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 15, 2006, to September 28, 2006. The last biweekly notice was published on September 26, 2006 (71 FR 56189).

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, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. 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 (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. 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 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: May 30, 2006.

Description of amendment request: The proposed amendment would revise technical specification (TS) Definitions 1.14, “LEAKAGE”, and 1.16, “PRESSURE BOUNDARY LEAKAGE”; revise TS 3/4.6.2, “Reactor Coolant System Operational Leakage”; add a new TS 3/4.4.5, “Steam Generator (SG) Tube Integrity;” add a new TS 6.8.4.g, “Steam Generator (SG) Program;” and add a new TS 6.9.1.12, “Steam Generator Tube Inspection Report”; as well as administrative and editorial changes. These changes are consistent with the NRC-approved Revision 4 to TS Task Force (TSTF) Standard TS change traveler, TSTF-449, “Steam Generator Tube Integrity.” The proposed changes are necessary in order to implement the guidance for the industry initiative on Nuclear Energy Institute (NEI) 97-06, “Steam Generator Program Guidelines.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on March 2, 2005 (70 FR 10298), on possible amendments concerning TSTF-449, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on May 6, 2005 (70 FR 24126). The licensee affirmed the applicability of the following NSHC determination in its application dated May 30, 2006.

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

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change requires a SG Program that includes performance criteria that will provide reasonable assurance that the SG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specification). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational LEAKAGE.

A[n] SGTR [SG tube rupture] event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. In the analysis of a[n] SGTR event, a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in the licensing basis plus the LEAKAGE rate associated with a double-ended rupture of a single tube is assumed.

For other design basis accidents such as [an] MSLB [main steam line break], rod ejection, and reactor coolant pump locked rotor[,] the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses typically assume that primary to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident induced stresses. The accident induced leakage criterion introduced by the proposed changes accounts for tubes that may leak during design basis accidents. The accident induced leakage criterion limits this leakage to no more than the value assumed in the accident analysis.

The SG performance criteria proposed change[s] to the TS[s] to identify the standards against which tube integrity is to be measured. Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. The performance criteria are only a part of the SG Program required by the proposed change[s] to the TS[s]. The program, defined by NEI 97-06, Steam Generator Program Guidelines, includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of design basis accidents are, in part, functions of the DOSE EQUIVALENT 1[I]-131 in the primary coolant and the primary to secondary LEAKAGE rates resulting from an accident. Therefore, limits are included in the plant technical specifications for operational leakage and for DOSE EQUIVALENT 1[I]-131 in primary coolant to ensure the plant is operated within its analyzed condition. The typical analysis of the limiting design basis accident assumes that primary to secondary leak rate after the accident is 1 gallon per minute with no more than [500 gallons per day or 720 gallons per day] in any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 1[I]-131 are at the TS values before the accident.

The proposed change does not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The proposed approach updates the current TSs and enhances the requirements for SG inspections. The proposed change does not adversely impact any other previously evaluated design basis accident and is an improvement over the current TSs.

Therefore, the proposed change does not affect the consequences of a[n] SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the consequences of an MSLB, rod ejection, or a reactor coolant pump locked rotor event, or other previously evaluated accident.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

The proposed performance based requirements are an improvement over the requirements imposed by the current technical specifications. Implementation of the proposed SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The result of the implementation of the SG Program will be an enhancement of SG tube performance. Primary to secondary LEAKAGE that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions.

The proposed change does not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change does not impact any other plant system or component. The change enhances SG inspection requirements.

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

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the [a] Margin of Safety

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of an SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the SG Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the SG Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TSs.

For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the TS.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Daniel S. Collins.

Tennessee Valley Authority (TVA), Docket No. 50-259 , Browns Ferry Nuclear Plant (BFN), Unit 1, Limestone County, Alabama

Date of amendment request: September 22, 2006 (TS-431).

Description of amendment request: The proposed amendment supplements a June 28, 2004, request to increase the licensed thermal power from 3293 megawatt thermal (MWt) to 3952 MWt, an approximate 20 percent increase in thermal power. This supplement requests interim approval of an increase in licensed thermal power from 3293 MWt to 3458 MWt with an attendant 30-psi increase in reactor pressure. This represents an approximate 5 percent increase above the original licensed thermal power (OLTP) of 3293 MWt. An interim approval would provide for operation at 105 percent power until such time as certain steam dryer analyses can be completed. The NRC staff's review of the remainder of the June 2004 application would resume upon receipt of the satisfactorily completed steam dryer analyses.

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

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

Response: No.

The probability (frequency of occurrence) of Design Basis Accidents occurring is not affected by the increased power level, because BFN Unit 1 continues to comply with the regulatory and design basis criteria established for plant equipment. An evaluation of the Boiling Water Reactor probabilistic risk assessments concludes that the calculated core damage frequency does not significantly change due to operation at 105% OLTP.

Scram setpoints (equipment settings that initiate automatic plant shutdowns) are established such that there is no significant increase in scram frequency due to operation at 105% OLTP. No new challenges to safety-related equipment result from operation at 105% OLTP.

The probability of Design Basis Accidents occurring is not affected by taking credit for containment overpressure in ensuring adequate NPSH [Net Positive Suction Head] for the BFN Unit 1 low pressure ECCS pumps. NRC Bulletin 96-03 requested that BWR [Boiling-Water Reactors] owners implement appropriate measures to minimize the potential clogging of the Emergency Core Cooling System (ECCS) suppression chamber strainers by potential debris generated by a LOCA [loss-of-coolant accident]. TVA installed new, high-capacity passive strainers on BFN Unit 1 of the same design as BFN Units 2 and 3. In addition, TVA's proposed resolution of NRC Bulletin 96-03 for BFN Unit 1 takes credit for containment overpressure to maintain adequate ECCS pump Net Positive Suction Head (NPSH). Containment pressure will increase following a pipe break occurring inside containment. Crediting containment overpressure in the analysis of the consequences of the Loss of Coolant Accident (LOCA) does not affect the precursors for the LOCA, nor does it affect the precursors for any other accident or transient analyzed in Chapter 14 of the BFN Updated Final Safety Analysis Report (UFSAR). Therefore, there is no increase in the probability of any accident previously evaluated.

The changes in consequences of hypothetical accidents, which would occur from 102% of the stretch power uprate reactor thermal power compared to those previously evaluated, are in all cases insignificant. The stretch power uprate accident evaluations do not exceed any of their NRC-approved acceptance limits. The spectrum of hypothetical accidents and transients has been investigated, and are shown to meet the plant's currently licensed regulatory criteria. In the area of core design, for example, the fuel operating limits such as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and Safety Limit Minimum Critical Power Ratio (SLMCPR) are still met, and fuel reload analyses will show plant transients meet the criteria accepted by the NRC. Challenges to fuel (ECCS performance) are evaluated, and shown to continue to meet the criteria of 10 CFR 50.46.

Challenges to the containment have been evaluated at the increased power level, and the containment and its associated cooling systems continue to meet the design and licensing criteria. Radiological release events (accidents) have been evaluated at the increased power level, and shown to be less than the limits of 10 CFR 50.67.

The radiological consequences of the design basis accident are not increased by taking credit for the post-LOCA suppression chamber airspace pressure. The containment will continue to function as designed. This proposed change only takes credit for containment pressure that would exist following a LOCA. Crediting this pressure in ensuring adequate ECCS NPSH will not result in an increase in containment leakage assumed in any analysis.

Therefore, the proposed amendment does not result in a significant increase in consequences or a significant increase in the probability or consequences of any 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.

Equipment that could be affected by operation at 105% OLTP has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario or equipment failure mode was identified. The full spectrum of accident considerations has been evaluated and no new or different kind of accident has been identified. Operation at 105% OLTP uses developed technology, and applies it within the capabilities of existing plant safety related equipment in accordance with the regulatory criteria, including NRC approved codes, standards and methods. No new power dependent accidents have been identified.

The BFN Unit 1 TS [Technical Specifications] require revision to implement operation at 105% OLTP. All revisions have been assessed, and it has been determined that the proposed change will not introduce a different accident than that previously evaluated.

The proposed use of the post-LOCA suppression chamber airspace pressure in the calculation of NPSH for the ECCS pumps does not introduce any new modes of plant operation or make physical changes to plant systems. Rather, the post-LOCA suppression chamber airspace pressure is a consequence of the conditions that would exist in the containment following a large pipe break inside containment. The proposed amendment does not introduce new equipment which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created.

Therefore, the change 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?

Response: No.

The calculated loads on all affected structures, systems and components will remain within their design allowables for all design basis event categories. No NRC acceptance criterion is exceeded. Because the BFN Unit 1 configuration and reactions to transients and hypothetical accidents does not result in exceeding the presently approved NRC acceptance limits, operation at 105% OLTP does not involve a significant reduction in a margin of safety.

The post-LOCA suppression chamber airspace pressure is a byproduct of the conditions that will exist in the containment after a line break inside containment. Conservative analyses have been performed that demonstrate that sufficient post-accident suppression chamber airspace pressure will be available to meet the NPSH requirements for the low pressure ECCS pumps. By enabling credit of these conditions for the low pressure ECCS pumps, adequate NPSH margin will be ensured, and accordingly, the ECCS pumps will meet their performance requirements. Therefore, the credit for containment overpressure 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: L. Raghavan.

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 51.12(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 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: December 2, 2005.

Brief description of amendment: The amendment revised the Oyster Creek Nuclear Generating Station Technical Specifications (TSs) to increase the allowable as-found main steam safety valve code safety function lift setpoint tolerance from ±1% to ±3%.

Date of Issuance: September 13, 2006.

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

Amendment No.: 261.

Facility Operating License No. DPR-16: The amendment revised the TSs.

Date of initial notice in Federal Register: January 17, 2006 (71 FR 2588).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated September 13, 2006.

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: June 7, 2005, as supplemented on May 12, 2006.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) to eliminate the use of the defined term Core Alterations. The amendments incorporate the changes reflected in TS Task Force (TSTF) Travelers 471-T (TSTF-471-T), “Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes,” and TSTF-51-A, “Revise containment requirements during handling irradiated fuel and core alterations.” In addition, the amendments revise TS 3.9.2, “Nuclear Instrumentation,” by replacing “Core Alterations” with “positive reactivity additions” in the Required Action for an inoperable source range monitor during refueling operations. The limiting conditions for operation in TS 3.9.4, “Shutdown Cooling (SDC) and Coolant Recirculation—High Water Level,” are also revised by replacing “core alterations” with “movement of fuel assemblies within containment.”

Date of issuance: September 21, 2006

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

Amendment Nos.: 279 and 256.

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

Date of initial notice in Federal Register: July 5, 2005 (70 FR 38716).

The May 12, 2006, letter 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 these amendments is contained in a Safety Evaluation dated September 21, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 3, 2005, as supplemented by letter dated July 6, 2006.

Brief description of amendment: The amendment revises the requirements of Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR).”

Date of issuance: September 20, 2006.

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

Amendment No. 209.

Renewed Facility Operating License No. DPR-23. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: May 24, 2005 (70 FR 29787). The supplemental letter provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 20, 2004, as supplemented by letters dated June 22, 2005, June 26, 2006, and September 18, 2006.

Brief description of amendment: The amendment revises Table 3.3.1-1, Functions 3, 14, 17.a., 20, and the footnote related to Function 20.

Date of issuance: September 22, 2006.

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

Amendment No. 210.

Renewed Facility Operating License No. DPR-23. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: November 23, 2004 (69 FR 68182). The letters dated June 22, 2005, June 26, 2006, and September 18, 2006, provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 22, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 27, 2004, as supplemented September 27, 2004, October 20, 2004, March 23, 2005, January 30, 2006 and May 25, 2006.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to incorporate a full-scope application of an alternate source term methodology in accordance with 10 CFR 50.67.

Date of issuance: September 15, 2006.

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

Amendment No.: 232.

Facility Operating License No. NPF-49: The amendment revised the TSs.

Date of initial notice in Federal Register: September 14, 2004 (69 FR 55468). The supplements contained clarifying information only, and did not change the initial no significant hazards consideration determination or expand the scope of the initial Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 13, 2005, as supplemented by letters dated June 13 and August 14, 2006.

Brief description of amendment: The amendment revised the Technical Specification (TS) surveillance requirements for the recirculation spray system.

Date of issuance: September 20, 2006.

Effective date: As of the date of issuance and shall be implemented prior to entering Mode 1 following refueling outage 3R11.

Amendment No.: 233.

Facility Operating License No. NPF-49: The amendment revised the TSs.

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61657). The supplements dated June 13 and August 14, 2006, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 29, 2005, as supplemented May 1, 2006, and September 8, 2006.

Brief description of amendments: The amendments requested authorization to revise the Updated Final Safety Analysis Report and the emergency operating procedures to allow an additional operator action to manually start one containment air return fan in the air return system in response to Nuclear Regulatory Commission Bulletin 2003-01, “Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors,” June 6, 2003.

Date of issuance: September 25, 2006.

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

Amendment Nos.: 231 and 227.

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

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61657).

The supplement dated May 1, 2006, 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 September 25, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 25, 2005, as supplemented July 28, 2005, and August 1, 2005.

Brief description of amendments: The amendments revised the Technical Specifications temperature limit for the standby nuclear service water pond from 91.5 °F to 95 °F.

Date of issuance: September 25, 2006.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance September 25, 2006.

Amendment Nos.: 232 and 228.

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

Date of initial notice in Federal Register: August 4, 2005 (70 FR 44946).

The supplements dated July 28, 2005, and August 1, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 25, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 13, 2005, as supplemented March 20, 2006.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) to correct a nonconservative TS associated with spent fuel storage in the spent fuel pool. The licensee identified the nonconservative TS while comparing results from spent fuel pool criticality codes.

Date of issuance: September 27, 2006.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance September 27, 2006.

Amendment Nos.: 233 and 229.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the license and the technical specifications.

Date of initial notice in Federal Register: November 21, 2005 (70 FR 70104). The supplement dated March 20, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 29, 2005, as supplemented May 1, 2006.

Brief description of amendments: The amendments requested authorization to revise the Updated Final Safety Analysis Report and the emergency operating procedures to allow an additional operator action to manually start one containment air return fan in the air return system in response to Nuclear Regulatory Commission Bulletin 2003-01, “Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors,” June 6, 2003.

Date of issuance: September 25, 2006.

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

Amendment Nos.: 234 and 216.

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

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61657).

The supplement dated May 1, 2006, 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 September 25, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 29, 2005, as supplemented by letter dated May 18, 2006.

Brief description of amendment: By letter dated June 29, 2005, Entergy Operations, Inc., the licensee for Arkansas Nuclear One, Unit 2 (ANO-2), requested a license amendment to relocate the shutdown cooling (SDC) open permissive interlock (OPI) license condition from the operating license to the licensee's technical requirements manual. The license condition to maintain OPI operability was previously accepted by the NRC staff in a letter to the licensee, dated March 30, 2005, and incorporated into ANO-2's operating license. The OPI prevents the two SDC suction isolation valves from opening above a selected set point to separate the high-pressure reactor coolant system from the low-pressure SDC system.

Date of issuance: September 13, 2006.

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

Amendment No.: 267.

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

Date of initial notice in Federal Register: December 6, 2005 (70 FR 72671). The supplement dated May 18, 2006, 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 13, 2006.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: December 14, 2004, revised by letter dated August 30, 2006.

Brief description of amendment: The Technical Specification amendment relocates structural integrity requirements to the Final Safety Analysis Report.

Date of issuance: September 14, 2006.

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

Amendment No.: 224.

Facility Operating License No. DPR-35: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: March 1, 2005 (70 FR 9991).

The licensee originally requested for additional TS relocations in their submittal dated December 14, 2004. The NRC staff found these unacceptable. Therefore, the licensee revised the original application by letter dated August 30, 2006, reducing the scope of the application as originally noticed. Hence, there is no change to the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 14, 2006.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: June 2, 2005, supplemented by letter dated June 14, 2006.

Brief description of amendment: The amendment revised the Technical Specification (TS) reactor coolant system leakage detection instrumentation requirements and actions.

Date of issuance: September 20, 2006.

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

Amendment No.: 225.

Facility Operating License No. DPR-35: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: May 23, 2006 (71 FR 29676). The supplement dated June 14, 2006, provided additional information that did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2006.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania

Date of application for amendment: April 11, 2005, as supplemented December 2, 2005, and January 27, April 14, August 16, and September 1, 2006.

Brief description of amendment: The amendment revised the scope of the steam generator tubesheet inspections and subsequent repair using the F* inspection methodology.

Date of issuance: September 27, 2006.

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

Amendment No: 160.

Facility Operating License No. NPF-73: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: June 7, 2005 (70 FR 33214). The supplements dated December 2, 2005, and January 27, April 14, August 16, and September 1, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's issuance of Amendment No. 158 to Facility Operating License NPF-73 for BVPS-2, regarding steam generator tube integrity (Technical Specification Task Force (TSTF) Item 449) on September 7, 2006, resulted in renumbering and rewording the requirements as originally proposed by the licensee to fit the TSTF-449 format, but did not change the scope of the application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 27, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 7, 2005, as supplemented April 25, June 1, and August 3, 2006.

Brief description of amendments: The amendments include changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the steam generator (SG) tube surveillance program (SG tube integrity), and changes to the SG reporting requirements.

Date of issuance: September 7, 2006.

Effective date: As of the date of issuance to be implemented within 90 days for BVPS-1 and prior to entry into Mode 4 following the fall 2006 refueling outage for BVPS-2.

Amendment Nos.: 276 and 158.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications and Licenses.

Date of initial notice in Federal Register: December 20, 2005 (70 FR 75491). The supplements dated April 25, June 1, and August 3, 2006, 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 7, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 17, 2005, as supplemented May 12 and August 22, 2006.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) 3.7.7.1 (BVPS-1), “Control Room Emergency Habitability Systems,” and 3.7.7 (BVPS-2), “Control Room Emergency Air Cleanup and Pressurization System,” by dividing these TSs into two specifications, addressing control room emergency ventilation and control room air cooling functions separately. The amendments also improved consistency with the Standard TSs and improved consistency between the units.

Date of issuance: September 25, 2006

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

Amendment Nos.: 277 and 159

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: April 26, 2005 (70 FR 21458). The supplements dated May 12 and August 22, 2006, 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 25, 2006.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of application for amendments: February 1, 2006.

Brief description of amendments: The amendments revise the Technical Specification (TS) testing frequency for the Surveillance Requirements in TS 3.1.4, “Control Rod Scram Times,” based on the TS Task Force (TSTF) change traveler TSTF-222, Revision 1.

Date of issuance: September 12, 2006.

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

Amendment Nos.: 237 and 214.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the License and TSs.

Date of initial notice in Federal Register: May 9, 2006 (71 FR 27001).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 12, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 30, 2005, as supplemented July 21, 2006.

Brief description of amendment: The amendment revises the Virgil C. Summer Nuclear Station Technical Specifications to permit the use of a best estimate methodology in performing loss-of-coolant accident analyses.

Date of issuance: September 7, 2006.

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

Amendment No. 176.

Renewed Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: October 11, 2005 (70 FR 59087). The supplemental letter provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 7, 2006.

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: November 30, 2005, as supplemented by letter dated May 30, 2006.

Brief description of amendments: The proposed amendments revised the Technical Specification (TS) requirements related to steam generator tube integrity, based on the NRC-approved Revision 4 to TS Task Force (TSTF)-449, “Steam Generator Tube Integrity.” Date of issuance: September 19, 2006.

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

Amendment Nos.: Unit 2—204; Unit 3—196.

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

Date of initial notice in Federal Register: February 14, 2006 (71 FR 7812). The May 30, 2006, supplemental letter 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 September 19, 2006.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendment: March 9, 2004 (TS-434) as supplemented on November 15, 2004, and March 7, 2006.

Brief description of amendment: The amendment reduced the Allowable Value used for Reactor Vessel Water Level—Low, Level 3, for several instrument functions.

Date of issuance: September 18, 2006.

Effective date: September 18, 2006.

Amendment No.: 258.

Renewed Facility Operating License No. DPR-33: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19575). The supplements dated November 15, 2004, and March 7, 2006, 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 18, 2006.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendments: November 3, 2003, as supplemented May 6, 2004, and August 1, 2006.

Description of amendment request: The amendment revised Technical Specification (TS) Table 3.3.1.1 -1, Reactor Protection system Instrumentation, Function 7.b.

Date of issuance: September 19, 2006.

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

Amendment No.: 259.

Renewed Facility Operating License No. DPR-33: Amendment revised the TSs.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19575). The supplements dated May 6, 2004, and August 1, 2006, 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 19, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 16, 2004, as supplemented by letters dated March 11, 2005, November 4, 2005, and April 14, 2006.

Description of amendment request: To extend the channel calibration frequency requirements for instrumentation in the high-pressure coolant injection, reactor core isolation cooling, and reactor water core isolation cooling systems.

Date of issuance: September 21, 2006.

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

Amendment Nos.: 260, 297 and 255.

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

Date of initial notice in Federal Register: May 23, 2006 (71 FR 29680). The supplemental letters provided clarifying information that did not expand the scope of the original application or change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2006.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendment: July 9, 2004 (TS 436).

Brief description of amendment: The amendment revised Technical Specification (TS) Surveillance Requirement 3.6.1.3.10 to increase the allowed main steam isolation valve (MSIV) leak rate from 11.5 standard cubic feet per hour (scfh) per valve to 100 scfh for individual MSIVs with a 150 scfh combined leakage for all four main steam lines.

Date of issuance: September 27, 2006.

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

Amendment No.: 261.

Facility Operating License No. DPR-33: Amendment revised the TSs.

Date of initial notice in Federal Register: May 23, 2006 (71 FR 29680).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 27, 2006.

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: August 14, 2004, as supplemented on April 11, 2005, and July 11, 2006 (TS-02-01).

Brief description of amendments: The amendments revise Technical Specifications (TSs) relating to the reactor protection system and engineered safety features instrumentation. The Trip Setpoint column of TS Tables 2.2-1 and 3.3-4 will be renamed Nominal Trip Setpoint; inequality signs in TS Tables 2.2-1 and 3.3-4 will be removed; the trip setpoint and allowable value for the Intermediate Range Neutron Flux P-6 permissive will be revised; Minimum Channels Operable in TS Table 3.3-3 will be revised; editorial changes will be made to TS Table 3.3-4 to replace ± signs with inequalities; and a correction will be made to an alarm/trip setpoint in TS Table 3.3-6.

Date of issuance: September 13, 2006.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos. 310 and 299.

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

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60688). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 13, 2006.

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: September 30, 2004, as supplemented on May 25, 2006.

Brief description of amendments: The amendments revise the technical specifications to relocate the requirements for the emergency diesel generator start loss of power instrumentation and associated actions in the engineering safety features tables to a new limiting condition for operation (LCO). In addition, an upper allowable value limit has been added to the voltage sensors for loss of voltage and degraded voltage consistent with Technical Specification Task Force (TSTF) Item, TSTF-365, along with a lower allowable value limit for the degraded voltage diesel generator start and load shed timer. The auxiliary feedwater loss of power start setpoints and allowable values have been relocated to this new LCO.

Date of issuance: September 14, 2006.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos. 311 and 300.

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

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2900). The supplemental letter provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 14, 2006.

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 29, 2006.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.4.15, “RCS [Reactor Coolant System] Leakage Detection Instrumentation.” The TS changes delete the containment atmosphere gaseous radioactivity monitor from TS 3.4.15 and revise the existing conditions, required actions, completion times, and surveillance requirements in TS 3.4.15 to account for the monitor being deleted. The June 29, 2006, letter superceded the license amendment request in the August 26, 2005, letter to authorize changes to the Final Safety Analysis Report.

Date of issuance: September 26, 2006.

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

Amendment No.: 175.

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

Date of initial notice in Federal Register: July 24, 2006 (71 FR 41843). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 26, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 26, 2006.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.4.15, “RCS [Reactor Coolant System] Leakage Detection Instrumentation.” The TS changes delete the monitor from TS 3.4.15 and revise the existing conditions, required actions, completion times, and surveillance requirements in TS 3.4.15 to account for the monitor being deleted. The June 26, 2006, letter superceded the license amendment request in the August 26, 2005, letter to authorize changes to the Updated Safety Analysis Report.

Date of issuance: September 26, 2006.

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

Amendment No.: 166.

Facility Operating License No. NPF-42. The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 24, 2006 (71 FR 41848).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 26, 2006.

No significant hazards consideration comments received: No.