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Notice

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

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Information about this document as published in the Federal Register.

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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 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 July 28, 2011, to August 10, 2011. The last biweekly notice was published on August 9, 2011 (76 FR 48908).

ADDRESSES:

Please include Docket ID NRC-2011-0187 in the subject line of your comments. Comments submitted in writing or in electronic form will be posted on the NRC Web site and on the Federal rulemaking Web site http://www.regulations.gov. Because your comments will not be edited to remove any identifying or contact information, the NRC cautions you against including any information in your submission that you do not want to be publicly disclosed.

The NRC requests that any party soliciting or aggregating comments received from other persons for submission to the NRC inform those persons that the NRC will not edit their comments to remove any identifying or contact information, and therefore, they should not include any information in their comments that they do not want publicly disclosed.

You may submit comments by any one of the following methods.

  • Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-2011-0187. Address questions about NRC dockets to Carol Gallagher 301-492-3668; e-mail Carol.Gallagher@nrc.gov.
  • Mail comments to: 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.

You can access publicly available documents related to this notice using the following methods:

  • NRC's Public Document Room (PDR): The public may examine and have copied, for a fee, publicly available documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.
  • NRC's Agencywide Documents Access and Management System (ADAMS): 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. From this page, the public can gain entry into ADAMS, which provides text and image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC's PDR reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.
  • Federal Rulemaking Web Site: Public comments and supporting materials related to this notice can be found at http://www.regulations.gov by searching on Docket ID: NRC-2011-0187.Start Printed Page 52700

Notice of Consideration of Issuance of Amendments to Facility Operating 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 Title 10 of the Code of Federal Regulations (10 CFR), Section 50.92, 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. 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. NRC regulations are accessible electronically from the NRC Library on the NRC Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC 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 e-mail at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is Start Printed Page 52701participating; 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 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 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 Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC 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 e-mail 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 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. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

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 PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 13, 2011.

Description of amendment request: The proposed amendment would modify the Technical Specifications (TSs) as a result of a revised Fuel Handling Accident analysis. The new analysis determined that the current TSs may not be conservative for all scenarios. The proposed amendment would provide new applicability and/or action language in the TSs that includes load movements over irradiated fuel assemblies. Specifically, the amendment would modify the following TSs: TS 3.3.3.1 (Radiation Monitoring Instrumentation); TS 3.7.6.1 (Control Room Emergency Air Filtration System); TS 3.7.6.3 (Control Room Air Temperature—Operating); TS 3.7.6.4 (Control Room Air Temperature—Shutdown); TS 3.8.1.2 (A.C. Start Printed Page 52702[Alternating Current] Sources—Shutdown); TS 3.8.2.2 (DC Sources [Direct Current]—Shutdown); TS 3.8.3.2 (On Site Power Distribution—Shutdown); TS 3.9.3 (Decay Time); TS 3.9.4 (Containment Building Penetrations); and TS 3.9.7 (Crane Travel—Fuel Handling Building).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

This proposed change revises Technical Specifications applicability wording regarding the movement of fuel assemblies in containment and the fuel storage pool to include load movements over irradiated fuel assemblies. The proposed applicability is more comprehensive than the current Applicability. This change was driven by an analysis change and was not due to fuel handling equipment or fuel movement methods. Expanding the applicability of the relevant Technical Specifications is necessary to account for updated fuel drop analyses which demonstrate that the impacted spent fuel assemblies may be damaged.

Consequently, dropping of a non-irradiated fuel assembly, dummy fuel assembly, or other load could result in a Fuel Handling Accident that has radiological consequences. Changing the applicability of the relevant Technical Specifications does not affect the probability of a Fuel Handling Accident. The expanded applicability provides assurance that equipment designed to mitigate a Fuel Handling Accident is capable of performing its specified safety function.

The dose consequences due to failure of two assemblies remain within the Regulatory Guide 1.183 and 10 CFR 50.67 acceptance criteria limits. The Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Main Control Room (MCR) dose results and associated regulatory limits are presented below.

New analysisRegulatory guide 1.183 limit10 CFR 50.67 limit
EAB4.56 rem TEDE<6.3 rem TEDE<25 rem TEDE.
LPZ0.70 rem TEDE<6.3 rem TEDE<25 rem TEDE.
MCR0.824 rem TEDE<5 rem TEDE<5 rem TEDE.

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

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

Response: No.

The revised spent fuel handling analyses demonstrate that the impacted fuel assemblies may be damaged as the result of a dropped fuel assembly, dummy assembly, or load. The existing Technical Specifications regarding movement of fuel assemblies are not applicable for movement of non-irradiated fuel assemblies or other loads. A drop of these loads could cause radiological consequences during periods when the equipment required to mitigate those consequences is not required to be OPERABLE in accordance with the existing Technical Specifications.

The proposed changes to the Technical Specifications applicability language regarding the movement of these loads in containment and the fuel storage pool ensure that Limiting Conditions of Operation and appropriate Required Actions for required equipment are in effect during fuel movement. This provides assurance that the Fuel Handling Accident will remain within the initial assumptions of accident analyses.

Consequently, there is no possibility of a new or different kind of accident due to this change.

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

Response: No.

The proposed Technical Specifications change will not affect protection criterion for plant equipment and will not reduce the margin of safety. By extending the Applicability to the movement of non-irradiated fuel assemblies, the current margin of safety is maintained.

Consequently, there is no significant reduction in a margin of safety due to this change.

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

Attorney for licensee: Joseph A. Aluise, Associate General Council—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: June 2, 2011, as supplemented by letter dated August 1, 2011.

Description of amendment request: The proposed amendment would approve revision to the South Texas Project (STP), Units 1 and 2, Fire Protection Program related to the alternate shutdown capability. Specifically, STP Nuclear Operating Company (STPNOC) proposes to credit the following manual operator actions in the control room prior to evacuation due to a fire for meeting the alternate shutdown capability:

  • Main steam line isolation.
  • Closing the pressurizer power-operated relief valves block valves.
  • Securing all reactor coolant pumps.
  • Feedwater isolation.
  • Securing the startup feedwater pump.
  • Letdown isolation.
  • Securing the charging pumps.

In addition, STPNOC proposes to credit the automatic trip of the main turbine upon the initiation of a manual reactor trip for meeting the alternate shutdown capability. A thermal-hydraulic analysis will demonstrate that these operations will ensure that the reactor coolant system (RCS) process variables remain within those values predicted for a loss of normal alternating current (a-c) power, as required by Section III.L.1 of Appendix R of Title 10 of the Code of Federal Regulations (10 CFR) part 50.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The design function of structures, systems and component are not impacted by the proposed change. The proposed change involves crediting operations in the control room prior to evacuation in the event of a fire in order to meet safe shutdown performance criteria. The proposed action will not initiate an event. The proposed actions do not increase the probability of occurrence of a fire or any other accident previously evaluated.

The proposed operations are feasible and reliable and demonstrate that the unit can be safely shutdown in the event of a fire. No significant consequences result from the performance of the proposed operations.

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

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

Response: No.

The design function of structures, systems and component are not impacted by the proposed amendment. The proposed change involves operations in response to a fire. They do not involve new failure mechanisms Start Printed Page 52703or malfunctions that can initiate a new accident.

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

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

Response: No.

Thermal-hydraulic analysis demonstrates that the proposed operations to be performed in the control room will ensure that the RCS process variables remain within those values predicted for a loss of normal a-c power, as required by 10 CFR 50, Appendix R, Section III.L.1. The analysis demonstrates that a single spurious operation before control of the plant is achieved through the alternative or dedicated shutdown system will not adversely impact the results of the analysis. After control of the plant is achieved by the alternative or dedicated shutdown system, circuits subjected to fire-induced circuit failures are isolated from the control stations such that the safe shutdown operations will not be compromised.

The need to perform the proposed operations can be readily diagnosed and the operations can be performed in rapid succession by control room operators at their normal control station. The actions are straightforward and familiar to the operators. The actions have been verified that they can be performed through demonstration. The operations are backed up outside the control room such that assurance exists they should not be negated by subsequent spurious actuation signals from a postulated fire.

The automatic turbine trip action can reasonably be assumed to occur with the credited manual reactor trip action that is part of the current licensing basis.

Considerable defense-in-depth features exist in Fire Area 1 [control room is part of Fire Area 1] such that it is extremely unlikely that a fire would result in evacuation of the control room.

The proposed operations are feasible and reliable and demonstrate that the unit can be safely shutdown in the event of a fire. The operations ensure that performance goals of Appendix R, Section III.L.2 are met. The achievement of these goals provide adequate margin from challenging any safety limits.

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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Branch Chief: Michael T. Markley.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: June 17, 2011 (TS-SQN-2011-07).

Description of amendment request: The proposed amendment would revise the licensing basis and the Technical Specifications to permit the use of a more robust AREVA Advanced W17 high thermal performance (HTP) fuel at Sequoyah Nuclear Plant (SQN), Units 1 and 2. This new fuel has been selected to address fuel assembly distortion and its resultant fuel handling issues. The proposed 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 (AREVA Advanced W17 HTP) is planned to occur over two refueling cycles for each SQN unit.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The reactor fuel and the analyses associated with it are not accident initiators. The response of the fuel to an accident is analyzed using conservative techniques and the results are compared to approved acceptance criteria. These evaluation results will show that the fuel response to an accident is within approved acceptance criteria for cores loaded with the new AREVA Advanced W17 HTP fuel and cores loaded with both AREVA Advanced W17 HTP and AREVA Mark-BW fuel. Therefore, the change in fuel design does not affect accident or transient initiation or consequences.

The addition of limits on DNBR [departure from nucleate boiling ratio] and maximum local fuel pin centerline temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL [core safety limit] lines are consistent with the Standard Review Plan (SRP) for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with Nuclear Regulatory Commission (NRC) regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not impact the initiation or the mitigation of an accident.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN Unit 1 and that will be installed concurrently with the introduction of the new Advanced W17 HTP fuel for SQN Unit 2). Core flow is not an accident initiator and does not play a role in accident mitigation.

The core operating limits to be developed using the new methodologies will be established in accordance with the applicable limitations as documented in the appropriate NRC Safety Evaluation reports. The proposed change to add and remove various topical reports cited in Technical Specification 6.9.1.14.a (including adding revision numbers and revision dates to current cited topical reports) enables the use of appropriate methodologies to re-analyze certain events. The proposed methodologies will ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed change to the list of NRC-approved methodologies listed in Technical Specification 6.9.1.14.a is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident. The proposed change will update the listing of NRC-approved methodologies consistent with the transition to AREVA Advanced W17 HTP fuel. Changes to the calculated core operating limits may only be made using NRC-approved methods, must be consistent with all applicable safety analysis limits and are controlled by the 10 CFR 50.59 process. The list of methodologies in the Technical Specifications does not impact either the initiation of an accident or the mitigation of its consequences.

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

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

Response: No.

Use of AREVA Advanced W17 HTP fuel in the SQN, Units 1 and 2, reactor cores does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The operational characteristics of AREVA Advanced W17 HTP fuel are bounded by the safety analyses. The AREVA Advanced W17 HTP fuel design performs within fuel design limits and does not create the possibility of a new or different type of accident.

The addition of limits on DNBR and maximum local fuel pin centerline Start Printed Page 52704temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL lines are consistent with the SRP for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not affect any accident initiators that would create a new accident.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN, Unit 1, and that will be installed concurrently with the introduction of the new Advanced W17 HTP fuel for SQN, Unit 2). Core flow is not an accident initiator and does not play a role in accident mitigation and cannot create the possibility of a new or different kind of accident.

The proposed change to the list of topical reports used to determine the core operating limits is administrative in nature and has no impact on any plant configuration or on system performance. It updates the list of NRC-approved topical reports used to develop the core operating limits. There is no change to the parameters within which the plant is normally operated. The possibility of a new or different accident is not created.

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

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

Response: No.

Use of AREVA Advanced W17 HTP fuel does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The operational characteristics of AREVA Advanced W17 HTP fuel are bounded by the safety analyses. The AREVA Advanced W17 HTP fuel design performs within fuel design limits. The proposed changes do not result in exceeding design basis limits. Therefore, the licensed safety margins are maintained.

The addition of limits on DNBR and maximum local fuel pin centerline temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL lines are consistent with the SRP for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not impact licensed safety margins.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN Unit 1 and that will be installed concurrently with the introduction of the new Advanced W17 HTP fuel for SQN Unit 2). The proposed changes to core flow are provided to ensure licensed safety margins are maintained.

The proposed change to the list of topical reports in Technical Specification 6.9.1.14.a does not amend the cycle specific parameters presently required by the Technical Specifications. The individual Technical Specifications continue to require operation of the plant within the bounds of the limits specified in the COLR [core operating limits report]. The proposed change to the list of analytical methods referenced in the COLR is administrative in nature and does not impact the margin of safety.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.

NRC Branch Chief: Douglas A. Broaddus.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 23, 2011.

Description of amendment request: The proposed amendment would revise the Wolf Creek Generating Station Technical Specifications (TSs) 3.3.7, “Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation,” 3.3.8, “Emergency Exhaust System (EES) Actuation Instrumentation,” 3.7.10, “Control Room Emergency Ventilation System (CREVS),” 3.7.11, “Control Room Air Conditioning System (CRACS),” 3.7.13, “Emergency Exhaust System (EES),” 3.8.2, “AC [Alternating Current] Sources—Shutdown,” 3.8.5, “DC [Direct Current] Sources—Shutdown,” 3.8.8, “Inverters—Shutdown,” and 3.8.10, “Distribution Systems—Shutdown.” Specifically, the proposed amendment would: (1) Delete MODES 5 and 6 from the Limiting Condition for Operation (LCO) Applicability for the CREVS and its actuation instrumentation (TS 3.7.10 and TS 3.3.7, respectively); (2) delete the Required Action from TS 3.7.10 and TS 3.7.11 that requires verifying that the OPERABLE CREVS/CRACS train is capable of being powered by an emergency power source; (3) revise TS 3.7.13 by incorporating a 7-day Completion Time for restoring an inoperable EES train to OPERABLE status during shutdown conditions; (4) adopt NRC-approved Technical Specification Task Force (TSTF) Change Traveler TSTF-36-A, Revision 4, “Addition of LCO 3.0.3 N/A [not applicable] to shutdown electrical power specifications,” for TSs 3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 3.8.10; and (5) add a more restrictive change to the LCO Applicability for TSs 3.8.2, 3.8.5, 3.8.8, and 3.8.10 such that these LCOs apply not only during MODES 5 and 6, but also during the movement of irradiated fuel assemblies regardless of the MODE in which the plant is operating.

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.

Deleting MODES 5 and 6 from the LCO Applicability of TSs 3.3.7 and 3.7.10 does not significantly increase the consequences of any accident since it has been demonstrated that the radiological consequences to control room occupants from a waste gas decay tank rupture will remain much less than the regulatory limits with no mitigation from the CREVS in MODES 5 and 6. The acceptance criteria for this event will continue to be met.

Incorporation of a 7-day Completion Time for restoring an inoperable EES train during shutdown conditions (i.e., during movement of irradiated fuel assemblies in the fuel building) and the deletion of Required Actions for verifying the availability of an emergency power source when a CREVS/CRACS train is inoperable during the same conditions, are operational provisions that have no impact on the frequency of occurrence of the event for which the EES, CREVS and CRACS are designed to mitigate. These systems have no bearing on the occurrence of a fuel handling accident [(FHA)] as the systems themselves are not associated with any of the potential initiating sequences, mechanisms or occurrences—such as a failure of a lifting device or crane, or an operator error—that could cause an FHA. Since these systems are designed only to respond to an FHA as accident mitigators Start Printed Page 52705after the accident has occurred, and they have no bearing on the occurrence of such an event themselves, the proposed changes to the CREVS, CRACS, and EES Technical Specifications have no impact on the probability of an accident previously evaluated.

With respect to deleting the noted Required Actions in TS 3.7.10 and TS 3.7.11 (for verifying that the OPERABLE CREVS/CRACS train is capable of being powered from an emergency power source when one CREVS/CRACS train is inoperable), such a change does not change the LCO requirement for both CREVS/CRACS trains to be OPERABLE, nor to the LCO requirements of the TS requirements pertaining to electrical power sources/support for shutdown conditions. The change to the Required Actions would thus not be expected to have a significant impact on the availability of the CREVS and CRACS. That is, adequate availability may be still assumed such that these systems would continue to be available to provide their assumed function for limiting the dose consequences of an FHA in accordance with the accident analysis currently described in the [Updated Safety Analysis Report].

With respect to the Completion Time for an inoperable EES train, the consequences of a postulated accident are not affected by equipment Completion Times as long as adequate equipment availability is maintained. The proposed EES Completion Time is based on the Completion Time specified in the Standard Technical Specifications (STS) for which it may be presumed that the specified Completion Time is acceptable and supports adequate EES availability. As noted in the STS Bases, the 7-day Completion Time for restoring an inoperable EES train takes into account the availability of the other train. Since the STS-support Completion Time supports adequate EES availability, it may be assumed that the EES function would be available for mitigation of an FHA, thus limiting offsite dose to within the currently calculated values based on the current accident analysis. On this basis, the consequences of applicable, analyzed accidents (i.e., the FHA) are not increased by the proposed change.

The adoption of TSTF-36-A will not affect the equipment and LCOs needed to mitigate the consequences of a[n] FHA in the fuel building; however, this change will reduce the chances of an unnecessary plant shutdown due to activities in the fuel building that have no bearing on the operation of the rest of the plant and the reactor core inside the containment building.

[redundant paragraph omitted]

The changes to the shutdown electrical specifications will add an additional restriction that is consistent with the objective of being able to mitigate a fuel handling accident during all situations, including a full core offload, in which such an accident could occur.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. All design, material, and construction standards that were applicable prior to this amendment request will be maintained. There will be no changes to any design or operating limits.

The proposed changes will not adversely affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed changes do not physically alter safety-related systems nor affect the way in which safety related systems perform their functions. The proposed changes do not alter plant design or operation; therefore, these changes will not increase the probability of any accident.

All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. After a postulated release from a waste gas decay tank rupture no CREVS mitigation is required. The applicable radiological dose criteria will continue to be met.

Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

There are no proposed design changes nor are there any changes in the method by which any safety related plant SSC performs its specified safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. Equipment performance necessary to fulfill safety analysis missions will be unaffected. The proposed changes will not alter any assumptions required to meet the safety analysis acceptance criteria.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety related system as a result of this amendment.

The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems.

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

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

Response: No.

There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor [ ], nuclear enthalpy rise hot channel factor [ ], loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met. It has been demonstrated that the CREVS and its actuation instrumentation are not required to mitigate the control room radiological consequences of a waste gas decay tank rupture.

The proposed changes do not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications. None of the acceptance criteria for any accident analysis will be changed.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: April 22, 2011.

Description of amendment request: The proposed amendment would revise the Wolf Creek Generating Station Technical Specification (TS) 5.3, “Unit Staff Qualifications,” by making two administrative changes to TS 5.3.1.1. Specifically, these changes will remove the operator license applicants' education and experience eligibility requirements, and correct inadvertent omissions in previous amendments relative to the Licensed Operators' and Senior Operators' qualification requirements.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change is an administrative change to reinstate the qualification requirements for Licensed Operators and Senior Licensed Operators that were inadvertently eliminated through the issuance of Amendment No. 150 [issued Start Printed Page 52706November 26, 2002] and Amendment No. 159 [issued January 31, 2005], and to remove an unnecessary reference to a [National Academy for Nuclear Training] NANT guideline. The proposed change does not directly impact accidents previously evaluated. [Wolf Creek Nuclear Operating Company's (WCNOC's)] licensed operator training program is accredited by the NANT and is based on a systems approach to training consistent with the requirements of 10 CFR Part 55. Although licensed operator qualifications and training may have an indirect impact on accidents previously evaluated, the NRC considered this impact during the rulemaking process, and by promulgation of the revised 10 CFR Part 55 rule, concluded that this impact remains acceptable as long as the licensed operator training program is certified to be accredited and is based on a systems approach to training.

Therefore, the proposed change will not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change is an administrative change to reinstate the qualification requirements for Licensed Operators and Senior Licensed Operators that were inadvertently eliminated through the issuance of Amendment No. 150 and Amendment No. 159, and to remove an unnecessary reference to a NANT guideline. WCNOC's licensed operator training program is accredited by the National Academy for Nuclear Training and is based on a systems approach to training consistent with the requirements of 10 CFR Part 55. Although licensed operator qualifications and training may have an indirect impact on accidents previously evaluated, the NRC considered this impact during the rulemaking process, and by promulgation of the revised 10 CFR Part 55 rule, concluded that this impact remains acceptable as long as the licensed operator training program is certified to be accredited and is based on a systems approach to training.

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

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

Response: No.

The proposed change is an administrative change to reinstate the qualification requirements for Licensed Operators and Senior Licensed Operators that were inadvertently eliminated through the issuance of Amendment No. 150 and Amendment No. 159, and to remove an unnecessary reference to a NANT guideline. As noted previously, WCNOC's licensed operator training program is accredited and is based on a systems approach to training consistent with the requirements of 10 CFR Part 55. Licensed operator qualifications and training can have an indirect impact on the margin of safety. However, the NRC considered this impact during the rulemaking process, and by promulgation of the revised 10 CFR Part 55 rule, determined that this impact remains acceptable when licensees maintain a licensed operator training program that is accredited and based on a systems approach to training.

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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Notice of Issuance of Amendments to Facility Operating 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.

Notice of Consideration of Issuance of Amendment to Facility Operating License, 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' 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 Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr.resource@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendment: July 22, 2010, as supplemented by letter dated April 8, 2011.

Brief description of amendment: The amendment revised an element of the methodology used in evaluating the radiological consequences of design basis steam generator tube rupture (SGTR) accidents. Specifically, the amendment revised the Palo Verde Nuclear Generating Station (PVNGS) Updated Final Safety Analysis Report Section 15.6.6, “Steam Generator Tube Rupture,” to reflect a lower iodine spiking factor assumed for the coincident event Generated Iodine Spike (GIS) and the resulting reduction in the radiological consequences for the Limiting SGTRLOPSF [Steam Generator Tube Rupture with Loss of Offsite Power and Single Failure] Event.

Date of issuance: July 28, 2011.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: Unit 1—186; Unit 2—186; Unit 3—186.

Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Operating Licenses and the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: December 28, 2010 (75 FR 81669).

The supplemental letter dated April 8, 2011, 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.Start Printed Page 52707

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, Darlington County, South Carolina

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant Citrus County, Florida

Date of application for amendments: July 8, 2010, as supplemented by letters dated September 23 and November 30, 2010; February 28 and April 7, 2011.

Brief description of amendments: The amendments establish a fleet Cyber Security Plan (CSP) in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 73.54, “Protection of digital computer and communication systems and networks,” and in conformance with the model CSP contained in Appendix A of Nuclear Energy Institute (NEI) document NEI 08-09, “Cyber Security Plan for Nuclear Power Reactors,” Revision 6, dated April 2010. The licensees' submittals included the fleet CSP for Brunswick Steam Electric Plant, Units 1 and 2, H. B. Robinson Steam Electric Plant, Unit No. 2, Shearon Harris Nuclear Power Plant, Unit 1, and Crystal River Unit 3 Nuclear Generating Plant, the licensees' proposed changes to the facility operating licenses, and a proposed CSP implementation schedule for each facility.

The licensees' submittals dated November 30, 2010, and April 7, 2011, supplemented the licensees' CSP to address: (1) Scope of systems in response to the October 21, 2010, the Nuclear Regulatory Commission (NRC, Commission) decision; (2) records retention; and (3) implementation schedule. The licensee provided, in its letter dated April 7, 2011, a revised copy of the Carolina Power & Light Company and Florida Power Corporation, Cyber Security Plan, Revision 0 that incorporated all of the changes that the licensee had made to the following sections of their CSP: Scope and purpose, defense-in-depth protective strategies, document control and records retention and handling, and deviations from NEI 08-09, Revision 6.

Date of issuance: July 29, 2011.

Effective date: The license amendments are effective as of the date of their issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensees on April 7, 2011, and approved by the NRC staff with the license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: Brunswick 1: 258, Brunswick 2: 286, Robinson 2: 226, Shearon Harris 1: 136, and Crystal River 3: 238.

Renewed Facility Operating License Nos. DPR-71, DPR-62, DPR-23, and NPF-63; and Facility Operating License No. DPR-72.: Amendments changed the facility operating licenses.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62595).

The supplements dated September 23 and November 30, 2010; February 28, 2011, and the Updated No Significant Hazards Consideration in Enclosure 5 of the letter dated April 7, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a safety evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of application for amendment: July 27, 2010, as supplemented by letters dated September 29, 2010, November 22, 2010, and March 30, 2011.

Brief description of amendment: The amendment approves the cyber security plan and associated implementation schedule, and revises Paragraph 2.E of Facility Operating License No. NPF-43 for Fermi 2, to provide a license condition to require the licensee to fully implement and maintain in effect all provisions of the NRC-approved Cyber Security Plan. The proposed change is consistent with Nuclear Energy Institute (NEI) 08-09, Revision 6, Cyber Security Plan for Nuclear Power Reactors.

Date of issuance: July 28, 2011.

Effective date: This license amendment is effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on July 27, 2010, as supplemented by letters dated September 29, 2010, November 22, 2010, and March 30, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment No.: 185.

Facility Operating License No. NPF-43: Amendment revised the License.

Date of initial notice in Federal Register: December 7, 2010 (75 FR 76043).

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 27, 2011.

No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin

Date of application for amendment: June 1, 2010, as supplemented by letters dated January 18, 2011, March 14, 2011, and June 27, 2011.

Brief description of amendment: The amendment revised the Kewaunee licensing basis, approving the licensee to operate the load tap changers (LTCs) on two new transformers in the automatic mode. The LTCs are designed to compensate for potential offsite power voltage variations and will provide added assurance that acceptable voltage is maintained for safety-related equipment.

Date of issuance: July 29, 2011.

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

Amendment No.: 209.

Renewed Facility Operating License No. DPR-43: Amendment did not revise the Technical Specifications.

Date of initial notice in Federal Register: August 10, 2010 (75 FR 48374).

The supplements dated January 18, 2011, March 14, 2011, and June 27, 2011, provided additional information that clarified the application, did not expand the scope of the application, and Start Printed Page 52708did not change the Commission's proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut

Date of amendment request: July 21, 2010.

Description of amendment request: The amendment relocates Millstone Power Station, Unit No. 3 (MPS3) Technical Specification (TS) 3/4.7.14, “Area Temperature Monitoring,” and the associated Table 3.7-6, “Area Temperature Monitoring,” to the MPS3 Technical Requirements Manual.

Date of issuance: July 27, 2011.

Effective date: As of its date of issuance, and shall be implemented within 60 days.

Amendment No.: 250.

Renewed Facility Operating License No. NPF-49: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: March 22, 2011 (76 FR 16007).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 27, 2011.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: July 14, 2010.

Brief description of amendments: The amendments revised the Technical Specifications related to the adoption of technical specification task force technical change Traveler 52, Revision 3, to implement option B of Appendix J to Title 10 of the Code of Federal Regulations (10 CFR), part 50.

Date of Issuance: July 28, 2011.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: 375, 377, and 376.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: December 14, 2010 (75 FR 77909).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: June 10, 2009, as supplemented by letters dated December 18, 2009, and August 25, 2010.

Brief description of amendments: The amendments change the Technical Specifications (TSs) and authorize changes to the “Updated Final Safety Analysis Report” (UFSAR) to allow the use of CASMO-4/SIMULATE-3 methodology for application to reactor core designs containing low enrichment uranium fuel bearing lumped burnable and/or gadolinia integral absorbers.

Date of Issuance: August 2, 2011.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 377, 379, and 378.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and the TSs and authorized UFSAR changes.

Date of initial notice in Federal Register: March 19, 2010 (75 FR 13314).

The supplements dated December 15, 2009, and August 25, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 2, 2011.

No significant hazards consideration comments received: No.

Duke Power Company, LLC, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Duke Power Company, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Duke Power Company, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: July 28, 2010, as supplemented March 3, 2011.

Brief description of amendments: The amendments approve changes to each station emergency plans to allow changes to the minimum staffing requirement during emergencies.

Date of Issuance: July 29, 2011.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: Catawba 1 and 2-265/261.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and emergency plan.

Amendment Nos.: McGuire 1 and 2—263/243

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and emergency plan.

Amendment Nos. Oconee 1, 2 and 3—376/378/377

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and emergency plan.

Date of initial notice in Federal Register: September 7, 2010 (75 FR 54393).

The supplement dated March 3, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 29, 2011.

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, West Feliciana Parish, Louisiana

Date of amendment request: July 22, 2010, as supplemented by letters dated September 23 and November 30, 2010, and February 15 and April 4, 2011.

Brief description of amendment: The amendment approved the cyber security plan (CSP) and associated implementation schedule, and added new Paragraph 2.E to Facility Operating License No. NPF-47 to provide a license condition to require the licensee to fully implement and maintain in effect all provisions of the NRC-approved Cyber Security Plan. The proposed change is generally consistent with Nuclear Energy Institute (NEI) 08-09, Revision 6, “Cyber Security Plan for Nuclear Power Reactors.”

Date of issuance: July 29, 2011.

Effective date: This license amendment is effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on Start Printed Page 52709April 4, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment No.: 171.

Facility Operating License No. NPF-47: The amendment revised the Facility Operating License.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62596).

The supplemental letters dated September 23 and November 30, 2010, and February 15 and April 4, 2011, 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 July 29, 2011.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-286, Indian Point Nuclear Generating Unit 1, 2 and 3, (IP1, IP2, and IP3), Westchester County, New York

Date of application for amendment: July 8, 2010, as supplemented by letters dated February 18, April 1, and June 29, 2011.

Brief description of amendment: The licensee's application for the proposed amendments to the Facility Operating Licenses (FOLs) includes: (1) The proposed Cyber Security Plan (CSP), (2) an implementation schedule, and (3) a proposed statement to be added to the existing FOL Physical Protection license conditions requiring Entergy to fully implement and maintain in effect all provisions of the Commission-approved CSP as required by 10 CFR 73.54, “Protection of digital computer and communication systems and networks.” A Federal Register notice dated March 27, 2009, issued the final rule that amended 10 CFR Part 73. The regulations in 10 CFR 73.54, establish the requirements for a CSP. This regulation specifically requires each licensee currently licensed to operate a nuclear power plant under Part 50 of this chapter to submit a CSP that satisfies the requirements of the Rule. Each submittal must include a proposed implementation schedule, and implementation of the licensee's CSP must be consistent with the approved schedule. The background for this application is addressed by the NRC Notice of Availability, Federal Register Notice, Final Rule 10 CFR Part 73, Power Reactor Security Requirements, published on March 27, 2009 (74 FR 13926).

Date of issuance: August 2, 2011.

Effective date: These license amendments are effective as of the date of their issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on July 8, 2010, as supplemented by letters dated February 18, April 1, and June 29, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 55 for IP1, 266 for IP2, and 243 for IP3, respectively.

Facility Operating License Nos. DPR-5, DPR-26, and DPR-64: The amendment revised the Licenses.

Date of initial notice in Federal Register : October 12, 2010 (75 FR 62596).

The supplements dated February 18, April 1, and June 29, 2011, 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 amendment is contained in a Safety Evaluation dated August 2, 2011.

No significant hazards consideration comments received: Yes. The Safety Evaluation dated August 2, 2011, provides the discussion of the comments received from New York State.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant, Van Buren County, Michigan

Date of application for amendment: July 26, 2011, supplemented by letters dated September 27, 2010, November 30, 2010, February 15, 2011, and April 4, 2011.

Brief description of amendment: The amendment approves the cyber security plan and associated implementation schedule, and revises Paragraph 2.E of Facility Operating License No. DPR-20 for Palisades Nuclear Plant, to provide a license condition to require the licensee to fully implement and maintain in effect all provisions of the NRC-approved Cyber Security Plan. The proposed change is generally consistent with Nuclear Energy Institute (NEI) 08-09, Revision 6, Cyber Security Plan for Nuclear Power Reactors.

Date of issuance: July 27, 2011.

Effective date: This license amendment is effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on July 26, 2010, as supplemented by letters dated September 27, 2010, November 30, 2010, February 15, 2011, and April 4, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment No.: 243.

Facility Operating License No. DPR-20: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register : December 7, 2010 (75 FR 76044).

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 27, 2011.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: August 10, 2010, as supplemented by letter dated June 10, 2011.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.9.3, “Reactor Building Penetrations,” to allow reactor building flow path(s) providing direct access from the reactor building atmosphere to the outside atmosphere to be unisolated under administrative control, during movement of irradiated fuel assemblies. The proposed change is consistent with Technical Specification Task Force (TSTF) Technical Change Traveler TSTF-312, Revision 1, “Administratively Control Containment Penetrations.”

Date of issuance: August 10, 2011.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: 245.

Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications/license.

Date of initial notice in Federal Register : October 5, 2010 (75 FR 61526).Start Printed Page 52710

The supplemental letter dated June 10, 2011, 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 August 10, 2011.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendment: March 19, 2010, as supplemented by letters dated June 16, 2010, October 29, 2010, December 3, 2010, January 14, 2011, and March 23, 2011.

Brief description of amendment: The changes implement an extension of the Technical Specification (TS) allowed outage time (AOT) for the Unit 1 and Unit 2 Suppression Pool Cooling (SPC) mode of the Residual Heat Removal (RHR) system, the Residual Heat Removal Service Water (RHRSW) system, the Emergency Service Water (ESW) system, and the A.C. Sources—Operating (Emergency Diesel Generators) from 72 hours to seven (7) days in order to allow for repairs of the RHRSW system piping. The AOT extension would only be allowed once every other calendar year, for each unit, with the opposite unit shutdown, reactor vessel head removed, reactor cavity flooded, and certain other specific compensatory measures, in effect.

Date of issuance: July 29, 2011.

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

Amendment Nos.: 203 and 165.

Facility Operating License Nos. NPF-39 and NPF-85: These amendments revised the license and the technical specifications.

Date of initial notice in Federal Register : May 18, 2010 (75 FR 27828).

The supplements dated June 16, 2010, October 29, 2010, December 3, 2010, January 14, 2011, and March 23, 2011, 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 determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, Limerick Generating Station, Unit 1 and 2, Montgomery County, Pennsylvania

Exelon Generation Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of application for amendments: November 23, 2009, as supplemented by letters dated July 23, September 24, November 18, December 21, 2010, March 31, May 19, and July 11, 2011.

Brief description of amendments: The amendments were submitted in accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.4 and 10 CFR 50.90 and requests NRC approval of the Exelon Generation Company, LLC (Exelon) Cyber Security Plan (CSP), provides an Implementation Schedule, and adds a sentence to the existing Physical Protection license condition to require Exelon to fully implement and maintain in effect all provisions of the Commission approved CSP.

Date of issuance: August 10, 2011.

Effective date: These license amendments are effective as of the date of their issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on November 23, 2009 as supplemented by letters dated July 23, September 24, November 18, December 21, 2010, March 31, May 19, and July 11, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 168, 168, 175, 175, 194, 238, 231, 203, 190, 204, 166, 280, 281, 283, 249, 244, 275.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, NPF-39, NPF-85, DPR-16, DPR-44, DPR-56, DPR-29, DPR-30, DPR-50: The amendments revised the Licenses.

Date of initial notice in Federal Register : April 12, 2011 (75 FR 20379).

The July 23, September 24, November 18, December 21, 2010, March 31, May 19, and July 11, 2011, supplements contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 10, 2011.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of application for amendments: July 22, 2010, as supplemented by letters dated September 28, 2010, November 29, 2010, February 3, 2011, and April 6, 2011.

Brief description of amendments: The amendments to the Renewed Facility Operating Licenses (FOL) include: (1) The proposed BVPS-1 and 2 Cyber Security Plan (CSP), (2) an implementation schedule, and (3) a proposed sentence to be added to the existing renewed FOL Physical Protection license condition for BVPS-1 and 2 requiring FirstEnergy Nuclear Operating Company to fully implement and maintain in effect all provisions of the Commission-approved BVPS-1 and 2 CSP as required by Title 10 of the Code of Federal Regulations (10 CFR) 73.54, “Protection of digital computer Start Printed Page 52711and communication systems and networks.” A Federal Register notice dated March 27, 2009, issued the final rule that amended 10 CFR Part 73. The regulations in 10 CFR 73.54, establish the requirements for a CSP. This regulation specifically requires each licensee currently licensed to operate a nuclear power plant under part 50 of this chapter to submit a CSP that satisfies the requirements of the Rule. Each submittal must include a proposed implementation schedule and implementation of the licensee's CSP must be consistent with the approved schedule. The background for this application is addressed by the NRC Notice of Availability, Federal Register Notice, Final Rule, 10 CFR Part 73, Power Reactor Security Requirements, published on March 27, 2009 (74 FR 13926).

Date of issuance: July 28, 2011.

Effective date: These license amendments are effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on July 22, 2010, as supplemented by letters dated September 28, 2010, November 29, 2010, February 3, 2011, and April 6, 2011, and approved by the Nuclear Regulatory Commission (NRC) staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 287 for BVPS-1 and 174 for BVPS-2.

Facility Operating License Nos. DPR-66 and NPF-73: The amendments revised the License.

Date of initial notice in Federal Register : October 12, 2010, 75 FR 62599.

The supplements dated September 28, 2010, November 29, 2010, February 3, 2011, and April 6, 2011, 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 amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of application for amendments: July 28, 2010, as supplemented by letters dated September 27 and November 19, 2010, and April 5 and June 30, 2011.

Brief description of amendments: The amendment includes three parts: The proposed plan, an implementation schedule, and a sentence added to the existing Physical Protection license condition to require FPL to fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP) as required by amended Title 10 of the Code of Federal Regulations (10 CFR) part 73. The proposed CSP was submitted in accordance with 10 CFR 73.54, “Protection of digital computer and communication systems and networks.”

Date of issuance: July 29, 2011.

Effective date: These license amendments are effective as of the date of their issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on July 28, 2010, as supplemented by letters dated September 27 and November 19, 2010, and April 5 and June 30, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos: Unit 3—245 and Unit 4—241.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the licenses.

Date of initial notice in Federal Register: December 7, 2010 (75 FR 76045).

The supplements dated September 27 and November 19, 2010, and April 5 and June 30, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendment: July 19, 2010, as supplemented by letters dated September 28, 2010, November 30, 2010, and April 8, 2011.

Brief description of amendment: The amendments approve the Cyber Security Plan and associated implementation schedule, and revises License Condition 2.D of the Renewed Facility Operating Licenses for Units 1 and 2. The amendments specify that the licensee fully implement and maintain in effect all provisions of the Commission approved CSP as required by 10 CFR 73.54.

Date of issuance: July 28, 2011.

Effective date: These license amendments are effective as of the date of issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on April 8, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 315 (for Unit 1) and 299 (for Unit 2).

Facility Operating License No. DPR-74: Amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62600).

The supplemental letters contain clarifying information, did not change the scope of the license 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 July 28, 2011.

No significant hazards consideration comments received: No.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: July 14, 2010, as supplemented by letters dated September 27, 2010, November 17, 2010, April 5, 2011, and June 22, 2011.

Brief description of amendment: The amendment approves the Cyber Security Plan and associated implementation schedule, and revises License Condition Start Printed Page 527122.C.(5) of the Renewed Facility Operating License. The amendment specifies that the licensee fully implement and maintain in effect all provisions of the Commission approved CSP, as required by 10 CFR 73.54.

Date of issuance: July 29, 2011.

Effective date: This license amendment is effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on April 5, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment No.: 278.

Renewed Facility Operating License No. DPR-49: The amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: November 9, 2010 (75 FR 68836).

The supplemental letters contain clarifying information, did not change the scope of the license 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 July 29, 2011.

No significant hazards consideration comments received: No.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of application for amendments: January 27, 2010, as supplemented by letters dated August 30, 2010, and May 3, 2011.

Brief description of amendments: The amendments revise Technical Specification 3.8.3, “Diesel Fuel Oil and Starting Air,” to specify an increased minimum diesel fuel oil storage volume and associated surveillance requirement for the Emergency Diesel Generators.

Date of issuance: August 4, 2011.

Effective date: This license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance.

Amendment Nos.: 244 (for Unit 1) and 248 (for Unit 2).

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications and Renewed Facility Operating License.

Date of initial notice in Federal Register: November 30, 2010 (75 FR 74096).

The August 30, 2010, and May 3, 2011, supplements did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 4, 2011.

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: July 20, 2010, and supplemented by letters dated September 24, 2010, November 30, 2010, February 21, 2010, April 1, 2011, and May 26, 2011.

Brief description of amendment: The amendment approves the Cyber Security Plan (CSP) and associated implementation schedule, and revises License Condition 2.C.3 of the Renewed Facility Operating License DPR-22 for Monticello Nuclear Generating Plant. The amendment specifies that the licensee fully implement and maintain in effect all provisions of the Commission approved CSP as required by 10 CFR 73.54.

Date of issuance: August 2, 2011.

Effective date: This license amendment is effective as of the date of its issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on April 1, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment No.: 166.

Facility Operating License No. DPR-22. Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62604).

The licensee's supplemental letters contained clarifying information, did not change the scope of the original license 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 August 2, 2011.

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of application for amendment: July 20, 2010, and supplemented by letters dated September 24, 2010, November 30, 2010, February 21, 2011, April 1, 2011, and May 26, 2011.

Brief description of amendment: The amendments approve the Cyber Security Plan (CSP) and associated implementation schedule, and revise License Condition 2.C.(3) of the Facility Operating Licenses for each unit at Prairie Island Nuclear Generating Plant. The amendments specify that the licensee fully implement and maintain in effect all provisions of the Commission-approved CSP as required by 10 CFR 73.54.

Date of issuance: July 29, 2011.

Effective date: These license amendments are effective as of the date of their issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on April 1, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 202 (for Unit 1) and 189 (for Unit 2).

Facility Operating License Nos. DPR-42 and DPR-60. Amendments revised the Facility Operating Licenses

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62604).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: July 14, 2010, as supplemented by letters dated September 28, 2010, April 1, 2011, June 6, 2011, and July 6, 2011.

Brief description of amendments: The amendments approve the Cyber Security Plan (CSP) and associated implementation schedule for Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 Start Printed Page 52713and 2. In addition, the amendments revise the existing license condition regarding physical protection in the each of the three facility operating licenses (FOLs) to require the licensee to fully implement and maintain in effect all provisions of the Nuclear Regulatory Commission (NRC)-approved CSP. The amendment was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 73.54, which requires licensees currently licensed to operate a nuclear power plant under 10 CFR part 50 to submit a CSP for NRC review and approval.

Date of issuance: July 28, 2011.

Effective date: The license amendments are effective as of the date of issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee by letter dated June 6, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: 189, 300 and 283.

Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The amendments revised the FOLs.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62606).

The letters dated September 28, 2010, April 1, 2011, June 6, 2011, and July 6, 2011, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of application for amendments: July 22, 2010, as supplemented by letters dated September 29 and November 30, 2010, and March 31 and June 16, 2011.

Brief description of amendments: The amendments approved the cyber security plan (CSP) and associated implementation schedule, and revised Paragraph 2.E of Facility Operating License Nos. NPF-10 and NPF-15, respectively, for San Onofre Nuclear Generating Station, Units 2 and 3, to provide a license condition to require the licensee to fully implement and maintain in effect all provisions of the NRC-approved Cyber Security Plan. The proposed change is consistent with Nuclear Energy Institute (NEI) 08-09, Revision 6, “Cyber Security Plan for Nuclear Power Reactors.”

Date of issuance: July 28, 2011.

Effective date: These license amendments are effective as of the date of issuance. The implementation of the CSP, including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on March 31 and June 16, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: Unit 2—225; Unit 3—218.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: November 9, 2010 (75 FR 68836).

The supplemental letters dated September 29 and November 30, 2010, and March 31 and June 16, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364, Houston County, Alabama; Edwin I. Hatch Nuclear Plant, Units 1 and 2, Docket Nos. 50-321 and 50-366, Appling County, Georgia; Vogtle Electric Generating Plant, Units 1 and 2, Docket Nos. 50-424 and 50-425, Burke County, Georgia

Date of amendment request: July 16, 2010, as supplemented March 28 and April 11, 2011.

Brief description of amendment request: The amendments approve the licensee's Cyber Security Plan and Implementation Schedule.

Date of issuance: July 28, 2011.

Effective date: These license amendments are effective as of the date of their issuance. The implementation of the cyber security plan (CSP), including key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee by letter dated April 11, 2011, and approved by the NRC staff with these license amendments. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos: Farley 1 and 2—186/181; Hatch 1 and 2—265/209; Vogtle 1 and 2—162/144.

Facility Operating License (Farley) NPF-2 and NPF-8; (Hatch) DPR-57 and NPF-5; (Vogtle) NPF-68 and NPF-81: The amendments changed the licenses and the technical specifications.

Date of initial notice in Federal Register: April 12, 2011 (76 FR 20381)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama

Date of application for amendments: November 23, 2009, as supplemented on December 18, 2009; July 23 and October 1, 2010; April 7 and July 15, 2011 (TS-470).

Description of amendment request: On March 27, 2009, the Federal Register Notice (74 FR 13926) published the final rule that amended Title 10 of the Code of Federal Regulations (10 CFR) Part 73, “Physical Protection of Plants and Materials.” Specifically, the regulations in 10 CFR 73.54 “Protection of Digital Computer and Communication Systems and Networks,” establish the requirements for a cyber security program to protect digital computer and communication systems and networks against cyber attacks. The proposed amendment included the proposed Cyber Security Plan, its implementation schedule, and a revised Physical Protection license condition for Browns Ferry Nuclear Plant, Units 1, 2, and 3 to fully implement and maintain in effect all provisions of the Nuclear Regulatory Commission approved Cyber Security Plan as required by 10 CFR 73.54.

Date of issuance: July 29, 2011.

Effective date: This license amendment is effective as of the date of issuance. The implementation of the cyber security plan (CSP), including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on Start Printed Page 52714April 7, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: Unit 1—279, Unit 2—306, and Unit 3—265.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the licenses.

Date of initial notice in Federal Register: December 7, 2010 (75 FR 76046).

The above Federal Register notice was based on the supplement dated December 18, 2009. The supplements dated July 23 and October 1, 2010; April 7 and July 15, 2011, 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 July 29, 2011.

No significant hazards consideration comments received: No.

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: November 23, 2009, as supplemented on December 11 and December 18, 2009; July 23 and October 1, 2010; April 7 and July 15, 2011 (TS 09-06).

Brief description of amendment: On March 27, 2009, the Federal Register Notice (74 FR 13926) published the final rule that amended Title 10 of the Code of Federal Regulations (10 CFR) part 73, “Physical Protection of Plants and Materials.” Specifically, the regulations in 10 CFR 73.54 “Protection of Digital Computer and Communication Systems and Networks,” establish the requirements for a cyber security program to protect digital computer and communication systems and networks against cyber attacks. The proposed amendment included the proposed Cyber Security Plan, its implementation schedule, and a revised physical protection license condition for Sequoyah Nuclear Plant, Units 1 and 2 to fully implement and maintain in effect all provisions of the Nuclear Regulatory Commission approved Cyber Security Plan as required by 10 CFR 73.54.

Date of issuance: July 29, 2011.

Effective date: This license amendment is effective as of the date of issuance. The implementation of the Cyber Security Plan (CSP), including the key intermediate milestone dates and the full implementation date, shall be in accordance with the implementation schedule submitted by the licensee on April 7, 2011, and approved by the NRC staff with this license amendment. All subsequent changes to the NRC-approved CSP implementation schedule will require prior NRC approval pursuant to 10 CFR 50.90.

Amendment Nos.: Unit 1—329 and Unit 2—322.

Facility Operating License DPR-77 and DPR-79: Amendments revised the licenses.

Date of initial notice in Federal Register: December 7, 2010 (75 FR 76046).

The above Federal Register notice was based on the supplement dated December 18, 2009. The supplements dated July 23 and October 1, 2010; April 7 and July 15, 2011, 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 July 29, 2011.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

Date of application for amendment: November 25, 2009, as supplemented by letters dated April 22, May 14, August 24, September 29, and November 4, 2010, and February 23, 2011.

Brief description of amendment: The amendment revised Technical Specification 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” to provide a 24-hour Completion Time (CT) for restoration of an inoperable Balance of Plant (BOP) ESFAS train and extends the CTs associated with individual instrument channels in the BOP ESFAS train to maintain overall consistency of related TS actions.

Date of issuance: July 28, 2011.

Effective date: As of its date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: 201.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: May 18, 2010 (75 FR 27833).

The supplemental letters dated April 22, May 14, August 24, September 29, and November 4, 2010, and February 23, 2011, 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 July 28, 2011.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

Date of application for amendment: August 5, 2010, as supplemented by letters dated March 23, May 3, and July 25, 2011.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the guidance of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.” The amendment adopted NRC-approved Technical Specification Task Force (TSTF)-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk-Informed TSTF] Initiative 5b.” When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs.

Date of issuance: July 29, 2011.

Effective date: As of its date of issuance and shall be implemented within 180 days from the date of issuance.

Amendment No.: 202.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 11, 2011 (76 FR 1649).

The supplemental letters dated March 23, May 3, and July 25, 2011, 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.Start Printed Page 52715

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 29, 2011.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of application for amendments: July 12, 2010.

Brief Description of amendments: These amendments revise the Technical Specifications (TSs) to: (1) Correct an error in TS 3.12.E.5, (2) delete duplicative requirements in TS 3.12.E.2 and TS 3.12.E.4, (3) relocate the shutdown margin value in TS 3.12 and the TS 3.12 Basis to the Core Operating Limits Report (COLR), and 4) expand the TS 6.2 list of parameters defined in the COLR.

Date of issuance: July 28, 2011.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment Nos.: 275 and 275.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the licenses and the technical specifications.

Date of initial notice in Federal Register: May 17, 2011 (76 FR 28477).

The Commission's related evaluation of the amendments is contained in a safety evaluation dated July 28, 2011.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 11th day of August 2011.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

End Signature End Preamble

[FR Doc. 2011-21212 Filed 8-22-11; 8:45 am]

BILLING CODE 7590-01-P