NRC-2014-0917, Biweekly Sholly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: September 16, 2014

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Biweekly Sholly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: September 16, 2014
ML14247A154
Person / Time
Site: Hatch, Dresden, Peach Bottom, Oconee, Mcguire, Perry, Oyster Creek, Byron, Three Mile Island, Braidwood, Limerick, North Anna, Vermont Yankee, Ginna, Vogtle, Robinson, Clinton, San Onofre, Quad Cities, McGuire, LaSalle, 07000053  Constellation icon.png
Issue date: 09/05/2014
From: Lund A L
Division of Operating Reactor Licensing
To:
Clayton B
References
NRC-2014-0917
Download: ML14247A154 (37)
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Text

[7590-01-P] NUCLEAR REGULATORY COMMISSION

[NRC-2014-0917] 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 August 21, 2014 to September 3, 2014. The last biweekly notice was published on September 2, 2014.

2 DATES: Comments must be filed by [INSERT DATE 30 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER

]. A request for a hearing must be filed by [INSERT DATE 60 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER

]. 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):

Carol.Gallagher@nrc.gov.

  • 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: Beverly A. Clayton, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:

301-415-3475, e-mail: Beverly.Clayton@nrc.gov

.

3SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments.

A. Obtaining Information. Please refer to Docket ID NRC-2014-0917 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:

  • 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 e-mail 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.

4B. Submitting Comments. Please include Docket ID NRC-2014-0917 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. 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 considerat ion. Under the Commission's regulations in

§ 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 5increase 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 t he 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 shutdow n 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 6with 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 7contention 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.

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 8governmental 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 e-mail 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 det ailed 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 9submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a W eb 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 e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail 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 represent ative) 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 e-mail to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Meta System 10Help 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 proc eedings 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 11 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 fo r public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the "Obtaining Information and Submitting Comments" section of this document.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: June 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14184B384. Description of amendment request: The amendment would revise the Technical Specifications (TS) by reducing the allowed maximum rated thermal power (RTP) at which the unit can operate when select High Pressur e Injection (HPI) System equipment is inoperable. 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?

12 Response: No.

The proposed TS changes do not modify the reactor coolant system pressure boundary, nor make any physical changes to the facility design, material, or construction standards. The probability of any design basis accident (DBA) is not affected by this change, nor are the consequences of any DBA affected by this change. The new small break loss-of-coolant accident (SBLOCA) partial-power analysis demonstrates that all 10 CFR 50.46 acceptance criteria are satisfied. Radiological consequences for loss-of-coolant accident (LOCA) events are evaluated in ONS Updated Final Safety Analysis Report Section 15.15 for the Maximum Hypothetical Accident. The proposed changes will not impact assumptions and conditions previously used in the radiological consequence evaluations for the Maximum Hypothetical Accident. The proposed changes do not involve changes to any structures, systems, or components (SSCs) that can alter the probability for initiating a LOCA event.

Therefore, the proposed TS changes do 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 TS changes reduce the allowed power level that the unit may be operated at with select HPI equipment out-of-service. The changes do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. No new failure modes are identified, nor are any SSCs required to be operated outside the design bases.

Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed TS changes are supported by SBLOCA analyses which demonstrate that the acceptance criteria of 10 CFR 50.46 are satisfied.

These analyses were performed in accordance with the Evaluation Model described in AREVA Topical Report BAW-10192P-A. The new SBLOCA analysis assumes a lower initial core power level (50% of rated thermal power (RTP)) than what was previously analyzed in support of TS 3.5.2 (i.e., 75% of RTP). The resulting peak cladding temperature results for 13the new SBLOCA analysis are lower than the existing analysis. In addition, a supplemental evaluation demonstrated that failure to perform a desired operator action of maintaining secondary-side pressure at 300 psig by throttling the atmospheric dump valve during a SBLOCA did not result in adverse affects to the new SBLOCA analysis results. Therefore, it is concluded that the proposed amendment request will not result in a significant decrease in the margin of safety.

The NRC staff has reviewed the licensee's ana lysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street - EC07H, Charlotte, NC 28202-1802. NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Progress Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2 (HBRSEP2), Darlington County, South Carolina Date of amendment request: June 20, 2014. A publicly-available version is in ADAMS under Accession No. ML14188B015.

Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.9.b.2 for the Steam Generator (SG) Program accident-induced leakage performance criterion to correct an editorial error in the accident-induced leakage rate value for any design-basis accident other than a SG tube rupture. 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?

14Response: No. The proposed change is a correction to an editorial error in the specified accident induced leakage performance criterion of TS 5.5.9.b.2. The error in TS 5.5.9.b.2 being addressed by this proposed change was introduced at the time of the HBRSEP2 submittal of the NRC-approved Technical Specification Task Force (TSTF) traveler 449, Rev. 4, Steam Generator Tube Integrity. The accident-induced leakage performance criterion will continue to be within the limit assumed in the accident analysis. As a result, neither the probability nor the consequences of any accident previously evaluated will be affected.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new or different accidents result from the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis, it only corrects an editorial error in the accident-induced leakage performance criterion specified in the SG Program. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. This change will have no effect on the margin of safety. This proposed change corrects an editorial error in the accident-induced leakage performance criterion specified in the SG Program.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202. Acting NRC Branch Chief

Lisa M. Regner.

15 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: November 14, 2013. A publicly-available version is in ADAMS under Accession No. ML13323A516. Description of amendment request: The proposed amendment would eliminate operability requirements for secondary containment when handling sufficiently decayed irradiated fuel or a fuel cask following a minimum of 13 days after the permanent cessation of reactor operation.

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 do not modify the design or operation of equipment used to move spent fuel or to perform core alterations. The proposed changes cannot increase the probability of any previously analyzed accident because they are based on changes in Source Term, atmospheric dispersion and dose consequence analysis methodology, not in procedures or equipment used for fuel handling.

The conservative re-analysis of the FHA [fuel-handling accident] concludes that the radiological consequences are within the regulatory limits established 10 CFR 50.67. This conclusion is based on the Alternate Source Term and guidance provided in Appendix B of Regulatory Guide 1.183 and analyses of fission product release and transport path that does not take credit for dose mitigation provided by engineered safeguards including secondary containment and the SGT system. The results of the core alteration events, other than the FHA, remain unchanged from the original design-basis that showed these events do not result in fuel cladding damage or radioactive release.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

162. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Regulation in 10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design-basis radiological consequence analyses. This license amendment application evaluates the consequences of a design-basis fuel handling accident in accordance with this regulation and Regulatory Guide 1.183. The revised analysis concludes that the radiological consequences of the fuel handling accident are less than the regulatory allowable limits. Safety margins and analytical conservatisms are retained to ensure the analysis adequately bounds all postulated event scenarios. The selected assumptions and release models provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensates for large uncertainties in facility parameters, accident progression, radioactive material transport and atmospheric dispersion. The proposed TS applicability statements continue to ensure that the total effective dose equivalent (TEDE) at the boundaries of the control room, the exclusion area, and low population zone boundaries are below the corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).

Therefore, the proposed change does not involve 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

17Attorney 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.

Exelon Generation Company, LLC, Docket Nos. STN 50-456, STN 50-457 and 72-73, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454, STN 50-455 and 72-68, Byron Station, Units 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-10, 50-237, 50-249 and 72-37, Dresden Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373, 50-374 and 72-70, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-352, 50-353 and 72-65, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, et al., Docket No. 50-219 and 72-15, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-171, 50-277, 50-278 and 72-29, Peach Bottom Atomic Power Station, Units 1, 2 and 3, York and Lancaster Counties, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50-254, 50-265 and 70-53, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

18Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania Exelon Generation Company, LLC, Docket No. 50-320, Three Mile Island Nuclear Station, Unit 2, Dauphin County, Pennsylvania Date of amendment request: May 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14164A054. Description of amendment request: The proposed changes revise the Emergency Plans for the affected facilities to adopt the Nuclear Energy Institute's (NEl's) revised Emergency Action Level (EAL) schemes described in NEI 99-01, Revisi on 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC in a letter dated March 28, 2013. A publicly-available version can be found in ADAMS under Accession No.

ML12346A463. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: The proposed changes have been reviewed considering the applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and other applicable

NRC documents. Exelon has evaluated t he proposed changes to the affected sites' Emergency Plans and determined that the changes do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards, set forth in 10 CFR 50.92, "Issuance of amendment," is provided below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to Exelon's EAL schemes 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 19not reduce the functionality, performance, or capability of Exelon's 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 Plans. As discussed in Section 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, Exelon did not consider the risk insights regarding any specific accident initiation or 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to Exelon's EAL schemes 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 20will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. All Exelon 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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to Exelon's EAL schemes 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 limits, 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.

In conclusion, and based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes to adopt the EAL schemes established in NEI 99-01, Revision 6, as endorsed by the U.S.

Nuclear Regulatory Commission (NRC); (2) the changes will be in compliance with the NRC's regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

The NRC staff has reviewed the licensee's ana lysis 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.

21Attorney for licensee: Bradley Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief

Travis L. Tate.

FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Perry, OH Date of amendment request: March 25, 2014. A publicly-available version is in ADAMS under Accession No. ML14084A165.

Description of amendment request: The proposed changes are consistent with the NRC-approved Industry/Technical Specifications Task Force (TSTF) Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b." The proposed change relocates surveillance frequencies to a licensee controlled program, the Surveillance Frequency Control Program. This change is applicable to licensees using probabilistic risk guidelines contained in NRC-approved Nuclear Ener gy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies."

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 change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems 22and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (that is, no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident

previously evaluated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to the TS [technical specification]),

since these are not affected by changes to the surveillance frequencies.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, FENOC will perform a probabilistic risk evaluation using the guidance contained in NRC approved Nuclear Energy Institute (NEI) 04-10, Revision 1, in accordance with the TS Surveillance Frequency Control Program. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance 23frequencies consistent with Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications."

Therefore, the proposed change does not involve a significant reduction

in a margin of safety.

Based upon the reasoning presented above, FENOC concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.

The NRC staff has reviewed the licensee's ana lysis 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.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San Diego County, California Date of amendment request: March 21, 2014. A publicly-available version is in ADAMS under Accession No. ML14085A141.

Description of amendment request: The proposed amendment would revise the Operating License and associated Technical Specifications (TS) to reflect the permanent cessation of power operation. Because the licenses for SONGS, Units 2 and 3 no longer authorize emplacement or retention of fuel in the reactor vessel, the limiting conditions for operation and associated surveillance requirements that do not apply in the defueled condition are being proposed for deletion. The remaining portions of the TS are being proposed for revision and incorporation as the permanently defueled TS to provide a continuing acceptable level of safety, 24which addresses the reduced scope of postulated design basis accidents associated with a defueled plant, as described in the SONGS, Units 2 and 3 safety analyses.

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.

SONGS Units 2 and 3 have permanently ceased operation. The proposed amendment would modify t he SONGS Units 2 and 3 facility operating licenses and TS by deleting the portions of the licenses and TS that are no longer applicable to a permanently defueled facility, while modifying the remaining portions to correspond to the permanently shutdown condition. This change is consistent with the criteria set forth in 10 CFR 50.36 for the contents of TS.

Section 15 of the SONGS Updated Final Safety Analysis Report (UFSAR) described the design basis accident (DBA) and transient scenarios applicable to SONGS Units 2 and 3 during power operations. With the reactors in a permanently defueled condition, the fuel storage pools and their systems have been isolated and are dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. As a result of the certifications submitted by SCE [Southern California Edison] in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactors or to place or retain fuel in the reactors in accordance with 10 CFR 50.82(a)(2), most of the accident scenarios postulated in the UFSAR are no longer possible.

The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure: 1. The integrity of the reactor coolant boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition; or 3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in

10 CFR 50.43(a)(1) or 100.11.

25 The first two criteria (integrity of the reactor coolant pressure boundary and safe shut down of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after the termination of reactor operations at SONGS Units 2 and 3 and the permanent removal of the fuel from the reactor vessels (following 17 months of decay time after shut down) and purging of the contents of the waste gas decay tanks, none of the SSCs at SONGS Units 2 and 3 are required to be relied on for accident mitigation. Therefore, none of the SSCs at SONGS Units 2 and 3 meet the definition of a safety-related SSC stated in 10 CFR 50.2 (with the exception of the passive fuel storage pool structure).

The deletion of TS definitions and rules of usage and application, that are currently not applicable in a defueled condition, has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled status of SONGS Units 2 and 3 has no impact on the remaining DBA. The removal of limiting conditions for operation (LCOs) or surveillance requirements (SRs) that are related only to the operation of the nuclear reactors or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the defueled mode. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system inventory control, and containment integrity are no longer applicable at SONGS Units 2 and 3 as a permanently defueled plant. The analyzed accidents involving damage to the reactor coolant system, main steam lines, reactor core, and the subsequent release of radioactive material are no longer possible at SONGS Units 2 and 3.

Since SONGS Units 2 and 3 has permanently ceased operation, the future generation of fission products has ceased and the remaining source term will decay. The radioactive decay of the irradiated fuel since shut down of the reactor will have reduced the consequences of the FHA

[fuel handling accident] to levels well below those previously analyzed.

The relevant parameter (water level) associated with the fuel pool provides an initial condition for the FHA analysis and is included in the permanently defueled TS.

The fuel storage pool water level, fuel storage pool boron concentration, and spent fuel assembly storage TS are retained to preserve the current requirements for safe storage of irradiated fuel.

Fuel pool cooling and makeup related equipment and support equipment (e.g., electrical power systems) are not required to be continuously 26available since there is sufficient time to effect repairs, establish alternate sources of makeup flow, or establish alternate sources of cooling in the event of a loss of cooling and makeup flow to the fuel storage pool.

The deletion and modification of provisions of the administrative controls does not directly affect the design of SSCs necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shut down and defueled condition of the reactors.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition is the only operation currently allowed, and therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

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 have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents cannot result in different or more adverse failure MODES or accidents than previously evaluated because the reactor is permanently shut down and defueled and SCE is no longer authorized to operate the

reactors.

The proposed deletion of requirements of the SONGS Unit 2 and Unit 3 TS do not affect systems credited in the accident analysis. The proposed permanently defueled TS (PDTS) continue to require proper control and monitoring of safety significant parameters and activities.

The proposed restriction on the fuel pool level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA. The fuel storage pool water level, fuel storage pool boron concentration, and spent fuel assembly storage TS 27are retained to preserve the current requirements for safe storage of irradiated fuel.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (i.e., fuel cladding and spent fuel cooling). Since extended operation in a defueled condition is the only operation currently allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of 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 previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Because the 10 CFR Part 50 licenses for SONGS Units 2 and 3 no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The remaining credible accidents do not credit SSCs for mitigation. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact an accident.

The proposed changes are limited to those portions of TS and license that are not related to the safe storage of irradiated fuel. The requirements for SSCs that have been deleted from the SONGS TS Units 2 and 3 are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs involving the reactors are no longer possible because the reactors are permanently shut down and defueled and SCE is no longer authorized to operate the reactors.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety because the current design limits continue to be met for the accidents of concern.

The NRC staff has reviewed the licensee's ana lysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 28proposes to determine that the amendment requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief

Douglas A. Broaddus.

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: June 4, 2014. A publicly-available version is in ADAMS under Accession No. ML14156A477. Description of amendment request: The purpose of the proposed license amendment request is to address proposed changes related to departure from the plant-specific Design Control Document (DCD) Tier 1 (and corresponding Combined License Appendix C information) and Tier 2 material to reconcile differences in the various valve table entries. Because this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the lic ensee 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 requested amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes do not result in any physical changes to the plant, and therefore do not change any safety-related design requirement, qualification requirement or function. The proposed changes do not 29involve any accident initiating event or component failure, thus, the probabilities of the accidents previously evaluated are not affected. The proposed changes do not affect the radioactive material releases used in the accident analyses, thus, the radiological releases in the accident analyses are not affected. The proposed changes do not affect any postulated non-radioactive accident scenario as evaluated in UFSAR

[Updated Final Safety Analysis Report] Chapter 15.

Therefore, the requested amendment does not involve a significant increase in the probability or consequences of an accident previously

evaluated.

2. Does the requested amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not result in any physical changes to the plant, and therefore do not adversely affect any structure, system or component. No safety-related equipment qualification or design function is affected. The proposed changes do not introduce a new failure mode or create a new fault or sequence of events that could result in a radioactive material release.

Therefore, the requested amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not result in any physical changes to the plant, and therefore do not change valve performance, including containment isolation. No safety acceptance criterion would be exceeded or challenged. No safety related function would be affected. Valve qualification would not be affected.

The proposed changes do not affect compliance with existing design codes and regulatory criteria and do not affect any safety analysis.

Therefore, the requested amendment does not involve a significant reduction in a margin of safety.

30The NRC staff has reviewed the licensee's ana lysis 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, Blach & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief

Lawrence Burkhart.

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, 31pursuant 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.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of application for amendments: September 12, 2013, as supplemented by letters dated May 20 and July 22, 2014. Brief description of amendments: The amendments modify Technical Specification (TS) 3.3.2. Specifically, the change modifies setpoints associated with the auxiliary feedwater pump suction transfer on low suction pressure. Date of issuance: August 27, 2014.

Effective date: This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 273 and 253. A publicly-available version is in ADAMS under Accession No. ML14211A403; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

32Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and technical specifications. Date of initial notice in Federal Register: December 10, 2013 (78 FR 74179). The supplemental letters dated May 20 and July 22, 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 August 27, 2014. No significant hazards consideration comments received

No Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of application for amendments: September 28, 2012, as supplemented by letters dated February 15, 2013, May 7, 2013, May 24, 2013, June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August 5, 2013, August 22, 2013, August 29, 2013, September 13, 2013, October 11, 2013, October 15, 2013, October 31, 2013, December 6, 2013, December 20, 2013, January 17, 2014, January 31, 2014 (2 letters), February 20, 2014, February 28, 2014, March 10, 2014, March 17, 2014, April 11, 2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20, 2014.

33Brief description of amendments: The amendments authorize an increase in the maximum licensed thermal power level for PBAPS, Units 2 and 3, from 3514 megawatts thermal (MWt) to 3951 MWt, which is an increase of approximately 12.4 percent.

Date of issuance: August 25, 2014.

Effective date: For PBAPS, Unit 2, the amendment is effective as of its date of issuance and shall be implemented prior to startup from refueling outage P2R20. For PBAPS, Unit 3, the amendment is effective as of its date of issuance and shall be implemented prior to startup from refueling outage P3R20.

Amendments Nos.: 293 and 296. A publicly-available version is in ADAMS under Accession No. ML14133A046; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Facility Operating Licenses and the Technical Specifications. Date of initial notice in Federal Register: April 9, 2013 (78 FR 21168). The letters dated February 15, 2013, May 7, 2013, May 24, 2013, June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August 5, 2013, August 22, 2013, August 29, 2013, September 13, 2013, October 11, 2013, October 15, 2013, October 31, 2013, December 6, 2013, December 20, 2013, January 17, 2014, January 31, 2014 (2 letters), February 20, 2014, February 28, 2014, March 10, 2014, March 17, 2014, April 11, 2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20, 2014, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

34The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 25, 2014. No significant hazards consideration comments received

No. Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York Date of amendment request: February 28, 2013, as supplemented by letters dated June 19, and November 11, 2013 and January 22, March 14, March 26, and June 6, 2014. Brief description of amendment: The amendment revised the Ginna Nuclear Power Plant Technical Specifications (TSs) to revise the allowable containment average air temperature

from " 120 °F" to " 125 °F" for TS 3.6.5 "Containment Air Temperature." Date of issuance: August 12, 2014.

Effective date: As of the date of issuance to be implemented within 45 days. Amendment No.: 116. A publicly-available version is in ADAMS under Accession No. ML14232A125; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-18

Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register

November 26, 2013 (78 FR 70594). The supplemental letters dated June 19, and November 11, 2013, and January 22, March 14, March 26, and June 6, 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 35original 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 August 12, 2014. No significant hazards consideration comments received
No. Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Date of application for amendments: July 23, 2013, as supplemented August 5, 2014. Brief description of amendments: The amendments revise the Technical Specification (TS) requirements and add license conditions related to control room envelope habitability in accordance with the Nuclear Regulatory Commission approved Revision 3 of Technical Specification Task Force (TSTF) Standard Techni cal Specifications Change Traveler TSTF-448, "Control Room Habitability."

Date of issuance: August 29, 2014.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: Unit 1 - 268 and Unit 2 - 212. A publicly-available version is in ADAMS under Accession No. ML14147A410; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Renewed Facility Operating licenses and the Technical Specifications.

36Date of initial notice in Federal Register: September 3, 2013 (78 FR 54290). The supplement dated August 5, 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. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 29, 2014. No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County Date of application for amendment: February 22, 2013. Brief description of amendment: The amendment revised Technical Specification 3.1.6, "Control Bank Insertion Limits," to include text, in Condition A, stating, "for reasons other than Condition C." This text addition modifies Condition A, for control bank sequence or overlap limits, to include language currently in Condition B, for control bank insertion limits, this change would point to Condition C, which, if applicable, would allow the specified completion time to restore the control bank to within the insertion limit to be increased from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This would align the description of the sequence and overlap limit of Condition A with the description of control bank insertion limit Condition B. Date of issuance: August 27, 2014.

37 Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 272 and 254. A publicly-available version is in ADAMS under Accession No. ML14188C453; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register

April 30, 2013 (78 FR 25317). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 27, 2014. No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 5 th day of September 2014.

For the Nuclear Regulatory Commission.

/RA/

A. Louise Lund, Acting Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

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