E7-20679. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 staff) 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 September 27, 2007, to October 10, 2007. The last biweekly notice was published on October 9, 2007 (72 FR 57352).

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 10 CFR 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. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

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, Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. 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 persons 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 01F21, 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 within 60 days, 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 set forth 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 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, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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.

A request for a hearing or a petition for leave to intervene must be filed 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; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, [e-mail]; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at [phone], verification number is [phone]. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to [phone] or by e-mail to [e-mail]. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, 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 01F21, 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. 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 [phone], [phone] or by e-mail to [e-mail].

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

Date of amendment request: September 24, 2007.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to add a reference to Dominion Topical Report DOM-NAF-5, “Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),” to the list of approved analytical methods. The proposed changes would permit the application of the Dominion nuclear core design and safety analysis methods, including the methodology to perform core thermal-hydraulic analysis to predict critical heat flux and departure from nucleate boiling ratio for the Westinghouse 422 V+ fuel design. The proposed amendment would also: (1) Accommodate the use of the methodologies proposed in DOM-NAF-5, (2) delete one approved analytical method that will no longer be used, and (3) delete date and revision numbers from the current TS list of approved analytical methods, consistent with TS Task Force (TSTF) Change Traveler TSTF-363-A, Revision 0, “Revise Topical Report References in ITS [improved TSs] 5.6.5, COLR [Core Operating Limits Report],” dated August 4, 2003.

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 analysis methods of DOM-NAF-5 do not make any contribution to the potential accident initiators and thus do not increase the probability of any accident previously evaluated. The use of the approved Dominion analysis methodologies will not increase the probability of an accident because plant systems, structures, and components (SSC) will not be affected or operated in a different manner, and system interfaces will not change.

Since the applicable safety analysis and nuclear core design acceptance criteria will be satisfied when the Dominion analysis methods are applied to KPS, the use of the approved Dominion analysis methods does not increase the potential consequences of any accident previously evaluated. The use of the approved Dominion methods will not result in a significant impact on normal operating plant releases, and will not increase the predicted radiological consequences of postulated accidents described in the USAR [updated safety analysis report].

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

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

Response: No.

The use of Dominion analysis methods and the Dominion statistical design limit (SDL) for fuel departure from nucleate boiling ratio (DNBR) and fuel critical heat flux (CHF) does not impact any of the applicable core design criteria. All pertinent licensing basis limits and acceptance criteria will continue to be met. Demonstrated adherence to these limits and acceptance criteria precludes new challenges to SSCs that might introduce a new type of accident. All design and performance criteria will continue to be met and no new single failure mechanisms will be created. The use of the Dominion methods does not involve any alteration to plant equipment or procedures that might introduce any new or unique operational modes or accident precursors.

Therefore, the proposed amendment does not create 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.

Nuclear core design and safety analysis acceptance criteria will continue to be satisfied with the application of Dominion methods. Meeting the analysis acceptance criteria and limits ensures that the margin of safety is not significantly reduced. Nuclear core design and safety analysis acceptance criteria will continue to be satisfied with the application of Dominion methods. In particular, use of VIPRE-D with the proposed SDL provides at least a 95% probability at a 95% confidence level that DNBR will not occur (the 95/95 DNBR criterion). The required DNBR margin of safety for KPS, which is the margin between the 95/95 DNBR criterion and clad failure, is therefore not reduced.

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, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Acting Branch Chief: Travis L. Tate.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: August 28, 2007.

Brief description of amendments: Revision to the Operating License and Technical Specification (TS) 1.0, “Use and Application, and TS 3.7.17”, “Spent Fuel Assembly Storage,” to Revise Rated Thermal Power from 3458 megawatts thermal (MWt) to 3612 MWt.

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. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The impacts of the proposed Stretch Power Uprate (SPU) on plant systems, structures, and components (SSCs) were reviewed with respect to SSC design capability, and it was determined that following completion of plant changes to support the SPU, no system, structure, or component would exceed its design conditions or limits. Evaluations supporting those conclusions were performed consistent with proposed Technical Specification changes. Consequently, equipment reliability and structural integrity will not be adversely affected. Control system studies demonstrated that plant response to operational transients under SPU conditions will not significantly increase reactor trip frequency, so there will be no significant increase in the frequency of SSC challenges caused by reactor trip.

New systems are not needed to implement the SPU, and new interactions among SSCs are not created. The SPU does not create new failure modes for existing SSCs. Modified components do not introduce new failure modes relative to those of the components in their pre-modified condition. Consequently, new initiators of previously analyzed accidents are not created.

The fission product barriers—fuel cladding, reactor coolant pressure boundary, and the containment building—remain unchanged. The spectrum of previously analyzed postulated accidents and transients was evaluated, and effects on the fuel, the reactor coolant pressure boundary, and the containment were determined. These analyses were performed consistent with the proposed Technical Specification changes. The results demonstrate that existing reactor coolant pressure boundary and containment limits are met and that effects on the fuel are such that dose consequences meet existing criteria at SPU conditions.

There is no increase in the probability of an accident concerning the potential insertion of a fuel assembly in an incorrect location in the Spent Fuel Pool Region I/Region II racks as a result of the specified storage patterns. Luminant Power [Luminant Generation Company LLC] has used administrative controls to move fuel assemblies from location to location since the initial receipt of fuel on site. Fuel assembly placement will continue to be controlled pursuant to approved fuel handling procedures and in accordance with the Technical Specification for spent fuel rack storage configuration limitations.

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

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

Response: No.

New systems are not required to implement the SPU, and new interactions among SSCs are not created. The SPU does not create new failure modes for existing SSCs. Modified components do not introduce failures different from those of the components in their pre-modified condition. Consequently, no new or different accident sequences arise from SSC interactions or failures.

Training will be provided to address SPU effects, and the plant's simulator will be updated consistent with SPU conditions. Operating procedure changes are minor and do not result in any significant changes in operating philosophy. For these reasons, the SPU does not introduce human performance issues that could create new accidents or different accident sequences.

The increase in power level does not create new fission product release paths. The fission product barriers (fuel cladding, reactor coolant pressure boundary, and the containment building) remain unchanged.

The potential for criticality in the spent fuel pool is not a new or different type of accident. The potential criticality accidents have been reanalyzed to demonstrate that the pool remains subcritical.

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

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

Structural evaluations performed at SPU conditions demonstrated that calculated loads on affected SSCs remain within their design for all design basis event categories. American Society of Mechanical Engineers (ASME) Code fatigue limits continue to be met.

Fuel performance evaluations were performed using parameter values appropriate for a reload core operating at SPU conditions. Those evaluations demonstrate that fuel performance acceptance criteria continue to be met. Loss of Coolant Accident (LOCA) and non-LOCA safety analyses were performed assuming SPU conditions and consistent with the proposed Technical Specification change. Emergency core cooling system performance was shown to meet the criteria of 10 CFR 50.46. The non-LOCA events identified in the Final Safety Analysis Report (FSAR) Chapter 15 were shown to meet existing acceptance criteria.

The containment building response to mass and energy releases was evaluated assuming SPU conditions. The evaluations showed that temperature and pressure limits were met.

No plant changes associated with the SPU reduce the degree of component or system redundancy. Existing Technical Specification operability and surveillance requirements are not reduced by the proposed changes.

The proposed fuel storage requirements in Technical Specification 3.7.17 will provide adequate margin to assure that the fuel storage array (Region I and Region II) will always remain subcritical by the 5% margin recommended by the Nuclear Regulatory Commission (NRC).

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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Branch Chief: Thomas G. Hiltz.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: August 16, 2007.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) requirements related to control room envelope habitability in TS 3.7.9, “Control Room Emergency Air Treatment System (CREATS),” and TS section 5.5, “Programs and Manuals.” The changes are consistent with the Nuclear Regulatory Commission approved Industry/Technical Specification Task Force (TSTF)-448, Revision 3. The availability of this TS improvement was published in the Federal Register on January 17, 2007, 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), an analysis of the issue of no significant hazards consideration adopted by the licensee 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 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. 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 significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. 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 analysis adopted by the licensee 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: Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment requests: September 24, 2007.

Description of amendment requests: Approval of the revision to the San Onofre Nuclear Generating Station Unit 3 Technical Specification 5.5.2.15, “Containment Leakage Rate Testing Program.” The request is for a one-time extension from the currently approved 15-year interval since the last Integrated Leak Rate Test (ILRT) to a 16-year interval since the last ILRT.

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. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10 CFR 50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 15 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the probability or consequences of an accident since research documented in NUREG-1493, “Performance-Based Containment System Leakage Testing Requirements,” September 1995, has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. The NUREG concluded that reducing the Type A testing frequency to once per twenty years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The most recent Type A test at Unit 3 shows leakage to be below acceptance criteria, indicating a leak tight containment. Inspections required by the American Society of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) are performed in order to identify indications of containment degradation that could affect leak tightness. Type B and C testing required by Technical Specifications will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test extension will not represent a significant increase in the consequences of an accident.

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

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

Response: No.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10 CFR 50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 16 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing cannot create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10 CFR 50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 15 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing will not significantly reduce the margin of safety. The NUREG 1493, “Performance-Based Containment System Leakage Testing Requirements,” September 1995, generic study of the effects of extending containment leakage testing found that a 20 year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG 1493 found that, generically, the design containment leakage rate contributes about 0.1 percent to the individual risk and that the decrease in Type A testing frequency would have a minimal [e]ffect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the American Society of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants[”]) will further reduce the risk of a containment leakage path going undetected.

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

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

NRC Branch Chief: Thomas G. Hiltz.

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: September 19, 2007.

Description of amendment request: The proposed amendments would revise various Technical Specification (TS) setting limits and the overtemperature ΔT/overpower ΔT time constants in TS 2.3 and TS 3.7. The methodology for determining the revised setting limits and time constants is in agreement with methods 1 and 2 in ISA-RP67.04, Part II.

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 [Limited Safety System Settings] LSSSs and setting limits to ensure that safety limits are not exceeded as a result of normal and expected instrument drift between calibration intervals. The new allowable values (LSSSs and setting limits) were derived to meet the intent of RIS 2006-17, “NRC Staff Position on the Requirements of 10 CFR 50.36, `Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels,” dated August 24, 2006.

The proposed TS change does not change any of the previously evaluated accidents in the Updated Final Safety Analysis Report (UFSAR). Rather, the proposed change ensures that reactor trip system and engineered safety function actuation system actuations occur as designed and within safety limits. In addition, it increases the probability that a malfunctioning instrument channel will be identified.

This change is not considered to represent a significant increase in the probability or consequences of an accident, since it will decrease the probability of the malfunction of a system, structure or component (SSC), thereby decreasing the probability or consequences of an accident previously evaluated. Specifically, the change is conservative in nature since it will increase the likelihood that a malfunctioning instrument channel will be identified prior to that channel exceeding its safety limit.

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 LSSSs and setting limits to ensure that safety limits are not exceeded as a result of normal and expected instrument drift between calibration intervals.

The change is conservative and is intended to ensure the safety analysis is maintained. Specifically, the proposed change is intended to identify a malfunctioning channel prior to its exceeding the safety limit sooner than the current instrument setting methodology. Therefore the proposed change 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?

No. The proposed change revises LSSSs and setting limits to ensure that safety limits are not exceeded as a result of normal and expected instrument drift between calibration intervals. The new allowable values (LSSS and setting limits) were derived to meet the intent of RIS 2006-17, “NRC Staff Position on the Requirements of 10 CFR 50.36, ‘Technical Specifications,’ Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels,” dated August 24, 2006.

Channel statistical allowance (CSA) calculations have been performed on channels with an associated safety analysis limit to determine the instrument channel uncertainty. Channel operational test (COT) errors are associated with those portions of the instrument channel tested to verify channel operability. These COT errors were extracted from the CSA to derive an allowable value for the channel. The allowable value is set at a distance from the actual (nominal) trip setpoint equal to the COT errors (with some minimal additional margin on some channels). The overall result is a reduction in the distance between the allowable value and the nominal trip setpoint. Consequently, for a malfunctioning channel, the allowable value will be exceeded with less drift and, therefore, corrective action will be initiated sooner after implementation of the proposed change. This will increase the likelihood that the safety analysis limit for the channel is not exceeded.

The distance between the safety analysis limit and the nominal trip setpoint has not been decreased; therefore, the safety margin has [not been] reduced. The likelihood that a malfunctioning channel is identified prior to exceeding its safety analysis limit has increased. 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., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Branch Chief: Evangelos C. Marinos.

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.

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

Date of amendment request: March 29, 2007.

Brief description of amendment request: The proposed amendments would revise the Catawba Nuclear Station, Units 1 and 2, Technical Specification section 3.5.2.8, and the associated Bases and authorize changes to the Updated Final Safety Analysis Report concerning modifications to the emergency core cooling system sumps.

Date of publication of individual notice in Federal Register: August 13, 2007, (72 FR 45274).

Expiration date of individual notice: October 15, 2007.

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

Date of amendment request: April 30, 2007.

Brief description of amendment request: The proposed amendments would revise the Catawba Nuclear Station, Unit 2, Technical Specification Section 5.5.9 concerning modifications to the steam generator tube repair criteria.

Date of publication of individual notice in Federal Register: August 13, 2007, (72 FR 45272).

Expiration date of individual notice: October 15, 2007.

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 at 1 [phone], [phone] or by e-mail to [e-mail].

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

Date of application for amendment: May 15, 2006, as supplemented by letters dated October 6, 2006, December 12, 2006, May 31, 2007, July 25, 2007, and September 4, 2007.

Brief description of amendment: The amendment consists of changes to various technical specifications (TSs) regarding steam generator tube integrity. It is based on Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, “Steam Generator Tube Integrity,” and is adapted for the custom TSs used at TMI-1.

Date of issuance: September 27, 2007.

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

Amendment No.: 261.

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

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40744). The supplements dated October 6, 2006, December 12, 2006, May 31, 2007, July 25, 2007, and September 4, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 27, 2007.

No significant hazards consideration comments received: No.

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 application for amendments: May 2, 2007.

Brief description of amendments: Consistent with the Nuclear Regulatory Commission approved Technical Specification Task Force-427, Revision 2, the amendments add a new limiting condition for operation (LCO) 3.0.9, to the TS. LCO 3.0.9 will allow the licensee to delay declaring an LCO not met for equipment supported by barriers unable to perform their associated support function for up to 30 days provided that risk is assessed and managed.

Date of issuance: September 27, 2007.

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

Amendment Nos.: 282 and 259.

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

Date of initial notice in Federal Register: June 19, 2007 (72 FR 33781). The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated September 27, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 28, 2006, as supplemented by letter dated September 20, 2007.

Brief Description of amendments: The amendments changed Technical Specification (TS) 3.8.3, “Diesel Fuel Oil,” to allow the main fuel oil storage tank to be taken out of service for 14 days for inspection, maintenance, and associated repairs on a one-time basis.

Date of issuance: September 27, 2007.

Effective date: Date of issuance to be implemented within 60 days.

Amendment Nos.: 242 and 270.

Renewed Facility Operating License Nos. DPR-71 and DPR-62: Amendments changed the TSs.

Date of initial notice in Federal Register: January 3, 2007 (72 FR 148). The supplement dated September 20, 2007, 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 September 27, 2007.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: January 18, 2007.

Brief description of amendment: The amendment revises the expiration time limit of the reactor coolant system Pressure/Temperature limit graphs in Technical Specifications (TS); revises the adjusted reference temperature for the reactor vessel; and revises the Low Temperature Overpressure Protection (LTOP) arming temperature value specified in TSs. It also makes editorial changes in the use of inequality signs in TSs associated with the LTOP arming temperature in order to make them consistent.

Date of issuance: October 4, 2007.

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

Amendment No.: 235.

Facility Operating License No. DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: April 10, 2007 (72 FR 17946). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 4, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 25, 2006, as supplemented by letters dated June 15, September 7, September 20, and September 21, 2007.

Brief description of amendment: The amendment provides the Technical Specification (TS) changes and evaluations of the radiological consequences of design-basis accidents for implementation of a full-scope alternative source term methodology.

Date of issuance: September 28, 2007.

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

Amendment No.: 226.

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

Date of initial notice in Federal Register: February 27, 2007 (72 FR 8804). 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 amendment is contained in a Safety Evaluation dated September 28, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 29, 2007, as supplemented by letter dated August 20, 2007.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.5.5, “Trisodium Phosphate,” and the associated surveillance requirements by replacing the containment sump buffering agent, trisodium phosphate, with sodium tetraborate decahydrate (STB). In particular, the amendment revises the TS Limiting Condition for Operation (LCO) 3.5.5, with a new weight requirement for STB. The title of the TS section is also changed from “Trisodium Phosphate” to “Containment Sump Buffering Agent and Weight Requirements.”

Date of issuance: October 2, 2007.

Effective date: As of the date of issuance and shall be implemented during the 2007 refueling outage, prior to Mode 3 entry following refueling.

Amendment No.: 227.

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

Date of initial notice in Federal Register: July 10, 2007 (72 FR 37544). The supplemental letter 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 amendment is contained in a Safety Evaluation dated October 2, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 18, 2007, as supplemented by letters dated July 16 and September 20, 2007.

Brief description of amendment: The amendment changes Technical Specification (TS) Surveillance Requirement (SR) 3.5.2.9, to make the surveillance consistent with the plant design following planned modifications to the containment sump. Entergy Nuclear Operations' (ENO) modification removes the existing emergency core cooling system (ECCS) suction inlet screens. In lieu of the ECCS suction inlet screens, ENO is installing passive strainer assemblies on the 590 foot elevation of containment. The SR change was necessary to reflect the change in equipment.

Date of issuance: October 4, 2007.

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

Amendment No.: 228.

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

Date of initial notice in Federal Register: June 19, 2007 (72 FR 33782). 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 amendment is contained in a Safety Evaluation dated October 4, 2007.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2 (ANO-2), Pope County, Arkansas

Date of application for amendment: March 30, 2007, as supplemented on June 13, 2007.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.9.12, “Fuel Storage,” and its associated tables, figures, and surveillance requirements, TS 5.3, “Fuel Storage,” and adds TS 6.5.17, “Metamic Coupon Sampling Program.” The ANO-2 TS 3.9.12 is changed to: (1) Support higher fuel assembly uranium-235 (U-235) enrichment; (2) apply the appropriate loading restrictions; and (3) delete the dry cask loading restrictions. ANO-2 TS 5.3.1 b is changed to reflect a different spent fuel pool boron concentration that is needed to assure K-effective remains less than or equal to 0.95. ANO-2 TS 5.3.2a is modified to reflect a higher fuel assembly U-235 enrichment. A new coupon sampling program is added as TS 6.5.17, and TS 4.9.12.d is added to direct performance of the coupon sampling program.

Date of issuance: September 28, 2007.

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

Amendment No.: 273.

Renewed Facility Operating License No. NPF-6: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: May 8, 2007 (72 FR 26175). The supplement dated June 13, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 28, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 26, 2006, as supplemented by letter dated August 8, 2007.

Brief description of amendment: The amendment revised the technical specifications to allow the AREVA NP Inc. Advanced Mark-BW(A) fuel assemblies to be loaded into the Braidwood Station, Unit 1 core for operating Cycles 15, 16, and 17.

Date of issuance: October 4, 2007.

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

Amendment Nos.: 145/145.

Facility Operating License Nos. NPF-72 and NPF-77: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: (72 FR 152; January 3, 2007). The August 8, 2007, 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 amendments is contained in a Safety Evaluation dated October 4, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 18, 2006, as supplemented by letter dated, March 26, 2007.

Brief description of amendments: The amendments would modify the technical specifications (TS) to risk-inform requirements regarding selected required action end states consistent with the Nuclear Regulatory Commission (NRC)-approved industry and TS task force (TSTF-423), Revision 0, “Technical Specifications End States, NEDC-32988-A.” This TSTF was published in the Federal Register on March 23, 2006, as part of the consolidated line item improvement.

Date of issuance: September 27, 2007.

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

Amendment Nos.: 184/171.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register: May 8, 2007 (72 FR 26177). The March 26, 2007, 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 amendments is contained in a Safety Evaluation dated September 27, 2007.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

Date of application for amendment: February 9, 2007, as supplemented by letters dated August 8, August 23, and September 13, 2007.

Brief description of amendment: The amendment will address Generic Safety Issue 191 “Assessment of Debris Accumulation on PWR Sump Performance,” by implementing Technical Specification (TS) changes that reflect the use of a new recirculation spray system pump start signal due to a modification to the containment sump screens and replace the use of LOCTIC with the Modular Accident Analysis Program-Design Basis Accident calculation methodology to calculate containment pressure, temperature, and condensation rates for input to the SWNAUA code, which ultimately changes the aerosol removal coefficients used in dose consequence analysis.

Date of issuance: October 5, 2007.

Effective date: As of the date of issuance, and shall be implemented prior to the first entry into Mode 4 coming out of 1R18, which begins September 2007.

Amendment No: 280.

Facility Operating License No. DPR-66: The amendment revised the License and TS.

Date of initial notice in Federal Register: April 24, 2007 (72 FR 20383). The supplements dated August 8, August 23, and September 13, 2007, 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 October 5, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 26, 2007, as supplemented by letter dated July 11, 2007.

Brief description of amendments: The amendment conforms the license to reflect the direct transfer of Wisconsin Electric Power Company's ownership interest and the Nuclear Management Company's operating authority for the renewed Facility Operating License, Nos. DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2 (Point Beach) to FPL Energy Point Beach, LLC, as approved by order of the Commission order dated July 31, 2007. Transfer of the licenses will also authorize FPL Energy Point Beach, LLC, pursuant to the general license requirements in 10 CFR 72.210, to store spent fuel in the Independent Spent Fuel Storage Installation at Point Beach.

Date of issuance: September 28, 2007.

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

Amendment Nos.: 228, 233.

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

Date of initial notice in Federal Register: February 28, 2007 (72 FR 9035). The July 11, 2007, supplement contained clarifying information and did not change the staff's initial proposed finding of no significant hazards consideration. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 31, 2007.

No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: October 12, 2006.

Brief description of amendment: The amendment revises the number of fuel assemblies that are allowed to be stored in the spent fuel pool (SFP) from 1879 to 1321 in Technical Specification (TS) 4.3.3 and removes the reference to Type 4 SFP storage racks in TS limiting condition for operation 3.7.13.

Date of issuance: October 1, 2007.

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

Amendment No.: 103.

Renewed Facility Operating License No. DPR-18: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: November 7, 2006 (71 FR 65145). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2007.

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: April 28, 2006, and as supplemented by letters dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007.

Brief description of amendments: The change increased the minimum allowed boron concentration of the spent fuel pool and allowed credit for soluble boron, guide tube inserts made from borated stainless steel, and fuel storage patterns in place of Boraflex.

Date of issuance: September 27, 2007.

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

Amendment Nos.: Unit 2-213; Unit 3-205.

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: June 6, 2006 (71 FR 32606). The supplemental letters dated November 13 and December 22, 2006, May 7, June 15, July 27, and September 11, 2007, 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 September 27, 2007.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: February 13, 2007.

Brief description of amendments: The amendments revised the Technical Specifications for refueling interlocks.

Date of issuance: October 4, 2007.

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

Amendment Nos.: 253, 197.

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

Date of initial notice in Federal Register: March 27, 2007 (72 FR 14308). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 4, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 12, 2006, as supplemented on December 7, 2006, January 26, 2007, May 8, 2007, August 14, 2007, and August 22, 2007.

Brief description of amendments: The amendments revise the technical specifications to establish 674 feet as the minimum water level of the ultimate heat sink and 87 °F as the maximum supply header temperature of the emergency raw water cooling system.

Date of issuance: September 28, 2007

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

Amendment Nos.: 317 and 307.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the technical specifications.

Date of initial notice in Federal Register: August 15, 2006 (71 FR 46939). The supplements dated December 7, 2006, January 26, 2007, May 8, 2007, August 14, 2007, and August 22, 2007, 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 September 28, 2007.

No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: April 18, 2007, as supplemented by letters dated July 20, and October 2, 2007.

Brief description of amendments: Amendments revise the licenses to reflect changes in legal name of TXU Generation Company LP to Luminant Generation Company LLC.

Date of issuance: October 9, 2007.

Effective date: As of the date of issuance and shall be implemented within 7 days from the date of issuance.

Amendment Nos.: Unit 1-139; Unit 2-139.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: June 13, 2007 (72 FR 32685). The supplemental letters dated July 20 and October 2, 2007, 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 September 10, 2007.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 15th day of October, 2007.

For the Nuclear Regulatory Commission.

John P. Boska,

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