Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
Federal Register, Volume 79 Issue 236 (Tuesday, December 9, 2014)
Federal Register Volume 79, Number 236 (Tuesday, December 9, 2014)
Notices
Pages 73106-73115
From the Federal Register Online via the Government Printing Office www.gpo.gov
FR Doc No: 2014-28704
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NUCLEAR REGULATORY COMMISSION
NRC-2014-0260
Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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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 November 13, 2014 to November 26, 2014. The last biweekly notice was published on November 25, 2014.
DATES: Comments must be filed by January 8, 2015. A request for a hearing must be filed by February 9, 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-2014-0260. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. 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: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Mable Henderson, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-3760, email: mable.henderson@nrc.gov.
SUPPLEMENTARY INFORMATION:
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Obtaining Information and Submitting Comments
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Obtaining Information
Please refer to Docket ID NRC-2014-0260 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-2014-0260.
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 pdr.resource@nrc.gov. 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.
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Submitting Comments
Please include Docket ID NRC-2014-0260 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS.
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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.
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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.
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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/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.
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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
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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 hearing.docket@nrc.gov, 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 MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, 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.
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Entergy Nuclear Vermont Yankee, LLC., and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 12, 2014. A publicly-available version is in ADAMS under Accession No. ML14168A302.
Description of amendment request: The proposed amendment would revise the site emergency plan (SEP) and Emergency Action Level (EAL) scheme to reflect the reduced scope of offsite and onsite emergency planning and the significantly reduced spectrum of credible accidents that can occur for the permanently defueled condition.
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:
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Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the emergency plan and EAL scheme do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not affect accident initiators or precursors, nor does it alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and emergency response organization (ERO) to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently defueled condition.
The probability of occurrence of previously evaluated accidents is not increased, since most previously analyzed accidents can no longer occur and the probability of the few remaining credible accidents are unaffected by the proposed amendment.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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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 reduce the scope of the emergency plan and EAL scheme commensurate with the hazards associated with a permanently shutdown and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new or different kinds of accident initiators are created.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the emergency plan and EAL scheme and do not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised SEP will continue to provide the necessary response staff with the proposed changes.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.
NRC Branch Chief: Douglas A. Broaddus.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3 (WF3), St. Charles Parish, Louisiana
Date of amendment request: August 28, 2014. A publicly-available version is in ADAMS under Accession No. ML14241A305.
Description of amendment request: The amendment would revise the 10-year frequency of the Type A or Integrated Leak Rate Test (ILRT) that is required by Technical Specification (TS) 6.15, ``Containment Leakage Rate Testing Program,'' to be extended to 15 years on a permanent basis.
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:
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Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves changes to the WF3 Containment Leakage Rate Testing Program. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The primary reactor building function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the reactor building itself and the testing requirements to periodically demonstrate the integrity of the reactor building exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.
The integrity of the reactor building is subject to two (2) types of failure mechanisms which can be categorized as (1) activity based and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that the reactor building containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the reactor building itself combined with the reactor building inspections performed in accordance with ASME American Society for Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, the Maintenance Rule and regulatory commitments serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed amendment does not involve a significant increase in the consequences of an accident previously evaluate.
The proposed amendment adopts the NRC-accepted guidelines of Nuclear Energy Institute (NEI) 94-01, Revision 2-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' October 2008 (ADAMS Accession No. ML100620847) for development of the WF3 performance-based testing program. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than values assumed in the plant safety analyses. The potential consequences of extending the ILRT interval to fifteen (15) years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year within fifty (50) miles resulting from design basis accidents was estimated to be acceptably small and determined to be within the guidelines published in RG Regulatory Guide 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. WF3 has determined that the increase in Conditional Containment Failure Probability due to the proposed change would be very small. Therefore, it is
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concluded that the proposed amendment does not significantly increase the consequences of an accident previously evaluated.
Based on the above discussion, it is concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2-A, for the development of the WF3 performance-
based leakage testing program, and establishes a fifteen (15) year interval for the performance of the reactor building ILRT. The reactor building and the testing requirements to periodically demonstrate the integrity of the reactor building exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
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Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2-A, for the development of the WF3 performance-
based leakage testing program, and establishes a fifteen (15) year interval for the performance of the containment ILRT. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the Reactor Building Leakage Rate Testing Program, as defined in the TS, ensure that the degree of the reactor building structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. The overall reactor building leakage rate limit specified by the TS is maintained, and the Type A, Type B, and Type C containment leakage tests will be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 2-A.
Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment will not degrade in a manner that is not detectable by an ILRT. A risk assessment using the current WF3 risk model concluded that extending the ILRT test interval from ten (10) years to fifteen (15) years results in a very small change to the WF3 risk profile.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, (TMI-1) Dauphin County, Pennsylvania
Date of amendment request: October 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14304A083.
Description of amendment request: The amendment would change the TMI-1 technical specifications (TSs). Specifically, the proposed amendment would modify the TS Table 3.1.6.1, ``Pressure Isolation Check Valves between the Primary Coolant System & LPIS Low Pressure Injection System,'' maximum allowable leakage limits.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, along with NRC edits in square brackets:
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Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes will not alter the way any structure, system, or component (SSC) functions, and will not alter the manner in which the plant is operated. In addition, the proposed amendment will not impact the ability of any SSC to mitigate an accident as currently evaluated in the UFSAR Updated Final Safety Analysis Report.
This proposed change deletes certain Reactor Coolant System Pressure Isolation Valve (RCS PIV) allowable leakage surveillance testing criteria in consideration of the safety significance and design capabilities of the plant and current industry testing and maintenance practices. The proposed change is consistent with Improved Standard Technical Specification (ITS) NUREG 1430, ``Standard Technical Specifications, Babcock and Wilcox Plants,'' Revision 4, and current RCS PIV leak testing practices. The maximum allowable leakage rate of 5 gpm gallons per minute remains unchanged; only the leakage testing incremental testing acceptance criteria below the 5 gpm limit is being deleted. Since the testing frequency and maximum allowable leakage remains unchanged, the probability or consequence of an interfacing system loss-of-coolant accident (ISLOCA) is unaffected. There are no changes to the American Society of Mechanical Engineers ASME Operation and Maintenance OM Code leakage testing requirements and methods for this class of valves. Additionally, two typographical errors and one clerical error are being corrected which are administrative in nature.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed revision is not a result of changes to plant equipment, system design, or operating practices. The modified limiting condition of operation LCO requirement will allow some relaxation of the leak testing method acceptance criteria for the RCS PIVs, consistent with NUREG-1430. Since the functions of the associated systems will continue to perform without change, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Further, the proposed changes do not introduce any new failure modes. Additionally, two typographical errors and one clerical error are being corrected which are administrative in nature.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed revision to the RCS PIV leakage testing acceptance criteria will not result in changes to system design or setpoints that are intended to ensure timely identification of plant conditions that could be precursors to accidents or potential degradation of accident mitigation systems. Since testing frequency and maximum allowable leakage for the RCS PIVs remain unchanged, the margin associated with the identification of RCS PIV degradation is not significantly reduced. The confidence in the ability of the fission product barriers (fuel cladding, RCS boundary, containment) to limit the level of radiation dose to the public remains the same. Additionally, two typographical errors and one clerical error are being corrected which are administrative in nature.
Since the setpoints and design features that support the margin of safety are unchanged, and actions for inoperable systems continue to provide appropriate time limits and compensatory measures, the proposed changes will not significantly reduce the margin of safety.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are
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satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Esquire, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Meena Khanna.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California
Date of amendment request: November 25, 2013. A publicly-available version is in ADAMS under Accession No. ML13330A557.
Description of amendment request: The proposed amendments would revise the Technical Specifications (TSs) to permit the use of Risk-
Informed Completion Times (CTs) in accordance with Technical Specification Task Force (TSTF) traveler, TSTF-505, Revision 1, ``Provide Risk-Informed Extended Completion Times--RITSTF Risked-
Informed TSTF Initiative 4b.'' The proposed amendment would, in part, modify selected Required Actions to permit extending the CTs in accordance with a new TS-required risk-informed completion time (RICT) program. The availability of the model safety evaluation for TSTF-505 was published by the NRC staff in the Federal Register on March 15, 2012 (77 FR 15399,) for referencing in license amendment applications.
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:
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Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change permits the extension of CTs provided the associated risk is assessed and managed in accordance with the NRC-
approved Risk Informed Completion Time (RICT) Program. The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the change involves no change to the plant or its modes of operation. The proposed change does not increase the consequences of an accident previously evaluated because the design basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended CT are no different from those during the existing CT.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed change create the possibility of a different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not change the design, configuration, or method of operation of the plant. The proposed change does not involve a physical alteration of the plant (no new or different kind of equipment will be installed).
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change permits the extension of CTs provided risk is assessed and managed in accordance with the NRC-approved RICT Program. The proposed change implements a risk-informed configuration management program to assure that adequate margins of safety are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendment involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Michael T. Markley.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: September 11, 2014. A publicly-available version is in ADAMS under Accession No. ML14254A371.
Description of amendment request: The proposed changes would revise the Combined Licenses by clarifying the position on design diversity, specifically human diversity, as related to the Component Interface Module (CIM) and Diverse Actuation System (DAS) design.
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:
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Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The requested amendment proposes changes to licensing basis documents to clarify the position on the human diversity aspects of design diversity as related to the Component Interface Module (CIM) and Diverse Actuation System (DAS) design processes. A review confirmed that the clarified position on human diversity would not change the CIM or DAS design. The requested changes to information presented in the Tier 2* and Tier 2 supporting documentation clarify the level of human diversity applied. The change continues to comply with the regulatory guidance in NUREG/CR-6303 regarding credible defenses against a postulated Common Cause Failure (CCF) of the Plant Monitoring and Safety System. The proposed change does not affect the plant itself. The change does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, or component (SSC) or function is adversely affected. The change does not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected. This activity will not allow for a new fission product release path, nor will it result in a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. Because the proposed changes do not change any safety-
related SSC or function credited in the mitigation of an accident, the consequences of the accidents evaluated in the UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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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 clarify the position on human diversity and show that the CIM/DAS diversity meets the regulatory guidance in NUREG/CR-6303. The clarified descriptions do not affect the plant itself. Therefore, the proposed changes do not affect any safety-
related equipment itself, nor do they affect equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. No system or design function or equipment qualification would be adversely affected by the proposed changes. Furthermore, the proposed changes do not result in a new failure mode, malfunction, or sequence of events that could affect safety or safety-related equipment.
Therefore, the proposed amendment does not create the possibility of a new or different
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kind of accident from any accident previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed changes to information presented in referenced licensing basis documents clarify the position regarding human diversity and do not affect the plant itself. The proposed changes do not adversely affect the design, construction, or operation of any plant SSCs, including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: October 23, 2014. A publicly-available version is in ADAMS under Accession No. ML14296A758.
Description of amendment request: The proposed changes would revise the Combined Licenses (COLs) changing the description and scope of the Initial Test Program. 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:
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Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment is related to the conduct of the Initial Test Program. The proposed changes are made in compliance with the applicable regulatory guides, are only related to the general aspects of how the program is executed and do not change any technical content for preoperational or startup tests. No changes are made to any design aspect of the plant.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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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 is related to the conduct of the Initial Test Program. The proposed changes are made in compliance with the applicable regulatory guides, are only related to the general aspects of how the program is executed and do not change any technical content for preoperational or startup tests. These changes do not affect the design or analyzed operation of any system.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment is related to the conduct of the Initial Test Program. The proposed changes are made in compliance with the applicable regulatory guides, are only related to the general aspects of how the program is executed and do not change any technical content for preoperational or startup tests. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus no margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
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 16, 2014. A publicly-available version is in ADAMS under Accession No. ML14290A139.
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 changes to revise the VEGP Updated Final Safety Analysis Report (UFSAR), involving Tier 1 and associated Tier 2 departures to add or delete piping line numbers of existing piping lines, or updating the functional capability classification of existing process flow lines.
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:
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Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes involve updating piping line name/number or functional capability requirements. These changes do not affect any system design function. Adding or updating information for existing ASME Section III piping does not involve (i.e., cannot affect) any accident initiating event or component failure, thus, the probabilities of the accidents previously evaluated are not affected. The maximum allowable leakage rate specified in the Technical Specifications is unchanged and radiological material release source terms are not affected, thus, the radiological releases in the accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes to update piping line name/number or functional capability requirements do not adversely affect the design or quality of any structure, system, or component. Adding or updating ASME Section III piping line information for existing process piping lines to a licensing table does not create a new fault or sequence of events that could result in a radioactive material release.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
Page 73113
The COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes involve updating piping line name/number or functional capability requirements information for new/existing process piping lines. Adding or updating the ASME Section III piping line name/number or functional capability requirements in the tables would not affect any radioactive material barrier. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus, no margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
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Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1, Washington County, Nebraska
Date of amendment request: November 7, 2014. A publicly-available version is in ADAMS under Accession No. ML14311A158.
Brief description of amendment request: The proposed amendment revises a limited number of Technical Specification Surveillance Requirements by adding a note or footnote permitting a one-time extension from a refueling frequency (i.e., at least once per 18 months) to a maximum of 28 months. These surveillance requirements include (1) manual containment isolation actuation, (2) manual recirculation actuation and recirculation actuation logic, (3) steam generator level calibration, (4) visual examination of the high-
efficiency particulate air and charcoal filters in the containment recirculating air cooling and filtering system, (5) emergency diesel generators, and (6) residual heat removal system integrity. An extension is necessary because these tests will expire before the next refueling outage begins on April 11, 2015.
Date of publication of individual notice in Federal Register: November 17, 2014 (79 FR 68487).
Expiration date of individual notice: December 17, 2014 (public comments); January 17, 2015 (hearing requests).
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Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas
Date of amendment request: January 28, 2013, as supplemented by letter dated August 28, 2013.
Brief description of amendment: The amendment revised Technical Specification (TS) requirements related to direct current (DC) electrical systems as specified in TS Limiting Condition for Operation (LCO) 3.8.4, ``DC Sources--Operating,'' LCO 3.8.5, ``DC Sources--
Shutdown,'' and LCO 3.8.6, ``Battery Parameters.'' A new ``Battery Monitoring and Maintenance Program'' is now required under TS Section 5.5, ``Administrative Controls--Programs and Manuals.'' These changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC Electrical Rewrite--
Update to TSTF-360.'' The availability of this TS improvement was announced in the Federal Register on September 1, 2011 (76 FR 54510).
Date of issuance: November 24, 2014.
Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.
Amendment No.: 250. A publicly-available version is in ADAMS under Accession No. ML14254A133; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications/license.
Date of initial notice in Federal Register: April 30, 2013 (78 FR 25313). The supplemental letter dated August 28, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 24, 2014.
No significant hazards consideration comments received: No.
Page 73114
Exelon Generation Company, LLC, Docket No. 50-373, LaSalle County Station (LSCS), Unit 1, LaSalle County, Illinois
Date of amendment request: December 20, 2013, as supplemented by letters dated February 26, 2014, September 11, 2014 (2 letters), and October 14, 2014.
Brief description of amendment: The amendment revised the LSCS, Unit 1, pressure and temperature curves, Figures 3.4.11-1 through 3.4.11-3, in Technical Specification 3.4.11, ``RCS Reactor Coolant System Pressure and Temperature (P/T) Limits.''
Date of issuance: November 25, 2014.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 210. A publicly-available version is in ADAMS under Accession No. ML14288A151; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-11: Amendment revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 5, 2014 (79 FR 45490). The supplemental letters dated September 11, 2014 (2 letters) and October 14, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 25, 2014.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: November 6, 2013, supplemented by letters dated June 13, 2014, and August 15, 2014.
Brief description of amendments: The amendments revised the Technical Specification 3.6.13, Divider Barrier Integrity, Surveillance Requirement 3.6.13.5 for the divider barrier seal inspection for the Donald C. Cook Nuclear Plant, Units 1 and 2.
Date of issuance: November 20, 2014.
Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment Nos.: 324 for Unit 1 and 307 for Unit 2.
Renewed Facility Operating License Nos. DPR-58 and DPR-74: The amendments revise the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 19, 2014 (79 FR 9496). The supplemental letters dated June 13, 2014, and August 15, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 20, 2014.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: March 11, 2013, as supplemented by letter dated July 3, 2014.
Brief description of amendment: The amendment changes the reactor steam dome pressure value specified in technical specification (TS) 2.1.1, ``Reactor Core SLs Safety Limits,'' from 785 pounds per square inch gauge (psig) to 686 psig. This change resolves a 10 CFR part 21, ``Reporting of Defects and Noncompliance,'' condition concerning a potential to momentarily violate the safety limit specified in TS 2.1.1.1 during a pressure regulator failure maximum demand (open) transient.
Date of issuance: November 25, 2014.
Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment No.: 185. A publicly-available version is in ADAMS under Accession No. ML14281A318; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22: This amendment revises the Renewed Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: June 11, 2013 (78 FR 35064). The supplemental letter dated July 3, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 25, 2014.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota
Date of amendment request: May 23, 2013, as supplemented by letter dated March 25, June 26, and October 20, 2014.
Brief description of amendment: The amendments revised Technical Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to reference and allow use of Westinghouse report WCAP-16045-P-A, ``Qualification of the Two-Dimensional Transport Code PARAGON'' and WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear Data Methodology,'' to determine core operating limits.
Date of issuance: November 19, 2014.
Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment Nos.: Unit 1-211; Unit 2-199. A publicly-available version is in ADAMS under Accession No. ML14296A666; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-42, and DPR-60: These amendments revised the Renewed Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: August 20, 2013 (78 FR 51229). The supplements dated March 25, June 26, and October 20, 2014, 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 November 19, 2014.
No significant hazards consideration comments received: No.
Page 73115
Southern Nuclear Operating Company Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia
Date of amendment request: July 2, 2013, and revised by letters dated February 14, and June 20, 2014, and supplemented by letters dated August 28 and October 14, 2014.
Brief description of amendment: The amendment revises the design of connections between reinforced concrete and steel plate concrete composite construction included in the VEGP, Units 3 and 4 updated Final Safety Analysis Report (UFSAR) and changes to the Technical Report, ``APP-GW-GLR-602, AP1000 Shield Building Design Details for Select Wall and RC/SC Connections,'' (prepared by Westinghouse Electric Company and reviewed by the NRC as part of the design certification rule). This Technical Report is incorporated by reference in the VEGP, Units 3 and 4 UFSAR.
Date of issuance: November 21, 2014.
Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.
Amendment No.: 26. A publicly-available version is in ADAMS under Accession No. ML14322A275; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.
Date of initial notice in Federal Register: September 3, 2013 (78 FR 54287).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 21, 2014.
No significant hazards consideration comments received: No.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.
Date of initial notice in Federal Register: September 3, 2013 (78 FR 54287).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 21, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of December 2014.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
FR Doc. 2014-28704 Filed 12-8-14; 8:45 am
BILLING CODE 7590-01-P
