2011-22541. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

Background

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from August 11, 2011 to August 24, 2011. The last biweekly notice was published on August 23, 2011 (76 FR 52699).

ADDRESSES:

Please include Docket ID NRC-2011-0205 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-2011-0205. Address questions about NRC dockets to Carol Gallagher 301-492-3668; e-mail [e-mail].
• 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 [e-mail].
• 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-2011-0205.

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, then 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 [e-mail], or by telephone at 301-415-1677, to request (1) A digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on 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 [e-mail], 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 [e-mail].

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

Date of amendment request: July 12, 2011.

Description of amendment request: The proposed license amendments would revise Technical Specification (TS) 3.4.5, “Reactor Coolant System (RCS) Leakage Detection Instrumentation,” to define a new time limit for restoring inoperable RCS leakage detection instrumentation to operable status and establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable. These proposed changes would be consistent with Standard Technical Specifications Change Traveler (TSTF)-514, “Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation.” The availability of TSTF-514 was announced in the Federal Register on December 17, 2010 (75 FR 79048), as part of the consolidated line item improvement process.

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 clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the primary containment atmosphere gaseous radioactivity monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated.

Therefore, it is concluded that 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 proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS required operable RCS leakage detection instrumentation monitor is the primary containment atmosphere gaseous radioactivity monitor. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

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

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

Response: No

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the primary containment atmosphere gaseous radioactivity monitor. Reducing the amount of time the plant is allowed to operate with only the primary containment atmosphere gaseous radioactivity monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure.

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

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: April 29, 2011.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.4.15, “RCS [Reactor Coolant System] Leakage Detection Instrumentation,” to define a new time limit for restoring inoperable RCS leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable; and make TS Bases changes which reflect the proposed changes and more accurately reflect the contents of the facility design basis related to operability of the RCS leakage detection instrumentation. New Condition C is applicable when the reactor building atmosphere gaseous radioactivity monitor is the only operable TS-required monitor. New Condition C Required Actions require analyzing grab samples of the reactor building atmosphere every 12 hours and restoring another monitor within 7 days. These changes are consistent with NRC-approved Revision 3 to Technical Specification Task Force (TSTF) Standard Technical Specification (STS) Change Traveler TSTF-513, “Revise PWR [Pressurized-Water Reactor] Operability Requirements and Actions for RCS Leakage Instrumentation.” The availability of this TS improvement was announced in the Federal Register on January 3, 2011 (76 FRN 189), as part of the consolidated line item improvement process (CLIIP).

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 clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the reactor building atmosphere gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated.

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

2. Does the Proposed Change Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated?

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the reactor building atmosphere gaseous radiation monitor. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change maintains sufficient continuity and diversity of leak detection capability that the probability of piping evaluated and approved for Leak-Before-Break progressing to pipe rupture remains extremely low.

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

3. Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the reactor building atmosphere gaseous radiation monitor. Reducing the amount of time the plant is allowed to operate with only the reactor building atmosphere gaseous radiation monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: April 6, 2011.

Description of amendment request: The proposed amendment would modify the actions to be taken when the atmospheric gaseous radioactivity monitor is the only operable reactor coolant leakage detection instrument. The modified actions require additional, more frequent monitoring of other indications of Reactor Coolant System (RCS) leakage and provide appropriate time to restore another leakage detection instrument to operable status. This change is consistent with the U.S. Nuclear Regulatory Commission (NRC) approved safety evaluation on Technical Specification Task Force (TSTF) Traveler, TSTF-514-A, Revision 3, “Revised BWR [boiling-water reactor] Operability Requirements and Actions for RCS Leakage Instrumentation” dated November 24, 2010.

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, with NRC edits in brackets:

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

Response: No

The proposed changes [ ] modify the time allowed for the plant to operate when the only Operable RCS leakage detection instrumentation monitor is the atmospheric gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed changes [ ] modify the time allowed for the plant to operate when the only Operable RCS leakage detection monitor is the atmospheric gaseous radiation monitor. The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

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

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

Response: No.

The proposed changes [ ] increase the time allowed for the plant to operate when the only Operable RCS leakage detection instrumentation monitor is the atmospheric gaseous radiation monitor from 24 hours to 7 days. Increasing the amount of time the plant is allowed to operate with only the atmospheric gaseous radiation monitor Operable does not significantly decrease the margin of safety due to the addition of compensatory Required Actions to analyze grab samples of the primary containment atmosphere once per 12 hours and monitor Reactor Coolant System leakage by administrative means once per 12 hours. The overall likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure is maintained with the addition of the Required Actions.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, including the edits in brackets above, 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: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Richard B. Ennis, Acting.

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

Date of amendment request: June 2, 2011.

Description of amendment request: The proposed amendment would modify Technical Specification Limiting Condition for Operation 3.1.2, “Reactor Anomalies,” to allow performance of the surveillance on a comparison of predicted to actual (or monitored) effective core reactivity (K_eff). The reactivity anomaly verification is currently determined by a comparison of predicted vs. actual control rod density.

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 with changes by the NRC staff noted in brackets:

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

Response: No.

The proposed Technical Specifications changes do not [substantively] affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the reactivity anomaly surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the Technical Specification requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of k_eff. Present day on-line core monitoring systems, such as the one in use at Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 are capable of performing the direct measurement of reactivity.

Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident.

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

Response: No.

This Technical Specifications amendment request does not [substantively change] the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the reactivity anomaly Technical Specifications will only provide a new, more efficient method of detecting an unexpected change in core reactivity.

Since all systems continue to be operated within their design bases, no new failure modes are introduced and the possibility of a new or different kind of accident is not created.

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

Response: No.

This proposed Technical Specifications amendment proposes to change the method for performing the reactivity anomaly surveillance from a comparison of predicted to actual control rod density to a comparison of predicted to actual k_eff. The direct comparison of k_eff provides a technically superior method of calculating any differences in the expected core reactivity. The reactivity anomaly surveillance will continue to be performed at the same frequency as is currently required by the Technical Specifications, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, including the changes made by the NRC staff as noted in brackets, 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: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: G. Edward Miller, Acting.

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

Date of amendment request: December 20, 2010, as supplemented by the July 20, 2011 letter.

Description of amendments request: FPC will be constructing and operating an on-site independent spent fuel storage installation at CR-3, as a general licensee under the provisions of 10 CFR part 72, Subpart K to maintain full-core offload capacity in the spent fuel pools. The spent fuel pools are located in the CR-3 Auxiliary Building (AB). In support of future dry shielded canister/transfer cask loading operations, FPC is replacing the existing AB overhead crane with a new single failure proof crane designed in accordance with American Society of Mechanical Engineers (ASME) NOG-1-2004, “Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).” The licensee requested NRC approval of the following:

1. An exception to ASME NOG-1-2004 pertaining to the application of tornado wind and tornado generated missile loading to auxiliary building overhead crane (FHCR-5) and its support structure.

2. Revisions to the CR-3 Final Safety Analysis Report (FSAR) Sections 5.1.1.1.h and 9.6.1.5.a.5 to specifically identify the design parameters for FHCR-5 and its support structure.

3. Deletion of a commitment in FSAR Section 9.6.3.1, “Spent Fuel Assembly Removal,” due to the expansion of spent fuel storage over that originally credited in the CR-3 Safety Evaluation Report dated July 5, 1974.

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 not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed LAR [license amendment request] does not involve plant equipment used to operate or shut down the reactor or in the mitigation of accidents described in Chapter 14 of the FSAR. FHCR-5 will be restricted from movement over fuel stored in either of the spent fuel pools by administrative controls and designated safe load paths when moving spent fuel casks, and it will be single failure proof so a cask load drop accident affecting stored spent fuel is prevented. The change provides justification for an exception to a Code requirement pertaining to the design and qualification of the new single failure proof crane in the AB. The new crane will meet the design specifications in ASME NOG-1-2004, with the exception of Section 4134(c). The change also includes a commitment not to operate the crane if an Approaching or Potential Tropical Storm, an Approaching or Potential Hurricane, or a Tornado Watch or Warning has been declared for the site. The revised FSAR description of the crane will meet the intent of the original description and will ensure the crane will exceed the design requirements of the original design. With the replacement of the crane, the occurrence of a cask load drop accident is not considered credible. As a result, the proposed change does not increase the probability or consequences of a load drop accident previously evaluated that could impact stored fuel and/or pool structural integrity.

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

2. Does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The power generation portion of the plant is unaffected by the proposed change, which is limited to the design and analysis of a new overhead crane in the AB. The location and design functions of the AB overhead crane remain as they are currently described in the CR-3 FSAR. Overall, the design of the crane is being enhanced to single failure proof in order to reduce the likelihood of an uncontrolled lowering of the load due to an unforeseen malfunction or subcomponent failure. Portions of the design and analysis of the crane require NRC approval because they deviate from the NRC-endorsed design code for single failure proof cranes and the CR-3 licensing basis. The new single failure proof crane will be used to move a loaded or unloaded transfer cask between the cask loading pit, the decontamination pit, and the transfer trailer in the truck bay. Any credible event involving the fuel handling evolutions are bounded by existing FSAR analyses.

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

3. Does not involve a significant reduction in a margin of safety.

This proposed LAR involves the replacement of the existing non-single failure proof AB overhead crane with a new single failure proof crane. The new crane will meet the design specifications found in ASME NOG-1-2004, with the exception of Section 4134(c). ASME NOG-1-2004 has been endorsed by the NRC in Regulatory Issue Summary (RIS) 2005-25, Supplement 1, “Clarification of NRC Guidelines for Control of Heavy Loads,” as an acceptable means of meeting the criteria in NUREG-0554, “Single Failure Proof Cranes for Nuclear Power Plants.” The ASME NOG-1-2004 design code has been found by the NRC to provide adequate protection and safety margin against the uncontrolled lowering of the lifted load. The occurrence of a cask load drop accident is considered not credible when the load is lifted with a single failure proof lifting system meeting the guidance in NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants,” Section 5.1.6, “Single Failure Proof Handling Systems.” As a result, the proposed change has no adverse impact on new fuel, stored spent fuel, cooling capacity of the pool, or structural integrity of the pool. Similarly, the margin of safety for the operation and safe shutdown of the plant will not be affected by the proposed change.

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

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Douglas A. Broaddus.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 1, Rockingham County, New Hampshire

Date of amendment request: July 14, 2011.

Description of amendment request: The proposed change would replace the Technical Specification (TS) required 10-year surveillance frequency for testing the containment spray nozzles in accordance with TS surveillance 4.6.2.1.d with an event-based frequency. Specifically, verification that the spray nozzle is unobstructed would only be required following activities that could result in nozzle blockage.

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

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

The spray nozzles and the associated containment spray system (CBS) are designed to perform accident mitigation functions. The proposed change to reduce the frequency and remove specific details of surveillance testing that verifies the spray nozzles are unobstructed does not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions. The proposed change does not alter or prevent the ability of operable SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. The capability of the CBS system to perform its accident mitigation functions is not adversely affected by the proposed change.

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

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

The proposed change will not impact the accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a significant change in the method of plant operation, or new operator actions. The change does not make any physical modifications to the CBS system, changes to setpoints, or changes to the method of delivering borated water to the CBS spray nozzles. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis.

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

3. The proposed change does not involve a significant reduction in the margin of safety.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change does not involve a significant change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: M.S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Harold K. Chernoff.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP), Burke County, Georgia

Date of amendment request: July 26, 2011.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TSs). Specifically, the proposed change would revise the minimum indicated nitrogen cover pressure specified for the accumulators in TS Surveillance Requirement (SR) 3.5.1.3 from 617 psig (pounds per square inch, gauge) to 626 psig. The proposed change is necessary to account for the uncertainty associated with the accumulator pressure indication instrumentation. Currently, in accordance with NRC Administrative Letter 98-10, “Dispositioning of Technical Specifications that Are Insufficient to Assure Plant Safety,” VEGP is administratively controlling the minimum indicated accumulator pressure to greater than or equal to 626 psig. In addition, an editorial error in the text of TS SR 3.6.2.1 would also be corrected.

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

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

Response: No

The proposed amendment revises the minimum indicated nitrogen cover pressure specified for the SI [safety injection] accumulators in SR 3.5.1.3 from 617 psig to 626 psig. In addition, the proposed change includes an administrative change to correct an editorial error in the text of TS SR 3.6.2.1.

The SI accumulators are not a precursor to any accident previously evaluated. The SI accumulators are used to mitigate the consequences of accidents previously evaluated. The proposed change to the indicated minimum SI accumulator nitrogen cover pressure provides assurance that the requirements of the TS continue to bound the acceptance limits of the SI accumulators with respect to the assumptions in the LOCA [loss-of-coolant accident] analyses.

Thus, the proposed change does not affect the probability or the consequences of any accident previously evaluated. The proposed change to correct an editorial error in the text of SR 3.6.2.1 has no impact on the probability or consequences of any accident previously evaluated.

Therefore, it is concluded that 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 proposed change revises the minimum indicated nitrogen cover pressure specified for the SI accumulators in SR 3.5.1.3 from 617 psig to 626 psig. In addition, the proposed change includes an administrative change to correct an editorial error in the text of TS SR 3.6.2.1.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change to the requirements of the TS assure that the acceptance limits of the SI accumulators with respect to the assumptions in the LOCA analyses continue to be met, and correct an editorial error in the text of an SR. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety.

Therefore, it is concluded that 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

The proposed change revises the minimum indicated nitrogen cover pressure specified for the SI accumulators in SR 3.5.1.3 from 617 psig to 626 psig. In addition, the proposed change includes an administrative change to correct an editorial error in the text of TS SR 3.6.2.1.

The proposed change to the indicated SI accumulator nitrogen cover pressure provides assurance that the requirements of the TS continue to bound the acceptance limits of the SI accumulators with respect to the assumptions in the LOCA analyses. Thus the proposed change to the SI accumulator minimum nitrogen cover pressure assures the existing margin of safety is maintained. The proposed change to correct an editorial error in the text of SR 3.6.2.1 has no impact on the margin of safety.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216. NRC Branch Chief: Gloria Kulesa.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, Unit 2, Hamilton County, Tennessee

Date of amendment request: July 15, 2011 (TS-SQN-2011-01).

Description of amendment request: The proposed amendment would revise the technical specifications (TSs) requirements for steam generator (SG) tube inspections to reflect the replacement steam generators (RSGs) to be installed during Sequoyah Nuclear Plant (SQN), Unit 2, refueling outage 18 presently scheduled for the fall of 2012. Previous changes to the SQN, Unit 2, TSs to reflect the Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4, were approved by Nuclear Regulatory Commission (NRC) on May 22, 2007. The changes proposed in this amendment reflect the inspection requirements of TSTF-449, Revision 4. The RSG tubes will be made of Alloy 690 thermally treated (TT) material, and the existing SGs have Alloy 600 tubes. The revisions to TSs are required because the inspection frequency for Alloy 690 TT tube material, as defined in TSTF-449, differs from the inspection frequency for Alloy 600, and the tube repair processes and products in the existing TSs are not applicable to the RSGs.

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 for RSGs continues to implement the current SG Program that includes performance criteria which provide reasonable assurance that the RSG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specifications). This change removes repair criteria from the SG Program that were approved by previous License Amendments for the existing SGs which are not applicable to the RSGs. It removes references to use of repairs and reporting of repair results in other TS sections. This change removes inspection requirements that are designated for specific damage conditions in the existing SGs. The change also revises the inspection interval for 100 percent inspections of SG tubes and the maximum interval for inspection of a single SG consistent with Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4 for the Alloy 690 tube material in the RSGs. The revised inspection requirements are based on properties and experience with the improved Alloy 690 tube material. The revised inspection requirements will result in the same outcome that SG tube integrity will continue to be maintained.

This change continues to implement SG performance criteria for tube structural integrity, accident induced leakage, and operational leakage for the RSGs. Meeting the performance criteria provides reasonable assurance that the RSG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident (DBA). The performance criteria are only a part of the SG Program required by the existing TS. The program, defined by NEI [Nuclear Energy Institute] 97-06, “Steam Generator Program Guidelines,” includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. These features will continue to be implemented as they are currently approved. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of DBAs are, in part, functions of the Dose Equivalent 1-131 in the primary coolant and the primary to secondary leakage rates resulting from an accident. Therefore, limits are included in the TS for Operational Leakage and for Dose Equivalent 1-131 in the primary coolant to ensure the plant is operated within its analyzed condition. The analysis of the limiting DBA assumes that the primary to secondary leak rate, after the accident, is 1 gallon per minute with no more than 150 gallons per day in any one SG, and that the reactor coolant activity levels of Dose Equivalent 1-131 are at the TS values before the accident. The proposed change to the SG inspection program does not affect the design of the SGs, their method of operation, operational leakage limits, or primary coolant chemistry controls. The proposed change does not adversely impact any other previously evaluated DBA. In addition, the proposed changes do not affect the consequences of a main steam line break, rod ejection, a reactor coolant pump locked rotor event, or other previously evaluated accident.

Therefore, the proposed change does not affect the consequences of a[n] SG tube rupture accident and the probability of such an accident is unchanged.

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 license amendment does not affect the method of operation of the SGs, or the primary or secondary coolant chemistry controls. In addition, the proposed amendment does not impact any other plant system or component. The change modifies existing SG inspection requirements based on the RSG design and the properties and experience associated with their improved materials. The revised inspection requirements will result in the same outcome that SG tube integrity will continue to be maintained.

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

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

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a[n] SG is maintained by ensuring the integrity of its tubes. SG tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change to the SG inspection program does not affect tube design or operating environment. The existing SG Program is maintained in this change. The repair criteria that are being removed are specific to the existing SGs and are not applicable to the RSGs. If tube defects are detected that exceed limits in the RSGs, then the tube will be removed from service. The effective tube plugging percentage will continue to be tracked for all plugging in each SG in accordance with TS Section 6.9.1.16.1 to ensure the heat transfer function of the SGs is not adversely affected. For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the TS.

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.

Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket No. 50-278, Peach Bottom Atomic Power Station (PBAPS), Unit 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: June 28, 2011.

Brief description of amendment request: The proposed amendment would modify the PBAPS, Unit 3, Technical Specification Section 2.1.1 to revise Safety Limit Minimum Critical Power Ratio values.

Date of publication of individual notice in Federal Register: August 22, 2011 (76 FR 52357).

Expiration date of individual notice: September 21, 2011 (public comments) and October 21, 2011 (hearing requests).

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's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to [e-mail].

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit 1 and 2, Calvert County, Maryland

Date of application for amendments: March 22, 2011.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) to define a new time limit for restoring inoperable reactor coolant system (RCS) leakage detection instrumentation to operable status. The proposed TS changes are consistent with TS Task Force (TSTF)-513, “Revise PWR [pressurized-water reactor] Operability Requirements and Actions for RCS Leakage Instrumentation.”

Date of issuance: August 24, 2011.

Effective date: As of the date of issuance to be implemented within 90 days.

Amendment Nos.: 299 and 276.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: April 19, 2011 (76 FR 21920).

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated August 24, 2011.

No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, LLC,

Calvert Cliffs Nuclear Power Plant, Unit 1 and 2 (CCNPP),

Docket Nos. 50-317, 50-318,

Calvert County, Maryland,

Nine Mile Point Nuclear Station, LLC,

Nine Mile Point Nuclear Station, Unit 1 and 2 (NMPNS),

Docket Nos. 50-220, 50-410,

Oswego County, New York, and

R. E. Ginna Nuclear Power Plant, LLC,

R. E. Ginna Nuclear Power Plant (Ginna),

Docket No. 50-244, Wayne County, New York

Date of amendment request: July 16, 2010, as supplemented by letters dated April 4, and July 1, 2011.

Brief description of amendments: The amendments to the Renewed Facility Operating Licenses (FOLs) includes: (1) The U.S. Nuclear Regulatory Commission (NRC)-approved Cyber Security Plan (CSP) for CCNPP, NMPNS, and Ginna, (2) the CSP implementation schedule, and (3) the license condition added to the existing physical protection license condition for CCNPP, NMPNS, and Ginna, requiring the licensee to fully implement and maintain in effect all provisions of the NRC-approved CSP for CCNPP, NMPNS, and Ginna, as required by Title 10 of the Code of Federal Regulations (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 73.54. 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 19, 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 16, 2010, as supplemented by letters dated April 4, and July 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 Nos.: 298, 275 (CCNPP1 & CCNPP2), 209, 137 (NMPNS1 & NMPNS2), and 113 (Ginna),.

Renewed Facility Operating License Nos. DPR-53 and DPR-69 (CCNPP1 & CCNPP2), DPR-63, NPF-69, (NMP1 & NMP2), and DPR-18 (Ginna),: Amendments revised the Licenses.

Date of initial notice in Federal Register: October 12, 2010 (75 FR 62594). The supplement dated April 4, and July 1, 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 these amendments is contained in a Safety Evaluation dated August 19, 2011.

No significant hazards consideration comments received: Yes.

The State of Maryland had no comments. However, the New York State provided comments. The Safety Evaluation dated August 19, 2011, provides the discussion of the comments received from the New York State.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

Date of application for amendment: July 15, 2010, as supplemented by letters dated February 15 and April 4, 2011.

Brief description of amendment: The application for the proposed amendment to the Renewed Facility Operating License (FOL) includes: (1) The proposed JAFNPP Cyber Security Plan, (2) an implementation schedule, and (3) a proposed sentence to be added to the existing renewed FOL Physical Protection license condition for JAFNPP requiring Entergy to fully implement and maintain in effect all provisions of the Commission-approved JAFNPP Cyber Security Plan (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 cyber security program. 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 Cyber Security Program 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 19, 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 15, 2010, as supplemented by letters dated February 15 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.: 300.

Renewed Facility Operating License No. DPR-59: The amendment revised the License

Date of initial notice in Federal Register: August 20, 2010 (75 FR 51492). The supplements dated 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 NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 19, 2011.

No significant hazards consideration comments received: Yes.

The Safety Evaluation dated August 19, 2011, provides the discussion of the comments received from the New York State.

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

Date of application for amendment: September 23, 2010 as supplemented by letter dated. April 22, 2011.

Brief description of amendment: The amendment revised Technical Specification (TS) limiting condition for operation 3.7.6, “Main Turbine Bypass System (MTBS),” to control the reactor operational limits, as specified in the Clinton Power Station Core Operating Limits Report to compensate for the inoperability of the MTBS.

Date of issuance: August 17, 2011.

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

Amendment No.: 195.

Facility Operating License No. NPF-62: The amendment revised the TSs and license.

Date of initial notice in Federal Register: February 1, 2011 (76 FR 5618). The April 22, 2011 supplement contained clarifying information and did not change the NRC staff=s initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 17, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 22, 2010, supplemented by letter dated April 7, 2011.

Brief description of amendment: The changes relocate the list of pumps, fans, and valves in Technical Specification (TS) 4.5.1.1b, Sequence and Power Transfer Test, to the TMI-1 Updated Final Safety Analysis Report. In place of the TS equipment listing there will be a more general reference to the permanently-connected and automatically-connected emergency loads which are tested through the load sequencer. In addition, TS 4.5.1.2b, TS 4.5.2.2a, and TS 4.5.2.2b refer to this test and are revised to reflect the change to TS 4.5.1.1b.

Date of issuance: August 22, 2011.

Effective date: Immediately, and shall be implemented within 30 days.

Amendment No.: 276.

Renewed Facility Operating License No. DPR-50. Amendment revised the license and the technical specifications.

Date of initial notice in Federal Register: November 30, 2010 (75 FR 74095). The supplement dated April 7, 2011, modified the application such that the Federal Register notice was re-issued on May 3, 2011 (76 FR 24928). The revised notice did not change the NRC staff's proposed no significant hazards determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 22, 2011.

No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 1, Rockingham County, New Hampshire

Date of amendment request: December 29, 2010.

Description of amendment request: The proposed change deletes the Seabrook Technical Specification (TS) 3.4.10, “Structural Integrity,” while relocating the requirements of Surveillance Requirement 4.4.10 to TS 6.7.6.m.

Date of issuance: August 22, 2011.

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

Amendment No.: 126.

Facility Operating License No. NPF-86: The amendment revised the TS and the License.

Date of initial notice in Federal Register: May 31, 2011 (76 FR 31375).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 22, 2011.

No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 1, Rockingham County, New Hampshire

Date of amendment request: July 26, 2010, as supplemented by letters dated September 28, 2010, March 31, June 23, and August 4, 2011.

Description of amendment request: This amendment approves the NextEra Seabrook LLC, cyber security plan (CSP) for Seabrook Station, Unit 1. Additionally, the amendment adds a license condition requiring that the licensee fully implement and maintain in effect all provisions of the approved plan.

Date of issuance: August 23, 2011.

Effective date: The license amendment is effective as of its date of issuance. The implementation of the 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 March 31, 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.: 127.

Facility Operating License No. NPF-86: The amendment revised the License.

Date of initial notice in Federal Register: May 10, 2011 (76 FR 27097).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 23, 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 amendments: March 18, 2011, as supplemented by letters dated May 4 and June 2, 2011.

Brief description of amendments: The amendments modified the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program.

Date of issuance: August 16, 2011.

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

Amendment Nos.: 203, 190.

Facility Operating License Nos. DPR-42 and DPR-60: The amendments revised the Operating Licenses for both units.

Date of initial notice in Federal Register: May 10, 2011 (76 FR 27098). 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 Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 16, 2011.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

Date of application for amendment: August 5, as supplemented September 27, and November 30, 2010 and March 28, 2011.

Brief description of amendment: The amendments revised Paragraph 2.E of the renewed facility operating license 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 and associated implementation schedule.

Date of issuance: August 24, 2011.

Effective date: This license amendment is effective as of its 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 28, 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 No.: 184 .

Renewed Facility Operating License No. NPF-12: Amendment revised the license.

Date of initial notice in Federal Register: April 12, 2011 (76 FR 20380). The September 27, 2010, and March 28, 2011, supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change 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 24, 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 12, 2010, as supplemented by letters dated September 27, November 29, and December 30, 2010, and April 1, June 14, and June 29, 2011.

Brief description of amendment: The amendment approved the Callaway Plant, Unit 1, Cyber Security Plan and associated implementation schedule, and revised Paragraph 2.E of Facility Operating License No. NPF-30 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: August 17, 2011.

Effective date: This license amendment is effective as of the date of its 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 revised implementation schedule submitted by the licensee on June 29, 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.: 203.

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

Date of initial notice in Federal Register: November 9, 2010 (75 FR 68837). The supplemental letters dated September 27, November 29, and December 30, 2010, and April 1, June 14, 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 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 17, 2011.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 25th day of August 2011.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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