Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

As Enacted ( Federal Register)

Cited authorities 6

Cited in

Federal Register, Volume 77 Issue 200 (Tuesday, October 16, 2012)

Federal Register Volume 77, Number 200 (Tuesday, October 16, 2012)

Notices

Pages 63343-63355

From the Federal Register Online via the Government Printing Office www.gpo.gov

FR Doc No: 2012-25240

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NUCLEAR REGULATORY COMMISSION

NRC-2012-0236

Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

Background

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission 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 September 20, 2012 to October 3, 2012. The last biweekly notice was published on October 2, 2012 (77 FR 60146-

60160).

Addresses: You may access information and comment submissions related to this document, which the NRC possesses and are publicly available, by searching on http://www.regulations.gov under Docket ID NRC-2012-0236. You may submit comments by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-

3668; email: Carol.Gallagher@nrc.gov.

Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Fax comments to: RADB at 301-492-3446.

For additional direction on accessing information and submitting comments, see ``Accessing Information and Submitting Comments'' in the Supplementary Information section of this document.

Supplementary Information:

Accessing Information and Submitting Comments

Accessing Information

Please refer to Docket ID NRC-2012-0236 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and are publicly available, by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236.

NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. To begin the search,

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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. Documents may be viewed in ADAMS by performing a search on the document date and docket number.

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.
Submitting Comments

Please include Docket ID NRC-2012-0236 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 will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Notice 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 Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in section 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.

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 ``Rules of Practice for Domestic Licensing Proceedings'' 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 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

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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.

All documents filed in the NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at 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/apply-certificates.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 the 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 agency'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 Web site at http://www.nrc.gov/site-help/e-submittals.html, by email at 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. 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

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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 following three factors in 10 CFR 2.309(c)(1): (i) The information upon which the filing is based was not previously available; (ii) the information upon which the filing is based is materially different from information previously available; and (iii) the filing has been submitted in a timely fashion based on the availability of the subsequent information.

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov.

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina.

Date of amendment request: June 19, 2012.

Description of amendment request: The proposed license amendments would revise the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2. The TS change proposes to extend the Completion Time (CT) of TS 3.8.1 Required Action D.4 for an inoperable diesel generator (DG). A commensurate change is also proposed to extend the maximum CT of TS 3.8.1 Required Actions C.3 and D.4. The licensee stated that it will add a supplemental alternating current power source (i.e., a supplemental diesel generator) with the capability to power any E-bus within one hour from the Station Blackout (SBO) event, and with the capacity to bring the affected unit to cold shutdown, to support this request.

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 DGs are safety related components which provide backup electrical power supply to the onsite emergency power distribution system. The proposed changes do not affect the design of the DGs, the operational characteristics or function of the DGs, the interfaces between the DGs and other plant systems, or the reliability of the DGs. The DGs are not accident initiators; the DGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power. Extending the CT for a single DG would not affect the previously evaluated accidents since the remaining DGs supporting the redundant ESF engineered safety feature systems would continue to be available to perform the accident mitigation functions. Thus, allowing a DG to be inoperable for an additional 7 days for performance of maintenance or testing does not increase the probability of a previously evaluated accident.
  
  Deterministic and probabilistic risk assessments evaluated the effect of the proposed TS changes on the availability of an electrical power supply to the plant emergency safeguards features systems. These assessments concluded that the proposed TS changes do not involve a significant increase in the risk of power supply unavailability.
  
  There is small incremental risk associated with continued operation for an additional 7 days with one DG inoperable; however, the calculated impact on risk provides risk metrics consistent with the acceptance guidelines contained in RG Regulatory Guide 1.177 and 1.174 (References 7.2.1 and 7.2.2). This risk is judged to be reasonably consistent with the risk associated with operations for 7 days with one DG inoperable as allowed by the current TS.
  
  Specifically, the remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. A DG is required to operate only if both offsite power sources fail and there is an event which requires operation of the plant engineered safety features such as a design basis accident. The probability of a design basis accident occurring during this period is low.
  
  The consequences of previously evaluated accidents will remain the same during the proposed 14-day CT as during the current 7-day CT. The ability of the remaining TS required DG to mitigate the consequences of an accident will not be affected since no additional failures are postulated while equipment is inoperable within the TS CT. The standby AC alternating current power supply for each of the four safety-related load groups consists of one DG complete with its auxiliaries, which include the cooling water, starting air, lubrication, intake and exhaust, and fuel oil systems. The sizing of the DGs and the loads assigned among them is such that any combination of three out of four of these DGs is capable of shutting down the plant safely, maintaining the plant in a safe shutdown condition, and mitigating the consequences of accident conditions.
  
  Thus this change does not involve a significant increase in the probability or consequences of a previously analyzed accident.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed changes do not involve a change in the plant design, plant configuration, system operation, or procedures involved with the DGs. The proposed changes allow a DG to be inoperable for additional time. Equipment will be operated in the same configuration and manner that is currently allowed and designed for. The functional demands on credited equipment is unchanged. There are no new failure modes or mechanisms created due to plant operation for an extended period to perform DG maintenance or testing. Extended operation with an inoperable DG does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the extended CT.
  
  Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
  
  Response: No.
  
  Currently, if an inoperable DG is not restored to operable status within 7 days, TS 3.8.1, Condition H, requires the unit to be in MODE 3 (i.e., HOT SHUTDOWN) within a CT of 12 hours, and to be in MODE 4 (i.e., COLD SHUTDOWN) within a CT of 36 hours. This TS Condition is entered on both units resulting in a dual-unit shutdown. The proposed Technical Specification changes will allow steady state plant operation at 100 percent power for an additional 7 days for performance of DG planned reliability improvements and preventive and corrective maintenance.
  
  Deterministic and probabilistic risk assessments evaluated the effect of the proposed TS changes on the availability of an electrical power supply to the plant ESF systems. These assessments concluded that the proposed TS changes do not involve a significant increase in the risk of power supply unavailability.
  
  The DGs continue to meet their design requirements; there is no reduction in capability or change in design configuration. The DG response to LOOP loss of offsite power, LOCA loss-of-coolant accident, SBO station blackout, or fire is not changed by this proposed amendment; there is no change to the DG operating parameters. In the extended CT, as in the existing CT, the remaining operable DGs and paths are adequate to supply electrical power to the onsite emergency power distribution system. The proposed change does not alter a design basis or safety limit; therefore, it does not significantly reduce the margin of safety. The DGs will continue to operate per the existing design and regulatory requirements.
  
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  The proposed TS changes do not alter the plant design nor does it change the assumptions contained in the safety analyses. The standby AC power system is designed with sufficient redundancy such that a DG may be removed from service for maintenance or testing. The remaining DGs are capable of carrying sufficient electrical loads to satisfy the UFSAR updated final safety analysis report requirements for accident mitigation or unit safe shutdown. The proposed changes do not impact the redundancy or availability requirements of offsite power circuits or change the ability of the plant to cope with a SBO.
  
  Therefore, based on the considerations given above, the proposed changes do not involve a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: David T. Conley, Senior Counsel--Manager Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602.
  
  NRC Acting Branch Chief: Jessie Quichocho.
  
  Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina; Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida
  
  Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina; Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, Darlington County, South Carolina
  
  Date of amendment request: September 12, 2012.
  
  Description of amendment request: The proposed license amendments would revise the Facility Operating Licenses for the Brunswick Steam Electric Plant, Units 1 and 2, H. B. Robinson Steam Electric Plant, Unit 2, Shearon Harris Nuclear Power Plant, Unit 1, and Crystal River Unit No. 3 Nuclear Generating Plant. The NRC issued license amendments, dated July 29, 2011, that approved the licensees' cyber security plan and associated implementation milestone schedule. Milestone 6 requires the identification, documentation, and implementation of cyber security controls for critical digital assets that could adversely impact the design function of physical security target set equipment by no later than December 31, 2012. The license amendment request would change the existing facility operating licenses for the Physical Protection/
  
  Security license condition for these plants to reference the change to an implementation schedule milestone and a proposed Revised Cyber Security Plan Implementation Schedule for the scope of Milestone 6.
  
  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:
4. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The proposed change to the Cyber Security Plan Implementation Schedule is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and has no impact on the probability or consequences of an accident previously evaluated.
  
  Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
5. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed change to the Cyber Security Plan Implementation Schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and 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.
6. Does the proposed change involve a significant reduction in a margin of safety?
  
  Response: No.
  
  Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the Cyber Security Plan Implementation Schedule is administrative in nature. Because there is no change to these established safety margins as result of this change, the proposed change does not involve a significant reduction in a margin of safety.
  
  Therefore, the proposed change does not involve a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.
  
  NRC Acting Branch Chief: Jessie Quichocho.
  
  Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina
  
  Date of amendment request: August 6, 2012.
  
  Description of amendment request: The proposed change would revise the Technical Specification (TS) requirements for inoperable snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The change is consistent with NRC approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
  
  372, ``Addition of LCO 3.0.8, Inoperability of Snubbers.''
  
  The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed the applicability of the model NSHC determination in its application dated August 6, 2012.
  
  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:
7. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable
  
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  snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.
  
  Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
8. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?
  
  Response: No.
  
  The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.
  
  Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
9. Does the proposed change involve a significant reduction in the margin of safety?
  
  Response: No.
  
  The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG Regulatory Guide 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant.
  
  Therefore, this change does not involve a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: David T. Conley, Manager--Senior Counsel--
  
  Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.
  
  NRC Acting Branch Chief: Jessie F. Quichocho.
  
  Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina
  
  Date of amendment request: August 29, 2012.
  
  Description of amendment request: The proposed change combines two changes that affect the same Technical Specification (TS) sections into one license amendment. The first part proposes to implement revisions consistent with Technical Specification Task Force (TSTF)-510, Revision 2, ``Revision to Steam Generator (SG) Program Inspection Frequencies and Tube Sample Selection.'' The second part proposes to permanently revise TS 5.5.9 ``Steam Generator Program'' to exclude portions of the SG tube below the top of the SG tubesheet from periodic inspections by implementing the permanent alternate repair criteria ``H.*'' References 2, 3, 8, 23 and 32 referred to in the licensees analysis can be found in the license amendment request under ADAMS Accession No. ML12275A176.
  
  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:
10. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change modifies steam generator tube inspection frequencies and tube selection consistent with TSTF-510 and excludes the lower portion of steam generator tubes from inspection by implementing the alternate repair criteria (H*) on a permanent basis and does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.
  
  Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed change to the SG tube inspection and repair criteria are the SG tube rupture (SGTR) event and the main steam line break (MSLB) postulated accident.
  
  The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed SG tube inspection frequency and sample selection criteria will not cause the consequences of a SGTR to exceed those assumptions.
  
  With respect to the SGTR event, the required structural integrity margins of the SG tubes and the tube-to-tubesheet joint over the H* distance will be maintained. Tube rupture in tubes with cracks within the tubesheet is precluded by the constraint provided by the presence of the tubesheet and the tube-to-tubesheet joint. Tube burst cannot occur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tube sheet, and from the differential pressure between the primary and secondary side, and tube sheet rotation. The structural margins against burst, as discussed in Regulatory Guide RG 1.121, ``Bases for Plugging Degraded PWR Pressurized-Water Reactor Steam Generator Tubes'' (Reference 32) and Nuclear Energy Institute NEI 97-06, ``Steam Generator Program Guidelines,'' (Reference 8) are maintained for both normal and postulated accident conditions.
  
  For the portion of the tube outside of the tubesheet, the proposed change also has no impact on the structural or leakage integrity. Therefore, the proposed change does not result in a significant increase in the probability of the occurrence of a SGTR accident.
  
  At normal operating pressures, leakage from degradations below the proposed limited inspection depth is limited by the tube-to-
  
  tubesheet crevice. Consequently, negligible normal operating leakage is expected from degradation below the inspected depth within the tubesheet region. The consequences of an SGTR event are affected by the primary to secondary leakage flow during the event. However, primary to secondary leakage flow through a postulated tube that has been pulled out of the tubesheet is not affected by the proposed changes since the tubesheet enhances the tube integrity in the region of the hydraulic expansion by precluding tube deformation beyond its initial hydraulically expanded outside diameter. Therefore, the proposed change does not result in a significant increase in the consequences of an SGTR. In addition, the selected H* value envelopes the depth within the tubesheet required to prevent a tube pullout.
  
  The probability of a MSLB event is unaffected by the potential failure of a SG tube as the failure of a tube is not an initiator for a MSLB event. Therefore the proposed SG tube inspection frequency and sample selection criteria and the structural integrity margins of the SG tubes and the tube-to-tubesheet joint over the H* distance do not increase the probability of a MSLB event.
  
  Page 63349
  
  The leak rate factor of 1.82 for HBRSEP, for a postulated MSLB, has been calculated as shown in References 2, 3 and 23. HBRSEP Unit No. 2 will apply the factor of 1.82 to the normal operating leakage associated with the tubesheet expansion region in the condition monitoring and operational assessment. Through application of the limited tube sheet inspection scope, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) will not occur.
  
  When the TS operational leak rate limit of 75 gallons per day gpd or about 0.052 gallons per minute (gpm) through any one SG is multiplied by the MSLB leak rate factor applicable to HBRSEP Unit No. 2 of 1.82 (Table 9-7 in Westinghouse Commercial Atomic Power Report WCAP-17091-P, Reference 3) the maximum primary to secondary accident induced leak rate is less than 0.095 gpm and is bounded by the value of 0.11 gpm through the faulted SG used in the MSLB accident analyses. Since the existing limit on operational leakage continues to ensure that the MSLB assumed accident induced leakage will not be exceeded, the consequences of a MSLB accident are not increased.
  
  For the condition monitoring assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 1.82 and added to the total leakage from any other source and compared to the allowable accident induced leak rate. For the operational assessment, the difference in the leakage between the allowable leakage and the calculated accident induced leakage from sources other than the tubesheet expansion region will be divided by 1.82 and compared to the observed operational leakage.
  
  Based on the above, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
11. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed change modifies steam generator tube inspection frequencies and tube selection consistent with TSTF-510 and excludes the lower portion of steam generator tubes from inspection by implementing the alternate repair criteria (H*) on a permanent basis. The proposed change does not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures resulting from tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses.
  
  Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
12. Does the change involve a significant reduction in the margin of safety?
  
  Response: No.
  
  The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes. Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes but will limit inspection within the tubesheet to the portion of the tube from the top of the tubesheet to a distance H* below the top of the tubesheet.
  
  The proposed change modifies steam generator tube inspection frequencies and tube selection consistent with TSTF-510 and limits required inspection to the safety significant portion of the steam generator tubes. WCAP-17345, Rev. 2 (Reference 2) identifies the specific inspection depth (H*) below which any type of tube degradation is shown to have no impact on the performance criteria in NEI 97-06 Rev. 3, ``Steam Generator Program Guidelines'' (Reference 8) and TS 5.5.9, ``Steam Generator (SG) Program.'' Changes associated with inspection frequency and tube selection criteria are consistent with TSTF-510 and are based on recent industry experience and are more effective in managing the frequency of verification of tube integrity and sample selection than those required by current TSs.
  
  The proposed change maintains the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute 97-06, ``Steam Generator Program Guidelines'' (Reference 8), and NRC Regulatory Guide 1.121 ``Bases for Plugging Degraded PWR Steam Generator Tubes'' (Reference 32), are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. Regulatory Guide 1.121 describes a method acceptable to the NRC for meeting General Design Criteria (GDC) 14, ``Reactor Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant System Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure Boundary,'' and GDC 32, ``Inspection of Reactor Coolant Pressure Boundary,'' by reducing the probability and consequences of a SGTR. Regulatory Guide 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SGTR are reduced. This Regulatory Guide uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.
  
  For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, Westinghouse WCAP-17091-P, Rev. 0 (Reference 3) and WCAP-17345, Rev. 2 (Reference 2) define a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. When the TS operational leak rate limit of 75 gpd or about 0.052 gpm through any one SG is multiplied by the MSLB leak rate factor applicable to HBRSEP Unit No. 2 of 1.82 (Table 9-7 in WCAP-17091-P (Reference 3) the maximum primary to secondary accident induced leak rate is less than 0.095 gpm and is bounded by the value of 0.11 gpm through the faulted SG used in the MSLB accident analyses.
  
  Therefore, the proposed change does not involve a significant reduction in any margin of safety.
  
  The NRC staff has reviewed the license's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: David T. Conley, Manager--Senior Counsel--
  
  Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.
  
  NRC Acting Branch Chief: Jessie F. Quichocho.
  
  Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin
  
  Date of amendment request: July 30, 2012.
  
  Description of amendment request: The proposed amendment would modify KPS Technical Specifications requirements regarding steam generator tube inspections and reporting as described in Technical Specification Task Force (TSTF)-510, Revision 2, ``Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.''
  
  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:
13. Does the proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
  
  Response: No.
  
  Page 63350
  
  The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.
  
  Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
14. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.
  
  Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
15. Does the proposed change involve a significant reduction in the margin of safety?
  
  Response: No.
  
  The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
  
  Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
  
  Therefore, it is concluded that 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.
  
  NRC Acting Branch Chief: Istvan Frankl.
  
  Entergy Nuclear Operations, Inc., Docket No 50-286, Indian Point Nuclear Generating Unit 3, Westchester County, New York
  
  Date of amendment request: August 14, 2012.
  
  Description of amendment request: The proposed amendment would revise Technical Specification 3.5.4, ``Refueling Water Storage Tank,'' such that the non-seismically qualified piping of the spent fuel pool purification system may be connected to the refueling water storage tank (RWST) seismic piping by manual operation of a RWST seismically qualified boundary valve under administrative controls for a limited period of time (i.e., 14 days per fuel cycle for filtration for removal of suspended solids from the RWST water). This change will only be applicable until Refueling Outage 18 (Spring 2015) ends.
  
  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:
16. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The use of the SFP spent fuel pool Purification Loop to re-
  
  circulate the RWST does not involve any changes or create any new interfaces with the reactor coolant system or main steam system piping. Therefore, the connection of the SFP Purification Loop to the RWST would not affect the probability of these accidents occurring. The SFP Purification Loop is not credited for safe shutdown of the plant or accident mitigation. Administrative controls ensure that the SFP Purification Loop can be isolated as necessary in sufficient time to assure that the RWST volume will be adequate to perform the safety function as designed. Since the RWST will continue to perform its safety function and overall system performance is not affected, the consequences of the accident are not increased.
  
  Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
17. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The design of the RWST and the SFP Purification Loop to allow recirculation and purification has not been altered. Procedures for the operation of the plant have not been revised to create the possibility of a new or different type of accident. Contingent upon manual operator action, a SFP Purification Loop line break will not result in a loss of the RWST safety function. Similarly, an active or passive failure in the SFP Purification Loop will not be significantly different whether aligned to the SFP or the RWST.
  
  Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
18. Does the proposed amendment involve a significant reduction in a margin of safety?
  
  Response: No.
  
  The SFP Purification Loop is not credited for safe shutdown of the plant or accident mitigation. Adequate RWST volume will be maximized prior to purification and timely operator action can be taken to isolate the non seismic system from the RWST to assure it can perform its function. This will result in no significant reduction in the margin of safety.
  
  Therefore, the proposed change does not significantly reduce 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.
  
  Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
  
  NRC Branch Chief: George Wilson.
  
  Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit 3, Westchester County, New York
  
  Date of amendment request: August 14, 2012.
  
  Description of amendment request: The amendment changes the licensing basis regarding the emergency diesel generator (EDG) fuel oil storage requirements. The current licensing basis, which requires sufficient EDG fuel oil to operate two EDGs at minimum safeguards for seven days, will be revised to provide sufficient EDG fuel oil to operate three EDGs at modified rated capacity for seven days. The Conditions, Required Actions, and Completion Times of Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, and Starting Air,'' will be modified
  
  Page 63351
  
  to be more consistent with the Nuclear Regulatory Commission's improved Standard Technical Specifications. Finally, TS 3.8.3 will be modified to relocate specific numerical values for EDG fuel oil storage requirements from the TSs to the TS Bases in accordance with TS Task Force (TSTF) 501 Revision 1.
  
  Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
19. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The proposed change relocates the volume of diesel fuel oil required to support 7 day operation of the onsite diesel generators to licensee control, revises the action statement to reflect the volume equivalent to a 6 day supply and locates the volume in the TS Bases under licensee control, consolidates surveillance requirements and recalculates the fuel oil volume required for the EDG. The revised specific volume of fuel oil equivalent to a 7 and 6 day supply is calculated consistent with the NRC approved methodology described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems for Standby Diesel Generators'' and ANSI N195-1976, ``Fuel Oil Systems for Standby Diesel Generators.'' Because the requirement to maintain a 7 day supply of diesel fuel oil is not changed and is revised to be more consistent with the assumptions in the accident analyses, the consolidated surveillances are more conservative, and the actions taken when the volume of fuel oil is less than a 6 day supply have not changed, neither the probability or the consequences of any accident previously evaluated will be affected.
  
  Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
20. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis but ensures that the diesel generator operates as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions.
  
  Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
21. Does the proposed change involve a significant reduction in a margin of safety?
  
  Response: No.
  
  The proposed change relocates the volume of diesel fuel oil required to support 7 day operation of the onsite diesel generators, revises the action statement to reflect the volume equivalent to a 6 day supply, locates the volume in the TS Bases under licensee control, consolidates surveillance requirements and recalculates the fuel oil volume required for the EDG. Although the bases for the existing limits on diesel fuel oil are changed, no change is made to the accident analysis assumptions and no margin of safety is reduced as part of this change.
  
  Therefore, the proposed change does not involve a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
  
  NRC Branch Chief: George Wilson.
  
  South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, South Carolina
  
  Date of amendment request: August 29, 2012.
  
  Description of amendment request: The proposed change would amend Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, respectively, by adding four non-Class 1E containment electrical penetration assemblies (EPAs). Containment EPAs are a passive extension of containment which provide the passage of the electric conductors through a single aperture in the nuclear containment structure, while providing a pressure barrier between the inside and the outside of the containment structure.
  
  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:
22. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The additional containment EPAs are a passive extension of containment and provide a pathway for communication sic, passage of non-Class 1E electrical signals sic, conductors between the Auxiliary Building and Containment. The proposed containment electrical penetration assemblies are similar in form, fit and function to the current non-Class 1E containment electrical penetration assemblies. The maximum allowable leakage rate allowed by Technical Specifications is also unchanged. The new EPAs will meet the same design function as current EPAs; therefore, the additional penetrations do not involve a significant increase in the probability or consequences of an accident previously evaluated.
23. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed containment electrical penetration assemblies are similar in form, fit, and function to the current non-Class 1E containment electrical penetration assemblies. The new EPAs will meet the same design function as current EPAs. Because the new EPAs are virtually identical in design and function to the current EPAs, no new type of failure modes exist.
  
  Therefore, the proposed change will not create the possibility of a new or different kind of accident.
24. Does the proposed amendment involve a significant reduction in a margin of safety?
  
  Response: No.
  
  The proposed containment electrical penetration assemblies are similar in form, fit and function to the current non-Class 1E containment electrical penetration assemblies. The additional containment electrical penetration assemblies are an engineered passive extension of containment, and, therefore, do not affect containment or its ability to perform its design function. The addition of the new EPAs does not exceed or alter a design basis or safety limit and, therefore, does not significantly reduce 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.
  
  Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
  
  NRC Branch Chief: Mark E. Tonacci.
  
  Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia
  
  Date of amendment request: July 31, 2012.
  
  Description of amendment request: The proposed amendment would modify the Technical Specifications requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.''
  
  Page 63352
  
  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:
25. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  Response: No.
  
  The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.
  
  Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
26. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  Response: No.
  
  The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.
  
  Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
27. Does the proposed change involve a significant reduction in a margin of safety?
  
  Response: No.
  
  The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
  
  Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
  
  Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety. Based on the above, Dominion concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant hazards consideration'' is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied.
  
  Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
  
  NRC Branch Chief: Robert J. Pascarelli.
  
  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 are available for public inspection at the NRC's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr.resource@nrc.gov.
  
  Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut
  
  Date of amendment request: November 17, 2011.
  
  Description of amendment request: The proposed amendment would add Optimized ZIRLO\TM\ as an allowable fuel rod cladding material and add the Westinghouse topical report on Optimized ZIRLO\TM\ to the Millstone Power Station, Unit 3 Technical Specifications. In addition, a typographical error would be corrected.
  
  Date of issuance: September 24, 2012.
  
  Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 253.
  
  Renewed Facility Operating License No. NPF-69: Amendment revised the License and Technical Specifications.
  
  Date of initial notice in Federal Register: May 15, 2012 (77 FR 28629).
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 24, 2012.
  
  No significant hazards consideration comments received: No.
  
  Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut
  
  Date of amendment request: July 23, 2012.
  
  Description of amendment request: The proposed amendment would conform the Millstone Power Station Unit 3 (MPS3) licenses to reflect a name change for Central Vermont Public Service Corporation (CVPS) resulting
  
  Page 63353
  
  from a subsequent restructuring in which CVPS will be consolidated with Gaz Meacutetro's other electric utility subsidiary in Vermont, Green Mountain Power Corporation.
  
  Date of issuance: October 3, 2012.
  
  Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment No.: 254.
  
  Renewed Facility Operating License No. NPF-49: Amendment revised the License.
  
  Date of initial notice in Federal Register: July 20, 2012 (77 FR 42768).
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2012.
  
  No significant hazards consideration comments received: No.
  
  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 26, 2011, as supplemented April 16, 2012.
  
  Brief description of amendments: The amendments revised the Technical Specifications (TSs) to incorporate aspects of NUREG-1431, ``Standard Technical Specifications--Westinghouse Electric Company Plants,'' STS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' Condition E, regarding Diesel Generator starting air receiver pressure limits.
  
  Date of issuance: September 28, 2012.
  
  Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
  
  Amendment Nos.: Unit 1--267 and Unit 2--247.
  
  Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.
  
  Date of initial notice in Federal Register: May 29, 2012 (77 FR 31659).
  
  The supplement dated April 16, 2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.
  
  The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 28, 2012.
  
  No significant hazards consideration comments received: No.
  
  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: December 5, 2011.
  
  Brief description of amendments: The amendments revised Technical Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip System Instrumentation,'' Function 16(e) to replace the phrase ``Turbine Impulse Pressure'' with ``Turbine Inlet Pressure.''
  
  Date of issuance: October 1, 2012.
  
  Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
  
  Amendment Nos.: Unit 1--268 and Unit 2--248.
  
  Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.
  
  Date of initial notice in Federal Register: August 9, 2012 (77 FR 47677).
  
  The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2012.
  
  No significant hazards consideration comments received: No.
  
  Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, Unit 1, Oconee County, South Carolina
  
  Date of application of amendments: April 3, 2012
  
  Brief description of amendments: The amendment revised the Technical Specifications related to the integrated leak rate test of the reactor containment building.
  
  Date of Issuance: October 1, 2012.
  
  Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
  
  Amendment No.: 381.
  
  Renewed Facility Operating License No. DPR-38: Amendment revised the license and the technical specifications.
  
  Date of initial notice in Federal Register: July 10, 2012, 77 FR 40651.
  
  The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2012.
  
  No significant hazards consideration comments received: No.
  
  Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
  
  Date of amendment request: September 12, 2011, as supplemented by letters dated October 13, 2011, March 22, 2012, and April 3, 2012.
  
  Brief description of amendment: The amendment approved a change to the site Emergency Plan to relocate the existing backup emergency operations facility for RBS from its current location at the Entergy Operations-Baton Rouge Division Office, located at 1509 Government Street in Baton Rouge, Louisiana, approximately 23 miles southeast of RBS, to the Entergy Customer Service Center, located at 5564 Essen Lane in Baton Rouge, Louisiana, approximately 28 miles southeast of RBS.
  
  Date of issuance: September 24, 2012.
  
  Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.
  
  Amendment No.: 175.
  
  Facility Operating License No. NPF-47: The amendment revised the RBS Emergency Plan.
  
  Date of initial notice in Federal Register: December 27, 2011 (76 FR 80975). The supplemental letters dated October 13, 2011, March 22, 2012, and April 3, 2012, 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 September 24, 2012.
  
  No significant hazards consideration comments received: No.
  
  Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York
  
  Date of application for amendment: August 16, 2011, as supplemented by letters dated March 30, June 13, August 1, August 16, and September 14, 2012.
  
  Brief description of amendment: The amendment revises the James A. FitzPatrick's (JAF's) current licensing basis, in the Updated Final Safety Analysis Report, to support installation of new reserve station service transformers (RSST) with on-load tap changers (OLTC). The new RSSTs with OLTCs will compensate for the wider range of offsite power voltage variations so that acceptable voltages at the safety-related equipment will be better maintained. The new RSSTs provided with OLTCs would facilitate operations in the automatic mode.
  
  The OLTCs are sub-components of two new RSSTs that will be installed at JAF during the refueling outage scheduled for September 2012.
  
  Date of issuance: September 26, 2012.
  
  Effective date: As of the date of issuance, and shall be implemented
  
  Page 63354
  
  within 90 days The implementation of the amendment shall include revision of the Updated Final Safety Analysis Report as described in the licensee's application for this amendment.
  
  Amendment No.: 302.
  
  Renewed Facility Operating License No. DPR-59: The amendment revised the License and the Updated Final Safety Analysis Report.
  
  Date of initial notice in Federal Register: November 15, 2011 (76 FR 70768).
  
  The supplements dated March 30, June 13, August 1, August 16, and September 14, 2012 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 26, 2012.
  
  No significant hazards consideration comments received: No.
  
  FirstEnergy Nuclear Operating Company, Docket Nos. 50-334 and 50-412, Beaver Valley Power Station (BVPS), Units 1 and 2, Beaver County, Pennsylvania; Docket No. 50-346, Davis-Besse Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio; Docket No. 50-440, Perry Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
  
  Date of application for amendments: September 20, 2011, as supplemented by letter dated June 21, 2012.
  
  Brief description of amendments: The proposed amendments would change the licenses of BVPS, Units 1 and 2, DBNPS, and PNPP to reflect the name change of an owner licensee from ``FirstEnergy Nuclear Generation Corp.'' to ``FirstEnergy Nuclear Generation, LLC.'' The proposed amendment is administrative in nature. The proposed amendment will also correct errors regarding the name of FirstEnergy Nuclear Generation Corp in the DBNPS and PNNP Facility Operating Licenses.
  
  Date of issuance: October 2, 2012.
  
  Effective date: At date of issuance.
  
  Amendment Nos.: No. 290 to Facility Operating License No. DPR-66 and No. 177 to Facility Operating License No. NPF-73 for BVPS, Units 1 and 2, respectively, and No. 286 to Facility Operating License No. NPF-
  
  3 for DBNPS, and No. 161 to Facility Operating License No. NPF-58 for PNPP.
  
  Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58: The amendments revised the Facility Operating Licenses.
  
  Date of initial notice in Federal Register: October 18, 2011 (76 FR 64391).
  
  The supplemental letter dated June 21, 2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed finding of no significant hazards consideration determination published in the Federal Register.
  
  The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 2, 2012.
  
  No significant hazards consideration comments received: No.
  
  Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie Plant, Unit 2, St. Lucie County, Florida.
  
  Date of application for amendment: February 25, 2011, as supplemented by the letters dated November 4 and December 8, 2011, and April 30 and May 4 and 7, 2012.
  
  Brief description of amendment: This amendment raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight percent uranium-235 to a maximum lattice averaged value of 4.6 weight percent uranium-235. The Technical Specification changes associated with fuel stored in the spent fuel pool include increasing the maximum initial enrichment from 4.5 weight percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies, and definition of three special configurations referred to in the nuclear criticality safety analysis as inspection and maintenance configurations.
  
  Date of issuance: September 18, 2012.
  
  Effective date: This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
  
  Amendment No: 162.
  
  Renewed Facility Operating License No. NPF-16: Amendment revised the Operating License and the Technical Specifications.
  
  Date of initial notice in Federal Register: September 1, 2011 (76 FR 54503).
  
  The supplemental letters provided additional information that clarified the application and did not expand the scope of the application as originally noticed and published in the Federal Register.
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 18, 2012.
  
  No significant hazards consideration comments received: No.
  
  Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie Plant, Unit 2, St. Lucie County, Florida.
  
  Date of application for amendment: February 25, 2011 as supplemented by letters dated February 25, May 24, July 22, August 18 (three letters), August 25 (three letters), August 29, September 2, September 8 (two letters), September 22, October 5, October 10, October 12 (two letters), October 31, November 2, November 3, November 4, November 7, November 14 (three letters), November 23 (three letters), December 8, December 14, December 20, December 27, December 29, 2011, January 14, 2012 (two letters), January 18 (two letters), January 21 (two letters), February 29, March 6 (two letters), March 8, March 15, March 16, March 17 (two letters), March 25, March 31 (two letters), April 5 (two letters), April 6, April 10, April 19 (seven letters), April 30, May 4, May 7, May 18, and July 23, 2012.
  
  Brief description of amendments: The proposed amendments would increase the licensed core power level for St. Lucie Unit 2 from 2070 megawatts thermal (MWt) to 3020 MWt. This represents a net increase in the core thermal power of approximately 11.85 percent, including a 10-
  
  percent power uprate and a 1.7 percent measurement uncertainty recapture, over the current licensed thermal power level and is defined as an extended power uprate. The proposed amendments would change the renewed facility operating license and the technical specifications (TSs) to support operation at the increased core thermal power level, including changes to the maximum licensed reactor core thermal power, reactor core safety limits, and reactor protection system and engineered safety feature actuation system limiting safety system settings. Additional TS changes include reactor coolant system heatup and cooldown limitations, accumulator and refueling water storage tank boron concentrations, main steam safety valve lift settings, emergency diesel generator fuel storage and core operating limits report references. A complete list of the proposed TS changes and the licensee's basis for change can be found in Attachment 1 of the licensee's application (ADAMS Accession No. ML110730116).
  
  Date of issuance: September 24, 2012.
  
  Page 63355
  
  Effective date: This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
  
  Amendment No.: 163.
  
  Renewed Facility Operating License No. NPF-16: Amendment revised the Operating License and the Technical Specifications.
  
  Date of initial notice in Federal Register: September 1, 2011 (76 FR 54503).
  
  The supplemental letters provided additional information that clarified the application and did not expand the scope of the application as originally noticed and published in the Federal Register.
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 24, 2012.
  
  No significant hazards consideration comments received: No.
  
  Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska
  
  Date of amendment request: September 16, 2011, as supplemented by letters dated May 2, May 24, and September 17, 2012.
  
  Brief description of amendment: The amendment revised the Cooper Nuclear Station Technical Specifications (TS) and Operating License to implement a 24-month fuel cycle and adopt TS Task Force (TSTF) Traveler TSTF-493, Revision 4, ``Clarify Application of Setpoint Methodology for LSSS Limiting Safety System Settings Functions,'' Option A. Specifically, the amendment revised certain TS Surveillance Requirement frequencies that are specified as ``18 months'' by changing them to ``24 months'' in accordance with the guidance provided in NRC Generic Letter 91-04, ``Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
  
  Date of issuance: September 28, 2012.
  
  Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
  
  Amendment No.: 242.
  
  Renewed Facility Operating License No. DPR-46: Amendment revised the Facility Operating License and Technical Specifications.
  
  Date of initial notice in Federal Register: March 6, 2012 (77 FR 13371). The supplemental letters dated May 2, May 24, and September 17, 2012, 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 September 28, 2012.
  
  No significant hazards consideration comments received: No.
  
  NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa
  
  Date of application for amendments: May 1, 2012, as supplemented by letters dated June 27, 2012, and July 26, 2012.
  
  Brief description of amendments: The amendment revises existing TS 3.3.5.1, on a one-time basis only, by adding a note to TS Table 3.3.5.1-1, Function 1d, Modes 4 and 5. This one-time license amendment enables DAEC to re-coat the internal surface of the Suppression Chamber during Refueling Outage 23.
  
  Date of issuance: September 27, 2012.
  
  Effective date: This license amendment is effective as of the date of issuance and shall be implemented within 30 days from date of issuance.
  
  Amendment No.: 283.
  
  Renewed Facility Operating License No. DPR-49: Amendment revised the Renewed Facility Operating License and Technical Specifications.
  
  Date of initial notice in Federal Register: July 10, 2012 (77 FR 40654).
  
  The licensee's June 27, 2012, and July 26, 2012, supplemental letters contained clarifying information, did not change the scope of the original amendment request, did not change the NRC staff's initial proposed finding of no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.
  
  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 27, 2012.
  
  No significant hazards consideration comments received: Yes.
  
  Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
  
  Date of application for amendment: June 17, 2011, supplemented by letters dated July 27, 2011, November 14, 2011, March 23, April 26, May 15, May 24, and June 26, 2012 (TS-SQN-2011-07).
  
  Brief description of amendment: The amendments revised the licensing basis and the Technical Specifications (TSs) to permit the use of a more robust AREVA Advanced W17 high thermal performance fuel. This new fuel has been selected to address fuel assembly distortion and its resultant fuel-handling issues. The AREVA Advanced W17 HTP fuel assembly design consists of standard uranium dioxide fuel pellets with gadolinium oxide burnable poison and M5TM cladding. The new fuel design ensures mechanical compatibility with the existing fuel, reactor core, control rods, steam supply system, and fuel-handling system. The transition from the existing fuel (AREVA Mark-BW) to new fuel is planned to occur over two refueling cycles for each unit.
  
  Date of issuance: September 26, 2012.
  
  Effective date: As of the date of issuance and shall be implemented prior to startup from fall 2013 refueling outage (RFO) for Unit 1, and prior to startup from fall 2012 RFO for Unit 2.
  
  Amendment Nos.: 331 and 324.
  
  Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the License and TSs.
  
  Date of initial notice in Federal Register: August 23, 2011 (76 FR 52703). The supplement letters dated July 27, 2011, November 14, 2011, March 23, April 26, May 15, May 24, and June 26, 2012, 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 amendment is contained in a Safety Evaluation dated September 26, 2012.
  
  No significant hazards consideration comments received: No.
  
  For The Nuclear Regulatory Commission.
  
  Dated at Rockville, Maryland, this 5th day of October 2012.
  
  Michele G. Evans,
  
  Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
  
  FR Doc. 2012-25240 Filed 10-15-12; 8:45 am
  
  BILLING CODE 7590-01-P

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Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

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