), the criteria will continue to be met.
alter the design configuration, or method of licensee has provided its analysis of the Therefore, the proposed changes do not
operation of plant equipment beyond its involve a significant increase in the
normal functional capabilities. All ERO
issue of no significant hazards
probability or consequences of an accident
functions will continue to be performed as consideration, which is presented
previously evaluated.
required. The proposed changes do not create below: 2. Does the proposed change create the
any new credible failure mechanisms, 1. Does the proposed change involve a possibility of a new or different kind of
malfunctions, or accident initiators. significant increase in the probability or accident from any accident previously
Therefore, the proposed changes do not consequences of an accident previously evaluated?
create the possibility of a new or different evaluated? Response: No.
kind of accident from those that have been Response: No. The ABB–NV correlation was originally
previously evaluated. Overall protection system performance will developed for Combustion Engineering fuel
3. Does the proposed change involve a remain within the bounds of the accident designs, and has also been qualified and
significant reduction in a margin of safety? analyses since there are no design changes. licensed for Westinghouse fuel applications
Response: No. The design of the reactor trip system (RTS) for the fuel region below the first mixing
The proposed changes to the EAL scheme instrumentation will be unaffected, and thus, vane grid where the W–3 correlation is
to adopt the NRC-endorsed guidance in NEI the protection system will continue to currently applied. The WLOP correlation is
99–01, Revision 6 do not alter or exceed a function in a manner consistent with the developed for DNBR calculations at low
design basis or safety limit. There is no plant design basis. All applicable design, pressure conditions. The W–3A correlations,
change being made to safety analysis material, and construction standards will which are based exclusively on DNB
assumptions, safety limit, or limiting safety continue to be maintained. [departure from nucleate boiling] data from
system settings that would adversely affect The proposed changes will not affect any rod bundle tests, have a wider applicable
plant safety as a result of the proposed assumptions regarding accident initiators or range and are more accurate than the W–3
changes. There are no changes to setpoints or precursors nor adversely alter the design correlation, leading to increased DNB margin
environmental conditions of any SSC or the assumptions, conditions, and configuration in the plant safety analyses. The NRC-
manner in which any SSC is operated. of the facility or the intended manner in approved ABB–NV and WLOP correlation
Margins of safety are unaffected by the which the plant is operated and maintained. 95/95 DNBR limits with the VIPRE–W code
proposed changes to adopt the NEI 99–01, The proposed changes will not alter or are 1.13 and 1.18, respectively.
Revision 6 EAL scheme guidance. The prevent the ability of structures, systems, and The proposed change does not create the
applicable requirements of 10 CFR 50.47 and components (SSCs) from performing their possibility of a new or different kind of
10 CFR 50, Appendix E will continue to be intended functions to mitigate the accident from any accident previously
met. consequences of an initiating event within evaluated, as the change is simply allowing
Therefore, the proposed changes do not the assumed acceptance limits. the use of more accurate correlations when
involve any reduction in a margin of safety. The proposed changes do not physically evaluating DNBR. The change does not
The NRC staff has reviewed the alter safety-related systems nor affect the way involve any physical changes to the facility.
licensee’s analysis and, based on this in which safety-related systems perform their Likewise, revising TS 2.1.1.1 to present the
review, it appears that the three functions. TS 5.6.5.b continues to ensure that DNBR safety limit calculated using the WRB–
the analytical methods used to determine the 2 methodology, without uncertainties being
tkelley on DSK3SPTVN1PROD with NOTICES
standards of 10 CFR 50.92(c) are
core operating limits meet NRC reviewed and applied, does [sic] not introduce any new or
satisfied. Therefore, the NRC staff approved methodologies. TS 5.6.5.c, different failure mode from what has been
proposes to determine that the unchanged by this amendment application, previously been evaluated. The change does
amendment request involves no will continue to ensure that applicable limits not involve any change to a methodology,
significant hazards consideration. of the safety analyses are met. including how uncertainties are calculated
Attorney for licensee: Damon D. Obie, The proposed change to TS 2.1.1.1 to and accounted for, nor does it involve any
Associate General Counsel, Talen specify only the true DNBR [departure from physical change to the facility.
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![38764 Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices
Collectively, and based on the above, the Attorney for licensee: John O’Neill, Therefore, the proposed amendment does
proposed changes do not create the Esq., Pillsbury Winthrop Shaw Pittman not create the possibility of a new or different
possibility of a new or different kind of LLP, 2300 N Street NW., Washington, kind of accident from any accident
accident from any accident previously DC 20037. previously evaluated.
evaluated. NRC Branch Chief: Michael T. 3. Does the proposed change involve a
3. Does the proposed change involve a significant reduction in a margin of safety?
Markley. Response: No.
significant reduction in a margin of safety?
Response: No. Virginia Electric and Power Company, The change affects the NAPS EALs, but
The ABB–NV correlation was originally Docket Nos. 50–338 and 50–339, North does not alter any of the requirements of the
developed for Combustion Engineering fuel Anna Power Station, Unit Nos. 1 and 2 Operating License or the Technical
designs, and has also been qualified and Specifications. The proposed change does
(NAPS), Louisa County, Virginia
licensed for Westinghouse fuel applications not affect any of the assumptions used in the
for the fuel region below the first mixing Date of amendment request: May 4, accident analysis, nor does it affect any
vane grid where the W–3 correlation is 2015. A publicly-available version is in operability requirements for equipment
currently applied. The WLOP correlation is ADAMS under Accession No. important to plant safety.
developed for DNBR calculations at low ML15131A026. Therefore, the proposed change will not
pressure conditions. The W–3A correlations, Description of amendment request: result in a significant reduction in the margin
which are based exclusively on DNB data The proposed license amendment of safety.
from rod bundle tests, have a wider would revise an incorrect equipment The NRC staff has reviewed the
applicable range and are more accurate than
the W–3 correlation, leading to increased
mark number, due to an administrative licensee’s analysis and, based on this
DNB margin in the plant safety analyses. error in the current Emergency Action review, it appears that the three
Therefore, the proposed change does not Levels (EAL). The amendment would standards of 50.92(c) are satisfied.
involve a significant reduction in a margin of correct the equipment mark number Therefore, the NRC staff proposes to
safety. from the ‘‘GW–RI–178–1 Process Vent determine that the amendment request
The currently listed Safety Limit in TS Normal Range’’ monitor to the ‘‘VG–RI– involves no significant hazards
2.1.1.1 for DNBR of 1.22 is calculated with 180–1 Vent Stack B Normal Range’’ consideration.
some uncertainties statistically combined monitor for Initiating Condition (IC) Attorney for licensee: Lillian M.
into the 1.17 value calculated using the RA2, EAL RA2.1. Cuoco, Senior Counsel, Dominion
WRB–2 methodology. These uncertainties are
Basis for proposed no significant Resources Services, Inc., 120 Tredegar
combined using the RTDP [Revised Thermal
Design Procedure] methodology described in hazards consideration determination: Street, RS–2, Richmond, VA 23219.
WCAP–11397–P–A [‘‘Revised Thermal As required by 10 CFR 50.91(a), the NRC Branch Chief: Robert J.
Design Procedure,’’ April 1989 (proprietary)]. licensee has provided its analysis of the Pascarelli.
Callaway FSAR Section 4.4.1.1 discusses issue of no significant hazards
III. Notice of Issuance of Amendments
which uncertainties are statistically consideration, which is presented
combined into the correlation limit.
to Facility Operating Licenses and
below:
Revising TS 2.1.1.1 to present the DNBR Combined Licenses
1. Does the proposed change involve a
safety limit calculated using the WRB–2 During the period since publication of
significant increase in the probability or
methodology, without uncertainties being the last biweekly notice, the
consequences of an accident previously
applied, does not represent a change in Commission has issued the following
evaluated?
methodology, but rather allows for changes
Response: No. amendments. The Commission has
in calculated uncertainties using This administrative change affects the
methodologies previously approved for determined for each of these
NAPS EALs, but does not alter any of the amendments that the application
Callaway without requiring a license requirements of the Operating License or the
amendment. The proposed TS 2.1.1.1 complies with the standards and
Technical Specifications. The proposed
revision does not represent a change in change does not modify any plant equipment requirements of the Atomic Energy Act
methodology for performing analyses. and does not impact any failure modes that of 1954, as amended (the Act), and the
The proposed changes do not eliminate could lead to an accident. Additionally, the Commission’s rules and regulations.
any surveillances or alter the frequency of proposed change has no effect on the The Commission has made appropriate
surveillances required by the Technical consequences of any analyzed accident since findings as required by the Act and the
Specifications. The nominal RTS and ESFAS the change does not affect any equipment Commission’s rules and regulations in
[engineered safety features actuation system] related to accident mitigation.
trip setpoints (as well as the associated 10 CFR chapter I, which are set forth in
Based on this discussion, the proposed
allowable values) will remain unchanged. amendment does not increase the probability
the license amendment.
None of the acceptance criteria for any or consequences of an accident previously A notice of consideration of issuance
accident analysis will be changed. evaluated. of amendment to facility operating
As there is no change to the source term, 2. Does the proposed change create the license or combined license, as
radiological release, or [dose] mitigation possibility of a new or different kind of applicable, proposed no significant
functions assumed in the accident analysis, accident from any accident previously hazards consideration determination,
the proposed changes have no impact on the evaluated? and opportunity for a hearing in
radiological consequences of a design basis Response: No. connection with these actions, was
accident. The change affects the NAPS EALs by
Based on the above, the proposed changes correcting an incorrect radiation monitor published in the Federal Register as
do not involve a significant reduction in a reference, but does not alter any of the indicated.
margin of safety. requirements of the Operating License or the Unless otherwise indicated, the
Technical Specifications. It does not modify Commission has determined that these
The NRC staff has reviewed the any plant equipment and there is no impact amendments satisfy the criteria for
tkelley on DSK3SPTVN1PROD with NOTICES
licensee’s analysis and, based on this on the capability of the existing equipment categorical exclusion in accordance
review, it appears that the three to perform its intended functions. No system with 10 CFR 51.22. Therefore, pursuant
standards of 10 CFR 50.92(c) are setpoints are being modified. No new failure to 10 CFR 51.22(b), no environmental
modes are introduced by the proposed
satisfied. Therefore, the NRC staff change. The proposed amendment does not impact statement or environmental
proposes to determine that the introduce an accident initiator or any assessment need be prepared for these
amendment request involves no malfunctions that would cause a new or amendments. If the Commission has
significant hazards consideration. different kind of accident. prepared an environmental assessment
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![38766 Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices
Exelon Generation Company, LLC, in the associated P/T figures. No significant hazards consideration
Docket Nos. 50–317 and 50–318, Calvert Additionally editorial changes are comments received: No.
Cliffs Nuclear Power Plant, Unit Nos. 1 proposed related to the P/T figures
Florida Power and Light Company, et
and 2, Calvert County, Maryland including clarifications and updates to
al., Docket Nos. 50–335 and 50–389, St.
Date of amendment request: the associated titles, labeling, and notes. Lucie Plant, Unit Nos. 1 and 2, St. Lucie
November 13, 2013, as supplemented by Date of issuance: June 12, 2015.
County, Florida
Effective date: As of the date of
letter dated June 13, 2014. Date of amendment request: February
issuance and shall be implemented
Brief description of amendments: The 20, 2014, as supplemented by letters
within 90 days.
amendments revised Technical dated December 11, 2014, and January
Amendment No.: 168. A publicly-
Specification (TS) 3.4.9, ‘‘Pressurizer,’’ 13, January 28, April 18, and May 19,
available version is in ADAMS under
TS 3.6.6, ‘‘Containment Spray and 2015.
Accession No. ML15141A482;
Cooling Systems,’’ TS 3.6.8, ‘‘Iodine Brief description of amendments: The
documents related to this amendment
Removal System,’’ TS 3.7.8, ‘‘Control amendments revised the Technical
are listed in the Safety Evaluation
Room Emergency Ventilation System,’’ Specifications (TSs) by relocating
enclosed with the amendment.
and TS 3.7.12, ‘‘Penetration Room Facility Operating License No. NPF– specific surveillance frequency
Exhaust Ventilation System’’ to provide 58: This amendment revised the TSs requirements to a licensee-controlled
a short completion time to restore an and License. program with implementation of
inoperable system for conditions under Date of initial notice in Federal Nuclear Energy Institute (NEI) 04–10,
which existing TSs require a plant Register: September 30, 2014 (79 FR ‘‘Risk-Informed Technical Specification
shutdown. 58817). Initiative 5b, Risk-Informed Method for
Date of issuance: January 14, 2015. The February 27, 2015, supplement Control of Surveillance Frequencies’’
Effective date: As of the date of contained clarifying information and (ADAMS Accession No. ML071360456).
issuance and shall be implemented did not change the NRC staff’s initial The NEI 04–10 methodology provides
within 60 days of issuance. proposed finding of no significant reasonable acceptance guidelines and
Amendment Nos.: 309 and 287. A hazards consideration. methods for evaluating the risk increase
publicly-available version is in ADAMS The Commission’s related evaluation of proposed changes to surveillance
under Accession No. ML14307A842; of the amendment is contained in a frequencies, consistent with Regulatory
documents related to these amendments Safety Evaluation dated June 12, 2015. Guide 1.177, ‘‘An Approach for Plant-
are listed in the Safety Evaluation No significant hazards consideration Specific, Risk-Informed
enclosed with the amendments. comments received: No. Decisionmaking: Technical
Renewed Facility Operating License Specifications’’ (ADAMS Accession No.
Nos. DPR–53 and DPR–69: The FirstEnergy Nuclear Operating
ML003740176). The changes are
amendments revised the Licenses and Company, Docket No. 50–440, Perry
consistent with NRC-approved
TSs. Nuclear Power Plant, Unit No. 1, Lake
Technical Specification Task Force
Date of initial notice in Federal County, Ohio
(TSTF) Standard Technical
Register: July 22, 2014 (79 FR 42548). Date of amendment request: Specifications Change TSTF–425,
The supplement dated June 13, 2014, September 12, 2014. Revision 3, ‘‘Relocate Surveillance
provided additional information that Brief description of amendment: The Frequencies to Licensee Control—
clarified the application, did not expand amendment modified the Technical RITSTF [Risk Informed Technical
the scope of the application as originally Specification (TS) definition of Specifications Task Force] Initiative 5b’’
noticed, and did not change the NRC SHUTDOWN MARGIN (SDM) to require (ADAMS Accession No. ML090850642).
staff’s original proposed no significant calculation of SDM at the reactor The Federal Register notice published
hazards consideration determination. moderator temperature corresponding to on July 6, 2009 (74 FR 31996),
The Commission’s related evaluation the most reactive state throughout the announced the availability of TSTF–
of the amendments is contained in a operating cycle. The changes are 425, Revision 3. The amendments also
Safety Evaluation dated January 14, consistent with the approved Technical include editorial changes to the TSs,
2015. Specification Task Force (TSTF) administrative deviations from TSTF–
No significant hazards consideration Traveler, TSTF–535, Revision 0. 425, and other changes resulting from
comments received: No. Date of issuance: June 23, 2015. differences between the St. Lucie Plant
Effective date: As of the date of TSs and the TSs on which TSTF–425
FirstEnergy Nuclear Operating
issuance and shall be implemented was based.
Company, et al., Docket No. 50–440,
within 90 days of issuance. Date of issuance: June 22, 2015.
Perry Nuclear Power Plant, Unit No. 1,
Amendment No.: 169. A publicly- Effective date: As of the date of
Lake County, Ohio
available version is in ADAMS under issuance and shall be implemented
Date of application for amendment: Accession No. ML15160A028; within 90 days of issuance.
June 23, 2014, as supplemented by a documents related to this amendment Amendment Nos.: 223 and 173. A
letter dated February 27, 2015. are listed in the Safety Evaluation publicly-available version is in ADAMS
Brief description of amendment: The enclosed with the amendment. under Accession No. ML15127A066;
amendment updated the technical Facility Operating License No. NPF– documents related to these amendments
specification (TS) pressure and 58: The amendment revised the Facility are listed in the Safety Evaluation (SE)
temperature (P/T) figures using an NRC Operating License and Technical enclosed with the amendments.
tkelley on DSK3SPTVN1PROD with NOTICES
approved methodology to adjust the Specifications. Renewed Facility Operating License
P/T limit curves for the previously Date of initial notice in Federal Nos. DPR–67 and NPF–16: Amendments
missing data, addresses the reactor Register: November 25, 2014 (79 FR revised the TSs.
coolant system (RCS) vacuum condition 70216). Date of initial notice in Federal
that can occur under certain conditions, The Commission’s related evaluation Register: The NRC staff initially made
and aligns the heatup/cooldown of the amendment is contained in a a proposed determination that the
requirements of the TS with the limits Safety Evaluation dated June 23, 2015. amendment request dated February 20,
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![Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices 38767
2014, involved no significant hazards Amendment Nos.: 198, 308, 290. A Facility Operating License Nos. DPR–
consideration (July 22, 2014, 79 FR publicly-available version is in ADAMS 13, NPF–10, and NPF–15: The
42550). By letters dated December 11, under Accession No. ML15141A271; amendments revised the emergency
2014, and January 13, 2015, the licensee documents related to these amendments plan.
provided clarifying information that did are listed in the Safety Evaluation Date of initial notice in Federal
not expand the scope of the application enclosed with the amendments. Register: December 23, 2014 (79 FR
and did not change the NRC staff’s Renewed Facility Operating License 77049). The supplemental letters dated
original proposed no significant hazards Nos. NPF–57, DPR–70 and DPR–75: The
March 17, 2015, and April 29, 2015,
consideration (NSHC) determination, as amendments revised the Renewed
provided additional information that
published in the Federal Register on Facility Operating Licenses and EPPs.
clarified the application, did not expand
July 22, 2014 (79 FR 44550). Date of initial notice in Federal
Subsequently, by letter dated January the scope of the application as originally
Register: April 14, 2015 (80 FR 20024).
28, 2015, the licensee supplemented its noticed, and did not change the staff’s
The supplemental letter dated April 9,
amendment request with a proposed 2015, provided additional information original proposed no significant hazards
change that expanded the scope of the that clarified the application, did not consideration determination as
request. Therefore, the NRC published a expand the scope of the application as published in the Federal Register.
second proposed NSHC determination originally noticed, and did not change The Commission’s related evaluation
in the Federal Register on March 3, the staff’s original proposed no of the amendments is contained in a
2015 (80 FR 11477), which superseded significant hazards consideration Safety Evaluation dated June 5, 2015.
the notice dated July 22, 2014 (79 FR determination as published in the No significant hazards consideration
44550). The licensee’s supplements Federal Register. comments received: No.
dated April 18, 2015, and May 19, 2015, The Commission’s related evaluation
provided clarifying information that did of the amendments is contained in a Tennessee Valley Authority, Docket No.
not expand the scope of the submittal Safety Evaluation dated June 17, 2015. 50–260, Browns Ferry Nuclear Plant,
and did not change the NRC staff’s No significant hazards consideration Unit 2, Limestone County, Alabama
proposed NSHC determination, as comments received: Yes.
published in the notice dated March 3, The Commission’s related evaluation Date of amendment request: June 19,
2015 (80 FR 11477). of the public comments is contained in 2014, as supplemented by letter dated
The Commission’s related evaluation the safety evaluation dated June 17, December 2, 2014.
of the amendments is contained in an 2015. Brief description of amendment: The
SE dated June 22, 2015. amendment revised the Technical
Southern California Edison Company, et
No significant hazards consideration Specification (TS) 3.4.9, ‘‘RCS [Reactor
al., Docket Nos. 50–206, 50–361, 50–
comments received: No. Coolant System] Pressure and
362, and 72–041, San Onofre Nuclear
PSEG Nuclear LLC, Docket Nos. 50–354, Generating Station (SONGS), Units 1, 2, Temperature (P/T) Limits,’’ and Figures
50–272 and 50–311, Hope Creek and 3, and the Independent Spent Fuel 3.4.9–1 through 3.4.9–2. The P/T limits
Generating Station and Salem Nuclear Storage Installation, San Diego County, are based on proprietary topical report
Generating Station, Unit Nos. 1 and 2, California NEDC–33178P–A, Revision 1, ‘‘GE
Salem County, New Jersey [General Electric] Hitachi Nuclear
Date of amendment request: March Energy Methodology for Development of
Date amendment request: December 31, 2014, as supplemented by letters Reactor Pressure Vessel Pressure-
9, 2014, as supplemented by letter dated dated October 21, 2014, March 17, 2015, Temperature Curves.’’ NEDO–33178–A,
April 9, 2015. and April 29, 2015. Revision 1, is the non-proprietary
Brief description of amendments: The Brief description of amendment: The version of the NRC-approved topical
amendments revised the PSEG Nuclear amendments revised the SONGS
report.
LLC (PSEG) Environmental Protection emergency plan to reflect the low
Plans (non-radiological) (EPPs), likelihood of any credible accident at Date of issuance: June 2, 2015.
contained in Appendix B to the the facility in its permanently shutdown Effective date: As of the date of
renewed facility operating licenses for and defueled condition that could result issuance and shall be implemented
Hope Creek Generating Station and in radiological releases requiring offsite within 60 days of issuance.
Salem Nuclear Generating Station, Units protective measures and how, in the
Amendment No.: 314. A publicly
1 and 2, to clarify that PSEG must unlikely event of certain severe, beyond-
available version is in ADAMS under
adhere to the currently applicable design-basis accidents, sufficient time
biological opinion issued by the would be available to initiate Accession No. ML15065A049;
National Marine Fisheries Service. The appropriate mitigating actions, and if documents related to this amendment
amendments also simplify the Aquatic needed, for offsite authorities to are listed in the Safety Evaluation
Monitoring section of the EPPs, modify implement protective actions to protect enclosed with the amendment.
reporting requirements related to New the public health and safety. Renewed Facility Operating License
Jersey Pollutant Discharge Elimination Date of issuance: June 5, 2015. No. DPR–52: The amendment revised
System permits, modify the criteria for Effective date: As of the date of the Facility Operating License and TSs.
reporting Unusual or Important issuance and shall be implemented Date of initial notice in Federal
Environmental Events, and remove the within 60 days of issuance. Register: February 3, 2015 (80 FR
tkelley on DSK3SPTVN1PROD with NOTICES
requirement for PSEG to submit an Amendment Nos.: Unit 1—167; Unit 5819).
Annual Environmental Operating 2—229; Unit 3—222. A publicly-
Report. available version is in ADAMS under The Commission’s related evaluation
Date of issuance: June 17, 2015. Accession No. ML15126A461; of the amendment is contained in a
Effective date: As of the date of documents related to these amendments Safety Evaluation dated June 2, 2015.
issuance and shall be implemented are listed in the Safety Evaluation No significant hazards consideration
within 60 days. enclosed with the amendments. comments received: No.
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![38770 Federal Register / Vol. 80, No. 129 / Tuesday, July 7, 2015 / Notices
Time, Monday through Friday, Energy Northwest, Docket No. 50–397, Tennessee Valley Authority, Docket No.
excluding government holidays. Columbia Generating Station, Benton 50–390, Watts Bar Nuclear Plant, Unit 1
Participants who believe that they County, Washington (WBN–1), Rhea County, Tennessee
have a good cause for not submitting Date of application for amendment: Date of amendment request: May 29,
documents electronically must file an May 15, 2015, as supplemented by letter 2015, as supplemented by letter dated
exemption request, in accordance with dated May 19, 2015. June 5, 2015.
10 CFR 2.302(g), with their initial paper Brief description of amendment: The Description of amendment request:
filing requesting authorization to amendment extended the The amendment provided a one-time
continue to submit documents in paper implementation period for Amendment change to Technical Specification Table
No. 232, License Amendment Request 3.3.4–1, Function 4a, ‘‘Reactor Coolant
format. Such filings must be submitted
for Changing Technical Specification System (RCS) Hot Leg Temperature
by: (1) first class mail addressed to the
Table 3.3.1.1–1 Function 7, ‘‘Scram Indication,’’ to permit the temperature
Office of the Secretary of the indication for RCS Loop 4 to be
Commission, U.S. Nuclear Regulatory Discharge Volume Water Level—High,’’
which was issued on March 27, 2015 inoperable for the remainder of WBN–
Commission, Washington, DC 20555– 1 Operating Cycle 13.
0001, Attention: Rulemaking and (ADAMS Accession No. ML15063A010).
Date of issuance: June 12, 2015.
Adjudications Staff; or (2) courier, Amendment No. 232 was effective as of Effective date: June 12, 2015.
express mail, or expedited delivery the date of issuance (i.e., on March 27, Amendment No.: 100. A publicly-
service to the Office of the Secretary, 2015), and was required to be available version is in ADAMS under
Sixteenth Floor, One White Flint North, implemented prior to restarting from Accession No. ML15160A407;
refueling outage R–22, scheduled for documents related to this amendment
11555 Rockville Pike, Rockville,
spring 2015. Amendment No. 235 are listed in the Safety Evaluation
Maryland, 20852, Attention:
extends the implementation period for enclosed with the amendment.
Rulemaking and Adjudications Staff. Amendment No. 232 to prior to
Participants filing a document in this Facility Operating License No. (NPF–
restarting from refueling outage R–23, 90): The amendment revised the
manner are responsible for serving the scheduled for spring 2017. Technical Specifications.
document on all other participants. Date of issuance: June 11, 2015. Public comments requested as to
Filing is considered complete by first- Effective date: As of its date of proposed no significant hazards
class mail as of the time of deposit in issuance and shall be implemented consideration (NSHC): Yes. (The
the mail, or by courier, express mail, or prior to restarting from refueling outage Advocate & Democrat and The Herald-
expedited delivery service upon R–23, scheduled for spring 2017. News on June 7 and June 10, 2015 as
depositing the document with the Amendment No.: 235. A publicly- well as The Daily Post-Athenian on June
provider of the service. A presiding available version is in ADAMS under 5 and June 8, 2015.) The notice
officer, having granted an exemption Accession No. ML15154A800; provided an opportunity to submit
request from using E-Filing, may require documents related to this amendment comments on the Commission’s
a participant or party to use E-Filing if are listed in the Safety Evaluation proposed NSHC determination. No
the presiding officer subsequently enclosed with the amendment. comments were received.
determines that the reason for granting Renewed Facility Operating License The supplemental letter dated June 5,
the exemption from use of E-Filing no No. NPF–21: Amendment revised the 2015, provided additional information
longer exists. Renewed Facility Operating License to that clarified the application, did not
extend the implementation date of expand the scope of the application as
Documents submitted in adjudicatory originally noticed, and did not change
proceedings will appear in the NRC’s Amendment No. 232, issued on March
the staff’s original proposed no
electronic hearing docket which is 27, 2015, to prior to restarting from
significant hazards consideration
available to the public at http:// refueling outage R–23, scheduled for
determination as published in The
ehd1.nrc.gov/ehd/, unless excluded spring 2017.
Advocate & Democrat and The Herald-
pursuant to an order of the Commission, Public comments requested as to News on June 7, and June 10, 2015, as
or the presiding officer. Participants are proposed no significant hazards well as The Daily Post-Athenian, on
requested not to include personal consideration (NSHC): Yes. Public June 5, and June 8, 2015.
privacy information, such as social notice of the proposed amendment was The Commission’s related evaluation
published in the Tri-City Herald, located of the amendment, finding of exigent
security numbers, home addresses, or
in in Kennewick, Washington, from circumstances, state consultation, and
home phone numbers in their filings,
June 2 through June 4, 2015. The notice NSHC determination are contained in a
unless an NRC regulation or other law provided an opportunity to submit
requires submission of such safety evaluation dated June 12, 2015.
comments on the Commission’s Attorney for licensee: General
information. However, in some proposed NSHC determination. No Counsel, Tennessee Valley Authority,
instances, a request to intervene will comments were received. 400 West Summit Hill Drive, ET 11A,
require including information on local The Commission’s related evaluation Knoxville, TN 37902.
residence in order to demonstrate a of the amendment, finding of exigent NRC Branch Chief: Jessie F.
proximity assertion of interest in the circumstances, state consultation, Quichocho.
proceeding. With respect to copyrighted public comments, and final NSHC Dated at Rockville, Maryland, this 29th day
works, except for limited excerpts that determination are contained in a Safety of June 2015.
serve the purpose of the adjudicatory Evaluation dated June 11, 2015.
tkelley on DSK3SPTVN1PROD with NOTICES
For the Nuclear Regulatory Commission.
filings and would constitute a Fair Use Attorney for licensee: William A. George A. Wilson,
application, participants are requested Horin, Esq., Winston & Strawn, 1700 K Deputy Director, Division of Operating
not to include copyrighted materials in Street, NW., Washington, DC 20006– Reactor Licensing, Office of Nuclear Reactor
their submission. 3817. Regulation.
NRC Branch Chief: Michael T. [FR Doc. 2015–16539 Filed 7–6–15; 8:45 am]
Markley. BILLING CODE 7590–01–P
VerDate Sep<11>2014 20:31 Jul 06, 2015 Jkt 235001 PO 00000 Frm 00110 Fmt 4703 Sfmt 9990 E:\FR\FM\07JYN1.SGM 07JYN1](https://thefederalregister.org/doc/80_FR_38885/page/15.svg)
Document Created: 2015-12-15 13:27:22
Document Modified: 2015-12-15 13:27:22
| 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 August 6, 2015. A request for a hearing must be filed by September 8, 2015. | |
| Contact | Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: [email protected] | |
| FR Citation | 80 FR 38756 |
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