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                    Document headings vary by document type but may contain                    the following:                  
                -                       the agency or agencies that issued and signed a document                    
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###### Nuclear Regulatory Commission

          - [NRC-2009-0306]

## I. 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 June 18, 2009, to July 1, 2009. The last biweekly notice was published on June 30, 2009 (74 FR 31318).

## 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                           ( print page 34045)                        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.                

                    Written comments may be submitted by mail to the Chief, Rulemaking and Directives Branch (RDB), TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this                     **Federal Register**                     notice. Written comments may also be faxed to the RDB at 301-492-3446. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.                

                    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 Commission's PDR, located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site,                     *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 petitioner/requestor 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 petitioner/requestor 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 petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner/requestor is aware and on which the petitioner/requestor 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 petitioner/requestor to relief. A petitioner/requestor 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, which the NRC promulgated in August 28, 2007 (72 FR 49139). The E-Filing process requires participants to submit and serve all adjudicatory documents over the Internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. 

                    To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the petitioner/requestor should contact the Office of the Secretary by e-mail at                     *hearing.docket@nrc.gov*, or by calling (301) 415-1677, to request (1) a digital 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/or (2) creation of an electronic docket for the proceeding (even in instances in which the                           ( print page 34046)                        petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms ViewerTM                     to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM                     is free and is available at                     *http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html*. 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*.                

                    Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE 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 EIE 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 through the “Contact Us” link located on the NRC Web site at                     *http://www.nrc.gov/​site-help/​e-submittals.html*                     or by calling the NRC Meta-System Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The Meta-System Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at                     *MSHD.Resource@nrc.gov*.                 

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. 

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the request and/or petition should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). 

                    Documents submitted in adjudicatory proceedings will appear in NBC's electronic hearing docket which is available to the public at                     *http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp*, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a 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 submissions.                

                    For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site,                     *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*.                

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

*Date of amendments request:*                     May 13, 2009.                 

*Description of amendments request:*                     The amendment would incorporate Technical Specification Task Force Traveler (TSTF) 479-A, Revision 0, “Changes to Reflect Revision of 10 CFR 50.55a,” and TSTF 497-A, Revision 0, “Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less,” approved by the Nuclear Regulatory Commission (NRC) on December 6, 2005 and October 4, 2006 respectively. The proposed changes revise the Improved Standard Technical Specification administrative controls of the Inservice Testing Program to be consistent with the requirements of Title 10 of the                     *Code of Federal Regulations*                     Section 50.55a(f)(4) for pumps and valves classified as American Society of Mechanical Engineers (ASME) Code Class 1, Class 2, and Class 3.                

The proposed change replaces references to ASME Section XI of the Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants within Technical Specification (TS) 5.5.8. In TS 5.5.8.b, the proposed change applies the extension of Surveillance Requirement 3.0.2 to other normal and accelerated inservice testing frequencies of 2 years or less that were not included in the frequencies of the table listed in TS 5.5.8.a. 

*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. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

                        The proposed change will replace, within TS 5.5.8, references to Section XI of ASME Boiler and Pressure Vessel Code, with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). In addition the proposed change                               ( print page 34047)                            adds words to TS 5.5.8.b which applies the extension allowance of the Surveillance Requirement 3.0.2 to other normal and accelerated inservice testing frequencies of two years or less that were not included in the frequencies of the table listed in TS 5.5.8.a.                    

The proposed change is administrative, does not affect any accident initiators, does not affect the ability to successfully respond to previously evaluated accidents and does not affect radiological assumptions used in the evaluations. Thus, operation of the facility in accordance with the proposed change will not involve an increase in the probability or the consequences of an accident previously evaluated.

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

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will replace, within TS 5.5.8 references to Section XI of ASME Boiler and Pressure Vessel Code with references to the ASME OM Code. In addition the proposed change also adds words to TS 5.5.8.b which applies the extension allowance of Surveillance Requirement 3.0.2 to other normal and accelerated inservice testing frequencies of two years or less that were not included in the frequencies of the table listed in TS 5.5.8.a.

                        The proposed change does not involve a modification to the physical configuration of the plant (                        *i.e.*, no new equipment will be installed) or involve a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally there is no change in the types or increase in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.                    

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

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The proposed change will replace, within TS 5.5.8 references to Section XI of ASME Boiler and Pressure Vessel Code with references to the ASME OM Code. In addition the proposed change also adds words to TS 5.5.8.b which applies the extension allowance of Surveillance Requirement 3.0.2 to other normal and accelerated inservice testing frequencies of two years or less that were not included in the frequencies of the table listed in TS 5.5.8.a.

The proposed change does not involve a modification to the physical configuration of the operating units or change the methods governing normal plant operation. The proposed change incorporates revisions to the ASME Code that results in a net improvement in the measures for testing pumps and valves. The safety functions of the applicable pumps and valves will be maintained.

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 amendments request involves no significant hazards consideration. 

*Attorney for licensee:*                     Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.                 

*NRC Acting Branch Chief:*                     Richard Guzman.                

## DTE Energy, Inc., Docket No. 50-16, Enrico Fermi Atomic Power Plant, Unit 1, Monroe County, Michigan 

*Date of amendment request:*                     March 25, 2009.                 

*Description of amendment request:*                     The proposed amendment would add a license condition incorporating a site license termination plan (LTP) for the Enrico Fermi Atomic Power Plant, Unit 1 (Fermi-1). The proposed license condition includes LTP change control criteria specifying when changes to the LTP require prior Nuclear Regulatory Commission approval. Since Fermi-1 is completely within the boundary of Unit 2, the Fermi-1 property would become part of Unit 2 site upon successful completion of license termination activities.                 

*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. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The change allows for the approval of the LTP and provides the criteria for when changes to the LTP require prior NRC approval. This change does not affect possible initiating events for the three postulated accidents previously evaluated in the Fermi-1 Safety Analysis Report (SAR), as updated, or alter the configuration or operation of the facility. Safety limits, limiting safety system settings, and limiting control systems are no longer applicable to Fermi-1 in the permanently defueled condition, and are therefore not considered further. The proposed change does not affect the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and has no impact on plant operations.

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

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

                        No. The safety analysis for the facility remains accurate as described in the Fermi-1 SAR, as updated, Section 8.4, Postulated Radiological Accidents. There are sections of the LTP that make reference to the decommissioning activities still remaining (                        *e.g.,*                         removal of large components, decontamination, etc.), however these activities are performed in accordance with approved Fermi-1 work authorizations and undergo 10 CFR 50.59 screening prior to initiation. The plant conditions for which the postulated accidents have been evaluated are still valid and no new accident scenarios, failure mechanisms, or single failures are introduced by this amendment. The system operating procedures are not affected.                    

Therefore, the proposed changes will 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?

No. There are no changes to the design or operation of the facility resulting from this amendment. The proposed change does not affect the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and has no impact on plant shutdown operations. Accordingly, neither the postulated accident assumptions in the Fermi-1 SAR, as updated, nor the basis of the Technical Specifications are affected.

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

Based upon the reasoning presented 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:*                     David Pettinari.                

*NRC Branch Chief:*                     Andrew Persinko.                

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

*Date of amendment request:*                     June 3, 2009.                 

*Description of amendment request:*                     The proposed change will modify Technical Specification (TS) 2.1.1.1 to account for the Combustion Engineering (CE) 16 x 16 Next Generation Fuel (NGF) and different U.S. Nuclear Regulatory Commission (NRC) reviewed and approved Departure from Nucleate Boiling (DNB) correlations. These new correlations will be implemented in the safety analyses for the next fuel cycle of                           ( print page 34048)                        operation consistent with NRC-approved methodologies.                 

*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.                     

No changes to plant equipment or operating procedures are required due to the change in the safety limit for DNBR [Departure from Nucleate Boiling Ratio]. This change does not impact any of the accident initiators. The analyses of the reload are performed using NRC-approved methodologies to ensure the Specified Acceptable Fuel Design Limits (SAFDLs), of which DNBR is one, are not violated. The current DNBR setpoint continues to ensure automatic protective action and is initiated to prevent exceeding the proposed DNBR safety limit.

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.                    

The proposed change does not involve any plant modifications or change in the way the plant is designed to function. The proposed change is not associated with any accident precursor or initiator. The proposed change supports the loading and use of Next Generation Fuel (NGF) at Waterford 3 as previously approved by the NRC.

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 change involve a significant reduction in a margin of safety?

*Response:*                         No.                    

The use of the NRC-approved NGF WSSV-T correlation with the ABB-NV correlation to establish a new bounding DNBR safety limit of 1.24, preserves the DNBR margin of safety at a 95/95 level. The Core Protection Calculator (CPC) DNBR power adjustment addressable constant BERR1 is calculated based on the WSSV-T and ABB-NV CHF [Critical Heat Flux] correlations such that a CPC trip at a DNBR of 1.26 using the CE-1 CHF correlation ensures that the bounding DNBR safety limit of 1.24 for the WSSV-T and ABB-NV CHF correlations will not be violated during normal operation and AOOs [anticipated operational occurrences] to at least a 95/95 probability/confidence level.

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:*                     Terence A. Burke, Associate General Counsel—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.                

*NRC Branch Chief:*                     Michael T. Markley.                

## FPL 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 amendment request:*                     April 17, 2009.                

*Description of amendment request:*                     The proposed change would amend the Renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2, respectively, to reflect a change in the legal name of the Licensee from “FPL Energy Point Beach, LLC” to “NextEra Energy Point Beach, LLC.”                

*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.                    

This request is for administrative changes only. No actual facility equipment or accident analyses will be affected by the proposed changes. Therefore, this request will have no impact on 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.                    

This request is for administrative changes only. No actual facility equipment or accident analyses will be affected by the proposed changes and no failure modes not bounded by previously evaluated accidents will be created. Therefore, this request will 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.                    

                        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. This request is for administrative changes only. No actual plant equipment or accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety system settings, and will not relax the bases for any limiting conditions of operation. Therefore, these proposed changes will 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:*                     Antonio Fernandez, Esquire, Senior Attorney, FPL Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.                 

*NRC Branch Chief:*                     Lois M. James.                

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

*Date of amendment request:*                     May 21, 2009 (TSC 09-02).                 

*Description of amendment request:*                     The proposed amendment would revise the Technical Specifications (TSs) Section 6.8.4.k, “Steam Generator Program,” for Unit 2 including associated TSs Bases 3/4.4.5, “Steam Generator (SG) Tube Integrity,” to allow the implementation of SG tubing alternate repair criteria for axial indications in the Westinghouse Electric Company explosive tube expansion region below the top of the tubesheet and specify the W* distance for the SG cold legs.                 

*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.                    

Of the various accidents previously evaluated, the proposed changes only affects the steam generator tube rupture (SGTR) event evaluation and the postulated steam line break (SLB) accident evaluation. Loss-of-coolant accident (LOCA) conditions cause a compressive axial load to act on the tube. Therefore, since the LOCA tends to force the tube into the tubesheet rather than pull it out, it is not a factor in this amendment request. Another faulted load consideration is a safe shutdown earthquake (SSE); however, the seismic analysis of Westinghouse 51 Series steam generators (SGs) has shown that axial loading of the tubes is negligible during an SSE.

                        TVA's [Tennessee Valley Authority's] amendment request allows taking credit for                               ( print page 34049)                            how the tubesheet enhances the tube integrity in the Westinghouse Electric Company explosive tube expansion (WEXTEX) region by precluding tube deformation beyond its initial expanded outside diameter. For the SGTR and SLB events, the required structural margins of the SG tubes will be maintained due to the presence of the tubesheet. Tube rupture is precluded for axial cracks in the WEXTEX region due to the constraint provided by the tubesheet. Therefore, the normal operating 3ΔP margin and the postulated accident 1.43ΔP margin against burst are maintained.                    

The W* length supplies the necessary resistive force to preclude pullout loads under both normal operating and accident conditions. The contact pressure results from the WEXTEX process, thermal expansion mismatch between the tube and tubesheet and from the differential pressure between the primary and secondary side. Therefore, the proposed change results in no significant increase in the probability or the occurrence of an SGTR or SLB accident.

The proposed changes do not affect other systems, structures, components or operational features. Therefore, based on the above evaluation, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

                        The consequences of an SGTR event are primarily affected by the primary-to-secondary flow rate and the time duration of the primary-to-secondary flow during the event. Primary-to-secondary flow rate through a postulated ruptured tube (                        *i.e.*, complete severance of a single SG tube) is not affected by the proposed change since the flow rate is based on the inside diameter of a SG tube and the pressure differential. TVA's amendment request does not change either of these. The duration of primary-to-secondary leakage is based on the time required for an operator to determine that an SGTR has occurred, the time to identify and isolate the faulty SG, and ensure termination of radioactive release to the atmosphere from the faulty SG. TVA's amendment request does not affect the duration of the primary-to-secondary leakage because it does not change the control room indicators with which an operator would determine that an SGTR has occurred. The consequences of an SGTR are secondarily affected by primary-to-secondary leakage, which could occur due to axial cracks remaining in service in the WEXTEX region in a non-faulted SG. During a SGTR, the primary-to-secondary differential pressure is less than or equal to the normal operating differential pressure; therefore, the primary-to-secondary leakage due to axial cracks in the WEXTEX region of a non-faulted SG during a SGTR would be less than or equal to the primary-to-secondary leakage experienced during normal operation. Primary-to-secondary leakage is considered in the calculation determining the consequences of a SGTR and the value is bounding.                    

                        The postulated SLB has the greatest primary-to-secondary pressure differential, and therefore could experience the greatest primary-to-secondary leakage. TVA's amendment request allows axial cracks to remain in service in the WEXTEX region, which have the possibility of primary-to-secondary leakage during a postulated SLB accident. However, the primary-to-secondary leakage would be limited by the amount the crack face can open (compared to a similar free-span axial crack) and by the restriction resulting from the tube to tubesheet contact pressure which create a restricted leakage path from the upper tip of the crack to the top of the WEXTEX expansion. TVA's amendment request requires the aggregate leakage, (                        *i.e.*, the combined leakage for the tubes with axial cracks in the WEXTEX region) plus the combined leakage developed by other alternate repair criteria (ARC), to remain below the maximum allowable SLB primary-to-secondary leakage rate limit such that the doses are maintained to less than a fraction of the 10 CFR 100 limits and also less than the General Design Criteria (GDC)-19 limits.                    

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.                    

TVA's amendment request does not introduce any physical changes to the Sequoyah Unit 2 SGs. TVA's amendment request takes credit for how the tubesheet enhances the SG tube integrity in the WEXTEX region by precluding tube deformation beyond its initial expanded outside diameter and allows axial cracks in the WEXTEX region to remain in service if prescribed criteria are met. Removal of the existing primary water stress corrosion cracking (PWSCC) axial at dented tube support plate ARC incorporates the more conservative TS limit for SG tube plugging. A failure to meet SG tube integrity results in an SGTR. Because the circumferential cracks detected within the WEXTEX region are required to be plugged, it is highly unlikely that a W* tube would fail as a result of a circumferential defect. Therefore, a tube severance which would strike neighboring tubes and create a multiple tube rupture is not credible.

The proposed change does not introduce any new equipment or any change to existing equipment. No new effects on existing equipment are created.

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.                    

The amendment request maintains the structural margins of the SG tubes for both normal and accident conditions that are required by Regulatory Guide 1.121.

For primarily axially oriented cracking located within the tubesheet, tube burst is precluded because of the presence of the tubesheet. WCAP-14797 defines a length, W*, of degradation free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces (with applicable safety factor applied). Application of the W* criteria will preclude unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the W* criteria.

Plugging of the SG tubes reduces the reactor coolant flow margin for core cooling. Implementation of the proposed changes is expected to result in plugging of fewer tubes than with the current criteria. Thus, implementation of the proposed changes will maintain the margin of flow that may have otherwise been reduced by tube plugging.

It is concluded that the proposed changes do not result in a significant reduction of margin with respect to plant safety as defined in the Updated Final Safety Analysis Report or TSs.

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, ET 11A, Knoxville, Tennessee 37902.                 

*NRC Branch Chief:*                     Thomas H. Boyce.                

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

*Date of amendment request:*                     April 13, 2009.                 

*Description of amendment request:*                     The proposed change revises Technical Specifications (TS) Table 3.7-1 Operator Action 3.b and provides direction if the less-than-required-minimum-operable-channels-condition for the nuclear flux intermediate range occurs between 7% and 11% of rated power.                 

*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 license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

                        No. The proposed change revises TS Table 3.7-1 Operator Action 3.b and provides direction if the less-than-required-minimum-operable-channels-condition for the nuclear flux intermediate range occurs between 7% and 11% of rated power. Required action between 7% and 11% of rated power is currently not addressed in the Operator Action 3.b. The proposed TS change does not involve a physical change to any structures, systems, or components (SSCs) at Surry Power Station; nor does it change any of the previously evaluated accidents in the                               ( print page 34050)                            Updated Final Safety Analysis Report (UFSAR). Thus, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.                    

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

No. The proposed change revises TS Table 3.7-1 Operator Action 3.b, and provides required action between 7% and 11% of rated power, which is currently not addressed in the Operator Action 3.b. The proposed change does not involve a physical change to any SSCs, and there is no impact on their design function. The proposed change does not affect initiators of analyzed events. Therefore, the proposed change does not introduce any new failures that could 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?

No. The proposed change provides required action for inoperability of nuclear flux intermediate range instrumentation between 7% and 11% of rated power in TS Table 3.7-1 Operator Action 3.b. Margin of safety is established through the design of plant SSCs, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not impact the condition or performance of SSCs relied upon for accident mitigation or any safety analysis assumptions. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

*Attorney for licensee:*                     Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.                

*NRC Branch Chief:*                     Melanie C. Wong.                

## 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.                

## Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida

*Date of amendment request:*                     May 31, 2009.                

*Description of amendment request:*                     Revise Technical Specification 3.1.3.4, related to requirements for Control Element Assembly (CEA) drop time to increase the available margin for CEA drop time testing.                

*Date of publication of individual notice in the**Federal Register*                    : June 1, 2009 (74 FR 26261).                

*Expiration date of individual notice:*                     July 1, 2009 (Comments) and July 31, 2009 (Hearing).                

## 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 Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site,                     *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 at1 (800) 397-4209, (301) 415-4737 or by e-mail to                     *pdr.resource@nrc.gov*.                

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

*Date of application for amendment:*                     September 12, 2008.                

*Brief description of amendment:*                     The amendment revises Fermi 2 Plant Operating License, Appendix A, Technical Specifications (TS) to remove statements relating to Nuclear Power Plant Staff Working Hours in Section 5.2.2, “Unit Staff,” specifically deleting subsection 5.2.2.e. The requested amendment removes references to and limits imposed by Nuclear Regulatory Commission Generic Letter (GL) 82-12, “Nuclear Power Plant Staff Working Hours” from the administrative controls section of the Fermi 2 TS. The regulations in 10 CFR Part 26, Subpart I, supersede the guidelines in GL 82-12.                

*Date of issuance:*                     June 19, 2009.                

*Effective date:*                     As of the date of issuance and shall be implemented no later than October 1, 2009.                

*Amendment No.:*                     181.                

*Facility Operating License No. NPF-43:*                     Amendment revised the Technical Specifications and License.                

*Date of initial notice in**Federal Register*                    : December 2, 2008 (73 FR 73353).                 

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 19, 2009.

*No significant hazards consideration comments received:*                     No.                

## Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

*Date of application for amendment:*                     January 14, 2009.                

*Brief description of amendment:*                     The amendment: (1) Deleted Technical Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR 3.1.3.3, (2) removed the reference to SR 3.1.3.2 from Required Action A.2 of TS                           ( print page 34051)                        3.1.3, “Control Rod OPERABILITY,” (3) renumbered SRs 3.1.3.3 through 3.1.3.5 to reflect the deletion of SR 3.1.3.2, and (4) revised Example 1.4-3 in Section 1.4, “Frequency,” to clarify the applicability of the 1.25 surveillance test interval extension. The changes are in accordance with NRC-approved TS Task Force (TSTF) change traveler TSTF-475, Revision 1, “Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.”                 

*Date of issuance:*                     June 30, 2009.                 

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

*Amendment No.:*                     212.                 

*Facility Operating License No. NPF-21:*                     The amendment revised the Operating Licenses and Technical Specifications.                 

*Date of initial notice in**Federal Register*                    : February 10, 2009 (74 FR 6665).                 

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 30, 2009. 

*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:*                     January 25, 2008, as supplemented by letters dated April 14 and 29, 2009.                 

*Brief description of amendment:*                     The amendment revised Technical Specification (TS) 3.7.5, “Main Turbine Bypass System.” The change provides an alternative to the existing Limiting Condition for Operation for the Main Turbine Bypass System (MTBS). The revised TS will require that the MTBS be operable or that the Average Planar Linear Heat Generation Rate, the Minimum Critical Power Ratio, and the Linear Heat Generation Rate limits for the inoperable MTBS be placed in effect as specified in the Core Operating Limits Report.                 

*Date of issuance:*                     June 24, 2009.                 

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

*Amendment No.:*                     163.                 

*Facility Operating License No. NPF-47:*                     The amendment revised the Facility Operating License and Technical Specifications.                 

*Date of initial notice in**Federal Register*                    : March 11, 2008 (73 FR 13023). The supplemental letters dated April 14 and 29, 2009, 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 June 24, 2009. 

*No significant hazards consideration comments received:*                     No.                

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

*Date of application for amendment:*                     June 21, 2007, as supplemented by letter dated January 30, 2009.                 

*Brief description of amendment:*                     A change is proposed to the technical specifications (TSs) of Clinton Power Station, Unit 1, consistent with TS Task Force (TSTF) change TSTF-423 to the standard technical specifications (STSs) for boiling-water reactor plants to allow, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed consistent with the program in place for complying with the requirements of Title 10 of the                     *Code of Federal Regulations*                     (10 CFR) Section 50.65(a)(4). The proposed amendment would modify the TS to risk-informed requirements regarding selected required action end states provided in TSTF-423, Revision 0, “Technical Specification End States, NEDC-32988-A.”                 

*Date of issuance:*                     June 26, 2009.                 

*Effective date:*                     As of the date of issuance and shall be implemented within 120 days.                 

*Amendment No.:*                     187.                 

*Facility Operating License No. NPF-62:*                     The amendment revised the Technical Specifications and License.                 

*Date of initial notice in**Federal Register*                    : February 12, 2008 (73 FR 8070). The January 30, 2009, 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 June 26, 2009. 

*No significant hazards consideration comments received:*                     No.                

## Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida 

*Date of application for amendments:*                     June 30, 2008, as supplemented on August 13, 2008.                 

*Brief description of amendments:*                     The proposed amendment modified Technical Specification (TS) requirements related to Refueling Water Tank (RWT) minimum contained volume of borated water. The proposed changes would make permanent the current administrative RWT minimum level of 32.5 feet for both units.                 

*Date of Issuance:*                     June 30, 2009.                 

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

*Amendment Nos.:*                     209 and 157.                 

*Renewed Facility Operating License Nos. DPR-67 and NPF-16:*                     Amendments revised the Technical Specifications.                 

*Date of initial notice in**Federal Register*                    : February 10, 2009 (74 FR 6665). The supplement dated August 13, 2008, 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 June 30, 2009. 

*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:*                     June 26, 2008, as supplemented by letters dated March 16 and May 1, 2009.                 

*Brief description of amendments:*                     The amendments revise the Facility Operating Licenses by revising the licensing basis loss-of-coolant accident and main steam line break accident radiological dose consequences as currently described in the Prairie Island Nuclear Generating Plant Units 1 and 2 Updated Safety Analysis Report. The amendments also make concomitant amendments to Technical Specifications (TSs) 3.3.5, 3.4.17, and 3.6.3, which are necessary to implement the revised analyses.                 

*Date of issuance:*                     June 19, 2009.                 

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

*Amendment Nos.:*                     191, 180.                 

*Facility Operating License Nos. DPR-42 and DPR-60:*                     Amendments revised the Technical Specifications.                 

*Date of initial notice in**Federal Register:*                     September 9, 2008 (73 FR 52420). 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.                           ( print page 34052)    

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 19, 2009. 

*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:*                     June 27, 2008, as supplemented letters dated August 13, 2008, and February 5, 2009.                 

*Brief description of amendments:*                     The changes consist of revisions to the Technical Specifications in support of the replacement of the steam generators (SGs) at San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. The changes reflect revised SG inspection and repair requirements, and revised peak containment post-accident pressure resulting from installation of the replacement SGs.                 

*Date of issuance:*                     June 25, 2009.                 

*Effective date:*                     Upon issuance; to be implemented for Unit 2 prior to entry into Mode 4 during the Unit 2 Cycle 16 refueling outage return-to-service, and to be implemented for Unit 3 prior to entry into Mode 4 during the Unit 3 Cycle 16 refueling outage return-to-service.                 

*Amendment Nos.:*                     Unit 2—220; Unit 3-213.                 

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

*Date of initial notice in**Federal Register:*                     September 23, 2008 (73 FR 54867). The supplemental letters dated August 13, 2008, and February 5, 2009, 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 June 25, 2009.

*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:*                     June 3, 2008, as supplemented by letters dated January 9 and 23, 2009.                

*Brief description of amendment:*                     The amendment revised Technical Specifications (TSs) 3.3.7, 3.3.8, 3.7.10, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 3.8.10. The amendment (1) deleted MODES 5 and 6 from the Control Room Emergency Ventilation System and its actuation instrumentation in TS 3.7.10 and TS 3.3.7; (2) adopted U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) change traveler TSTF-36-A for TSs 3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 3.8.10; and (3) added a more restrictive change to the Limiting Condition for Operation (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.                

*Date of issuance:*                     June 30, 2009.                 

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

*Amendment No.:*                     192.                 

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

*Date of initial notice in**Federal Register:*                     October 7, 2008 (73 FR 58678). The supplemental letters dated January 9 and 23, 2009, 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 June 30, 2009. 

*No significant hazards consideration comments received:*                     No.                

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

*Date of application for amendments:*                     June 9, 2008.                 

*Brief description of amendments:*                     These amendments revise the Technical Specifications to revise action statements in TS 3.12 for insertion limit and shutdown margin requirements, revise the applicability for the operability of the rod position indication and bank demand position indication systems, revise/add action statements for rod position indication, and add action statements for group step demand counters.                 

*Date of issuance:*                     June 25, 2009.                 

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

*Amendment Nos.:*                     265, 264.                 

*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:*                     July 29, 2009 (73 FR 43957).                

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 25, 2009.

*No significant hazards consideration comments received:*                     No.                

Dated at Rockville, Maryland, this 6th day of July 2009.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

 [FR Doc. E9-16347 Filed 7-13-09; 8:45 am]

BILLING CODE 7590-01-P