E6-22492. 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 December 8, 2006 to December 21, 2006. The last biweekly notice was published on December 19, 2006 (71 FR 75987).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazard 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 01F21, 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].

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 amendments request: September 28, 2006.

Description of amendment request: The proposed amendment would modify technical specification (TS) requirements of TS 3.8.3, “Diesel Fuel Oil,” to include a new Condition A with associated Required Action and Completion Time. The proposed Condition A allows the main fuel oil storage tank to be inoperable for up to 14 days for the purpose of performing inspection, cleaning, or repair activities.

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

1. 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 alter the assumption of the accident analyses or the Technical Specification Bases. The inclusion of provisions to permit internal inspection of the main fuel oil storage tank during plant operation does not impact the availability of the EDGs to perform their intended safety function. Furthermore, while the main fuel oil storage tank is out of service, the availability of on-site and off-site fuel oil sources ensures that an adequate supply of fuel oil remains available. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical change to the design of the Diesel Fuel Oil System, nor does it alter the assumptions of the accident analyses. The inclusion of provisions to permit internal inspection and cleaning of the main fuel oil storage tank during plant operation does not introduce any new failure modes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change alters the method of operation of the Diesel Fuel Oil System. However, the availability of the EDGs to perform their intended safety function is not impacted and the assumptions of the accident analyses are not altered. 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Branch Chief: D. Pickett.

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

Date of amendment request: November 27, 2006.

Description of amendment request: The proposed amendment would modify the Technical Specifications (TSs) by relocating references to specific American Society for Testing and Materials (ASTM) standards for fuel oil testing to licensee-controlled documents and adding alternate criteria to the “clear and bright” acceptance test for new fuel oil.

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 by a reference to a generic analysis published in the Federal Register on February 22, 2006 (71 FR 9179), which is presented below:

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

Response: No.

The proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Requirements to perform testing in accordance with applicable ASTM standards are retained in the TS as are requirements to perform surveillances of both new and stored diesel fuel oil. Future changes to the licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, tests and experiments,” to ensure that such changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to recognize more rigorous testing of water and sediment content. Relocating the specific ASTM standard references from the TS to a licensee-controlled document and allowing a water and sediment content test to be performed to establish the acceptability of new fuel oil will not affect nor degrade the ability of the emergency diesel generators (DGs) to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, the changes do not involve 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.

The proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs.

Therefore, the changes do 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 changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Instituting the proposed changes will continue to ensure the use of applicable ASTM standards to evaluate the quality of both new and stored fuel oil designated for use in the emergency DGs. Changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there is no significant reduction in a margin of safety.

The “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The margin of safety provided by the DGs is unaffected by the proposed changes since there continue to be TS requirements to ensure fuel oil is of the appropriate quality for emergency DG use. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies while maintaining sufficient controls to preserve the current margins of safety.

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

Attorney for licensee: David G. Pettinari, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: April 11, 2006.

Description of amendment request: The proposed amendments would (Item 1) revise the Technical Specifications (TSs) and delete the license conditions related to steam generator (SG) tube integrity and (Item 2) revise an organizational description in TS 5.2.1 that is solely administrative in nature and unrelated to the SG tube integrity TSs.

The changes related to SG tube integrity are consistent with the consolidated line-item improvement process (CLIIP), Nuclear Regulatory Commission-approved Revision 4 to Technical Specification Task Force (TSTF) Standard TS Change Traveler, TSTF-449, “Steam Generator Tube Integrity.”

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 is presented below:

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

(Item 1) SG Tube Integrity

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 (steam generator tube rupture) SGTR event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. In the analysis of a 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 MSLB, 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 to the TS 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 to the TS. 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 1-131 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 1-131 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 0.27 gallons per minute with no more than 135 gallons per day in any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS values before the accident.

The proposed change 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 SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the consequences of an MSLB (main steamline break), rod ejection, or a reactor coolant pump locked rotor event, or other previously evaluated accident.

(Item 2) Organization Description Revision in TS 5.2.1

The proposed change revises an organizational description in TS 5.2.1 to conform to an application for consent to the indirect transfer of control of the renewed facility operating licenses. The proposed change does not affect the operation of any equipment, and is solely administrative in nature; therefore, the proposed change has no impact on any accident probabilities or consequences.

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

(Item 1) SG Tube Integrity

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.

(Item 2) Organization Description Revision in TS 5.2.1

There are no new accident causal mechanisms created as a result of this proposed change. No changes are being made to the plant that will introduce any new accident causal mechanisms. This change is solely administrative in nature and does not impact any plant systems that are accident initiators; therefore, no new accident types are being created.

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

(Item 1) SG Tube Integrity

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.

(Item 2) Organization Description Revision in TS 5.2.1

Margin of safety is related to confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. This proposed change is solely administrative in nature and does not affect the performance of the barriers. Consequently, no safety margins will be impacted.

Attorney for licensee: Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street EC07H, Charlotte, NC 28202.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: November 16, 2006.

Description of amendment request: The proposed amendments would authorize revision to revise the Updated Final Safety Analysis Report (UFSAR) to describe the flood protection measures for the auxiliary building.

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

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

No. This License Amendment Request (LAR) proposes the use of a realistic seismic evaluation of the Auxiliary Building sprinkler system (high pressure service water) piping which demonstrates that these non-Category I (non-seismic) self-actuating sprinkler systems will not fail during a Maximum Hypothetical Earthquake (MHE) and clarifies Duke's commitment toward Auxiliary Building flood protection measures in the Updated Final Safety Analysis Report (UFSAR). The proposed change does not affect any Chapter 15 accident analyses. Operation in accordance with the amendment authorizing this change would not involve any accident initiation sequences or change the consequences of any accident analyzed.

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?

No. This LAR proposes the use of a realistic seismic evaluation of the Auxiliary Building sprinkler system (high pressure service water) piping which demonstrate that these non-Category I (non-seismic) self-actuating sprinkler systems will not fail during a MHE and clarifies Duke's commitment toward Auxiliary Building flood protection measures in the UFSAR. Operation in accordance with this proposed amendment will not result in a change in the parameters governing plant operation and will not generate any new accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

No. This LAR proposes the use of a realistic seismic evaluation of the Auxiliary Building sprinkler system (high pressure service water) piping, which demonstrates that these non-Category I (non-seismic) self-actuating sprinkler systems will not fail during a MHE and clarifies Duke's commitment toward Auxiliary Building flood protection measures in the UFSAR. Operation in accordance with this proposed amendment will not result in a change in the parameters governing plant operation and will not affect any Chapter 15 accident analyses. 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: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, Charlotte, NC 28202.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: April 11, 2006.

Description of amendment request: The proposed amendment would add Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber. The proposed changes are consistent with approval of TS Task Force (TSTF) change TSTF-372, Revision 4, “Addition of LCO 3.0.8, Inoperability of Snubbers.”

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on November 24, 2004 (69 FR 68412).

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 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 allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in Regulatory Guide 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: September 26, 2006.

Description of amendment request: The proposed amendment would allow up to eight AREVA NP Inc. Modified Advanced Mark-BW(A) fuel assemblies containing M5 alloy to be placed in nonlimitng Braidwood Station, Unit No. 1 core regions for evaluation during Cycle 14, 15, and 16. The proposed amendment would also remove all references to Joseph Oat spent fuel storage racks that have been physically removed from the spent fuel pool.

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. The proposed TS [technical specification] change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The AREVA Advanced Mark-BW(A) fuel is similar in design to the Westinghouse fuel that will be co-resident in the core. The Advanced Mark-BW(A) fuel assemblies are also similar in design to the Advanced Mark-BW assemblies using M5 alloy material for the cladding, structural tubing, and grids generically approved for use in Westinghouse 3- and 4-loop designed pressurized water reactors with 17 × 17 fuel rod arrays. The AREVA Advanced Mark-BW(A) fuel assemblies will be placed in nonlimiting regions (i.e., locations) of the core. The Cycle 14, 15, and 16 reload designs will meet all applicable design criteria. EGC [Exelon Generation Company, LLC] will use the NRC-approved standard reload design models and methods to demonstrate that all applicable design criteria will be met. Evaluations will be performed as part of the cycle specific reload safety analysis for the operation of the AREVA Advanced Mark-BW(A) fuel to confirm that the acceptance criteria of the existing safety analyses continue to be met. Operation of the AREVA Advanced Mark-BW(A) fuel will not significantly increase the predicted radiological consequences of accidents postulated in the Updated Final Safety Analysis Report.

The proposed change regarding removal of all references in TS to the Joseph Oat spent fuel racks is administrative and deletes unnecessary wording relating to equipment that is physically removed from the Braidwood Station spent fuel pool and therefore does not alter the design, configuration, operation, or function of any plant system, structure or component. As a result, the administrative change does not affect the outcome of any previously evaluated accidents.

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

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

The AREVA Advanced Mark-BW(A) fuel is similar in design to the Westinghouse fuel that will be co-resident in the core. The Advanced Mark-BW(A) fuel assemblies are also similar in design to the Advanced Mark-BW assemblies using M5 alloy material for the cladding, structural tubing, and grids generically approved for use in Westinghouse 3- and 4-loop designed pressurized water reactors with 17 x 17 fuel rod arrays. The Braidwood Station, Unit [No.] 1 cores in which the fuel operates will be designed to meet all applicable design criteria and ensure that all pertinent licensing basis criteria are met. Demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of accident. The reload core designs for the cycles in which the AREVA Advanced Mark-BW(A) fuel will operate will demonstrate that the use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies in nonlimiting core regions (i.e., locations) is acceptable. The relevant design and performance criteria will continue to be met and no new single failure mechanisms will be created. The use of AREVA Advanced Mark-BW(A) fuel does not involve any alteration to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors.

The proposed change regarding removal of all references in TS to Joseph Oat spent fuel racks is administrative and deletes unnecessary wording relating to equipment that is physically removed from the Braidwood Station spent fuel pool and therefore does not alter the design, configuration, operation, or function of an plant system, structure or component. As a result, the administrative change does not create any new or different kind of accident.

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

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

Operation of Braidwood Station, Unit [No.] 1 with up to eight AREVA Advanced Mark-BW(A) fuel assemblies in nonlimiting core regions (i.e., locations) does not change the performance requirements on any system or component such that any design criteria will be exceeded. The normal limits on core operation defined in the Braidwood Station TS will remain applicable for the use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies during Cycles 14, 15, and 16. The reload core designs for the cycles in which the AREVA Advanced Mark-BW(A) fuel will operate will specifically evaluate any pertinent differences, including both mechanical design differences and the past irradiation history, between the AREVA Advanced Mark-BW(A) fuel product, and the Westinghouse fuel product that will be co-resident in the core. The use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies will be specifically evaluated during the reload design process using reload design models and methods as approved by the NRC.

The proposed change regarding removal of all references in TS to Joseph Oat spent fuel racks is administrative and deletes unnecessary wording relating to equipment that is physically removed from the Braidwood Station spent fuel pool and therefore does not alter the design, configuration, operation, or function of an plant system, structure or component. As a result, the administrative change does not affect the ability of any operable structure, system, or component to perform its designated safety function.

Based on this evaluation, 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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Michael L. Marshall, Jr.

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

Date of amendment request: October 19, 2006.

Description of amendment request: The proposed amendments would revise Technical Specification 4.6.2.1.d to allow the frequency of air or smoke flow testing of the containment spray nozzles to be reduced from 10 years to an activity-related frequency following maintenance that could cause a blockage.

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

No. The proposed change revises the surveillance frequency from once per 10 years to following activities that could result in nozzle blockage. The containment spray system nozzles are passive components and are not considered as an initiator of any analyzed event. The proposed change will not impact the ability of the containment spray system to mitigate the consequences of an accident. Industry experience indicates that containment spray systems of similar design are highly reliable and not susceptible to plugging due to the open design of the nozzles, the location of the nozzles high in the containment dome, and the corrosion resistant materials used for construction of the system. The alternative frequency of this surveillance has no impact on the probability of failure of associated active systems. Therefore, there is no significant increase in the probability or consequences of previously evaluated accidents due to the extended surveillance frequency.

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

No. The proposed amendment provides an alternative frequency for performance of the spray nozzle surveillance test. The containment spray nozzles are used for accident mitigation only. Potential unidentified blockage of the containment spray nozzles will not result in the initiation of an accident. The change does not involve a physical alteration of the plant nor does it result in an operational condition different from that which has already been considered in the accident analyses. Therefore, the change does not create the possibility of a new or different kind of accident or malfunction from any accident previously evaluated.

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

No. The alternative frequency of spray nozzle testing has no significant impact on the consequences of any analyzed accident and does not significantly change the failure probability of any equipment that provides protection for the health and safety of the public. The containment spray system will continue to be capable of maintaining containment temperature and pressure below design values. Therefore, there is no significant reduction in the margin of safety.

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Douglas V. Pickett (Acting).

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

Date of amendment request: October 19, 2006.

Description of amendment request: The proposed amendments would revise various Technical Specifications (TSs) to address requirements that should have been changed as part of previously approved amendments. These amendments included TS changes regarding relocation of administrative requirements to licensee controlled programs such as the Topical Quality Assurance Report (TQAR), handling of recently irradiated fuel in accordance with TS Task Force change traveler TSTF-51, and Auxiliary Feedwater Actuation System (AFAS) trip and bypass requirements. The proposed amendments also correct some typographical errors.

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

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

These proposed license amendments require no plant hardware or operational modifications. The proposed changes either correct various administrative errors or incorporate changes that have been justified by previously approved license amendments and should have been made as part of those submittals. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No modifications to either plant hardware or operational procedures are required to support these proposed license amendments; hence, no new failure modes are created. The proposed changes either correct various administrative errors or incorporate changes that have been justified by previously approved license amendments and should have been made as part of those submittals. Therefore, operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The TS corrections proposed by these license amendments are administrative in nature in that they either correct typographical errors (e.g., letter dates and transient limits) or are justified by previous license amendments (i.e., relocation of administrative programs to the TQAR, the implementation of TSTF-51 for recently irradiated fuel, and correct inconsistencies introduced by AFAS trip and bypass requirements). Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Douglas V. Pickett (Acting).

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

Date of amendment request: November 3, 2006.

Description of amendment request: The proposed amendment would modify the Technical Specifications (TS) to reflect a proposed plant modification that will replace the reactor coolant system resistance temperature detectors (RTDs) and bypass piping with fast response thermowell detectors mounted directly in the primary loop piping. The specific TS requirements affected include the notes in Unit 2 TS surveillance requirement for channel calibration of the overtemperature differential temperature (OTΔT) and overpower differential temperature (OPΔT) reactor trip system functions. The proposed change also affects the Unit 1 and Unit 2 TS allowable values for OTΔT and OPΔT reactor trip systems functions.

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 of occurrence or consequences of an accident previously evaluated?

Response: No.

The resistance temperature detectors (RTD) bypass system is the hardware associated with Reactor Coolant System instrumentation having control, indication, and protection functions. The RTD bypass system is not considered a precursor to any previously analyzed accident. The system is relied upon to mitigate the consequences of some accidents. The new system replacing the RTD bypass system will perform the same control, indication, and protection functions, and, similarly, will not be considered a precursor to any accident. The capability of the system to mitigate the consequences of the previously analyzed accidents will not be significantly affected. Therefore, replacement of the existing RTD bypass system with the new system will not increase the probability of occurrence of an accident, and will not increase 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 replacement of the existing RTD bypass with the new system would not create new failure modes, and the replacement system is not an initiator of any new or different kind of accident. The proposed deletion of the note in Technical Specification (TS) Surveillance Requirement 3.3.1.15, and proposed changes to Allowable Values in TS Table 3.3.1-1 do not affect the interaction of the replacement system with any system whose failure or malfunction can initiate an accident. Therefore, the proposed change does not create the possibility of a new [or] different kind of accident from any previously evaluated.

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

Response: No.

Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the models and associated assumptions used to analyze the system's performance. The replacement system will continue to perform the same temperature detection function to the same level of reliability as defined in the Donald C. Cook Nuclear Plant Updated Final Safety Analysis Report. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

The Nuclear Regulatory Commission (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 requests involve no significant hazards consideration.

Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: L. Raghavan.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of amendment requests: December 14, 2006.

Description of amendment requests: The proposed amendments would delete Section 2.G of the Diablo Canyon Power Plant Facility Operating Licenses, which requires reporting of violations of the requirements in Sections 2.C, 2.E, and 2.F of the Facility Operating License.

The NRC staff issued a notice of opportunity to comment in the Federal Register on August 29, 2005 (70 FR 51098), on possible amendments to eliminate the license condition involving reporting of violations of other requirements (typically in License Condition 2.C) in the operating license, 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 model for referencing in license amendment applications in the Federal Register on November 4, 2005 (70 FR 67202). The licensee affirmed the applicability of the NSHC determination in its application dated December 14, 2006.

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 is presented below:

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

Response: No.

The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the 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 deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Antonio Fernandez, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: David Terao.

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

Date of amendment request: December 15, 2006.

Description of amendment request: The amendment request would revise the Technical Specifications (TSs) to adopt NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-372, “Addition of LCO [Limiting Condition for Operation] 3.0.8, Inoperability of Snubbers.” The amendment would add (1) a new LCO 3.0.8 addressing when one or more required snubbers are unable to perform their associated support function(s) (i.e., the snubber is inoperable) and (2) a reference to LCO 3.0.8 in LCO 3.0.1 on when LCOs shall be met.

The NRC staff issued a notice of opportunity for comment in the Federal Register on November 24, 2004 (69 FR 68412), on possible license amendments adopting TSTF-372 using the NRC's consolidated line item improvement process (CLIIP) for amending licensee's TSs, which included a model safety evaluation (SE) and model no significant hazards consideration (NSHC) determination. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252), which included the resolution of public comments on the model SE. The May 4, 2005, notice of availability referenced the November 24, 2004, notice. The licensee has affirmed the applicability of the following NSHC determination in its application.

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 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 allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering [a] supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in [NRC] RG [Regulatory Guide] 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and the management of plant risk[, which is required by the proposed LCO 3.0.8]. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: David Terao.

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

Date of application for amendment: April 11, 2006, as supplemented by letter dated November 9, 2006.

Brief description of amendment: The amendment modifies Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR),” to add a U.S. Nuclear Regulatory Commission-approved topical report to the listing of analytical methods in TS 5.6.5.b. This change will allow for the use of the S-RELAP5 thermal-hydraulic analysis code for the non-loss-of-coolant accident analyses at HBRSEP2.

Date of issuance: November 29, 2006.

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

Amendment No. 211.

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

Date of initial notice in Federal Register: August 29, 2006 (71 FR 51224).

The supplemental letter dated November 9, 2006, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 29, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 27, 2005, as supplemented by letter dated August 22, 2006.

Brief description of amendment: The amendment changes consist of the following changes to the plant Technical Specifications (TSs):

• Adoption of Technical Specification Task Force (TSTF)-258, Revision 4; regarding changes to TS Section 5.0, Administrative Controls.
• Adoption of TSTF-308, Revision 1; regarding the determination of cumulative and projected dose contributions in the Radioactive Effluents Control Program (RECP).
• Revision of the IP2 definition for dose equivalent iodine-131 based on NUREG-1431, Revision 3.
• Revision of the IP2 RECP requirements based on NUREG-1431, Revision 3.
• Revision of the IP3 Explosive Gas and Storage Tank Radioactivity Monitoring Program requirements based on NUREG-1431.

Date of issuance: December 13, 2006.

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

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

Date of application for amendment: February 14, 2006.

Brief description of amendment: The amendment eliminated the requirement to verify containment isolation valves that are maintained locked, sealed, or otherwise secured closed from the monthly position verification. A new surveillance requirement, (SR) 4.6.1.1.d, was also added to replace the existing note and reflects the SR for similar devices located inside containment. In addition, a new note was included to allow verification by use of administrative means of the valves and blind flanges that are located in high-radiation areas. In this regard, the amendment adopts TS Task Force (TSTF) Improved Standard TS Change Traveler No. 45 (TSTF-45-A), “Exempt Verification of Containment Isolation Valves that are Not Locked, Sealed, or Otherwise Secured.”

Date of issuance: December 18, 2006.

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

Amendment No.: 269.

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

Date of initial notice in Federal Register: April 11, 2006 (71 FR 18373).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 18, 2006.

No significant hazards consideration comments received: No.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendment: January 12, 2006.

Brief description of amendment: The amendments revised Technical Specification 3.6.6, “Containment Spray and Cooling Systems,” Surveillance Requirement 3.6.6.3, governing containment cooling train cooling water flow rate, from “ >2660 gallons per minute (gpm) to each train” to “ >2660 gpm to each cooler,” to accurately reflect the plant design.

Date of issuance: December 6, 2006.

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

Amendment Nos.: 149, 149, 143 and 143.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register: April 25, 2006 (71 FR 23954)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 6, 2006.

No significant hazards consideration comments received: No.

Amendment No.: 250 and 232

Facility Operating License Nos. DPR-26 and DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: February 14, 2006 (71 FR 7807).

The letter dated August 22, 2006, supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: June 21, 2006, as supplemented December 12, 2006.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.7.3, “Intake Cooling Water System,” Action a, to increase the allowed outage time for one inoperable intake cooling water pump from 7 days to 14 days.

Date of issuance: December 12, 2006.

Effective date: December 12, 2006.

Amendment Nos.: 232 and 227.

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

Date of initial notice in Federal Register: September 12, 2006 (71 FR 53717). The December 12, 2006, Supplement did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 28, 2006.

Brief description of amendment: The amendment modifies the Technical Specification (TS) requirements for inoperable snubbers by adding Limiting Condition for Operation 3.0.8. This change is based on the NRC-approved Technical Specification Task Force (TSTF) standard TS change TSTF-372, Revision 4. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on May 4, 2005 (70 FR 23252).

Date of issuance: December 14, 2006.

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

Amendment No.: 263.

Facility Operating License No. DPR-49: The amendment revises the TSs.

Date of initial notice in Federal Register: (71 FR 43534) August 1, 2006.

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

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of application for amendment: May 11, 2006.

Brief description of amendment: The amendment revised NMP2 Technical Specification (TS) 3.1.7, “Standby Liquid Control (SLC) System,” (SLCS) by increasing the minimum required NMP2 SLCS pump test discharge pressure specified in TS Surveillance Requirement 3.1.7.7 from 1235 psig to 1320 psig.

Date of issuance: December 14, 2006.

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

Amendment No.: 117.

Facility Operating License No. NPF-69: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: September 26, 2006 (71 FR 56192).

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

No significant hazards consideration comments received: No.

Nuclear Management Company, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of application for amendment: September 15, 2005, as supplemented on April 13, August 21, and August 22, 2006.

Brief description of amendment: The amendment revised the MNGP licensing basis by implementing the full-scope alternative source term methodology, resulting in revision of portions of the Technical Specifications to reflect this licensing basis change.

Date of issuance: December 7, 2006.

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

Amendment No: 148.

Facility Operating License No. DPR-22: Amendment revised the Facility Operating License and Technical Specifications.

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.

Date of initial notice in Federal Register: February 14, 2006 (71 FR 7808).

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

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

Date of application for amendment: January 13, 2006, as supplemented by letter dated September 29, 2006.

Brief description of amendment: The amendment revised TS 5.6.5, “Core Operating Limits Report (COLR),” by adding Westinghouse Topical Report WCAP-16009-P-A, “Realistic Large-Break LOCA [Loss-of-Coolant Accident] Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),” dated January 2005, as an approved analytical method for determining the core operating limits for Diablo Canyon Power Plant, Unit No. 2.

Date of issuance: December 20, 2006.

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

Amendment No.: 192.

Facility Operating License No. DPR-82: The amendment revised the Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register: February 28, 2006 (71 FR 10076).

The September 29, 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 amendment is contained in a Safety Evaluation dated December 20, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 29, 2006, as supplemented July 6, 2006.

Brief description of amendments: The amendments revised the Technical Specifications for containment tendon surveillance to provide consistency with the requirements of the regulations.

Date of issuance: December 12, 2006.

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

Amendment Nos.: 147, 127.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: May 9, 2006 (71 FR 27004). The supplement dated July 6, 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 December 12, 2006.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: December 15, 2005, as supplemented by letters dated June 12 and September 8, 2006 (TS-05-10).

Brief description of amendment: The amendment revises the existing steam generator tube surveillance program and was modeled after the U.S. Nuclear Regulatory Commission's approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4. TSTF-449 is part of the consolidated line item improvement process.

Date of issuance: November 3, 2006.

Effective date: As of the date of issuance and shall be implemented prior to entering Mode 4 during startup from the Unit 1 Cycle 7 refueling outage.

Amendment No.: 65.

Facility Operating License No. NPF-90: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: March 28, 2006 (71 FR 15489 ). 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 amendment is contained in a Safety Evaluation dated: November 3, 2006.

No significant hazards consideration comments received: No. 65.

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 6, 2006 (TS-06-04).

Brief description of amendments: The amendment revises Technical Specification 3.1.3.2, “Position Indication Systems—Operating,” for the Sequoyah Nuclear Plant, Units 1 and 2, to allow for the use of an alternate means other than movable incore detectors to monitor the position of a control or shutdown rod should problems occur with the analog rod position indication system. The use of this alternate method will reduce the frequency of flux mapping using movable incore detectors to determine the position of the non-indicating rod. This will reduce the wear on the movable incore detector system that is also used to complete other required TS surveillances.

Date of issuance: December 11, 2006.

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

Amendment Nos.: 315 and 304.

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 46938). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated

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: May 1, 2006 (TS-05-10).

Brief description of amendments: The amendments modify the Technical Specification (TS) Section 6.0, “Administrative Controls,” to adopt a Nuclear Regulatory Commission-approved topical report that extends the burnup limit of the Mark-BW fuel design with M5 alloy. These amendments also incorporate Technical Specification Task Force Traveler (TSTF) 363, Revision 0, “Revised Topical Report References in Improved Technical Specification 5.6.5, Core Operating Limits Report.” TSTF-363 makes administrative changes to the format of referenced topical reports in the TSs.

Date of issuance: November 16, 2006.

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

Amendment Nos. 314 and 303.

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

Date of initial notice in Federal Register: June 20, 2006 (71 FR 35459).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 16, 2006.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: June 16, 2006.

Brief description of amendment: The amendment revises TS Section 5.7.2.11, “Inservice Testing Program”, consistent with Technical Specification Task Force (TSTF) Traveler 479, Revision 0, “Changes to Reflect Revision of 10 CFR 50.55a” and TSTF 279, Revision 0, “Remove ‘applicable supports’ from Inservice Testing Program.” The changes replace references to Section XI of the ASME Boiler and Pressure Vessel Code with the ASME Operation and Maintenance Code for inservice testing (IST) activities and removes reference to “applicable supports” from the IST program. In addition, the changes limit the applicability of Surveillance Requirement 3.0.2 provisions to other normal and accelerated frequencies specified as two years or less in the IST program.

Date of issuance: December 18, 2006.

Effective date: As of the date of issuance and shall be implemented no later than the start of the second 10-year IST interval.

Amendment No. 66.

Facility Operating License No. NPF-90: Amendment revises the technical specifications.

Date of initial notice in Federal Register: August 15, 2006 (71 FR 46939). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 18, 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: May 9, 2006.

Brief description of amendment: The amendment revised Technical Specifications (TS) 1.1, “Definitions,” and TS 3.4.16, “RCS [Reactor Coolant System] Specific Activity.” The revisions replaced the current Limiting Condition for Operation (LCO) 3.4.16 limit on RCS gross-specific activity with limits on RCS Dose Equivalent I-131 and Dose Equivalent Xe-133 (DEX). The conditions and required actions for LCO 3.4.16 not being met, as well as surveillance requirements for LCO 3.4.16, are revised. The modes of applicability for LCO 3.4.16 are extended. The current definition of Ē—Average Disintegration Energy in TS 1.1 is replaced by the definition of DEX. In addition, the current definition of Dose Equivalent I-131 in TS 1.1 is revised to allow alternate NRC-approved thyroid dose conversion factors.

Date of issuance: December 18, 2006.

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

Amendment No.: 178.

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

Date of initial notice in Federal Register: June 20, 2006 (71 FR 35461).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 18, 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: May 9, 2006.

Brief description of amendment: The amendment revised Technical Specifications (TS) 1.1, “Definitions,” and TS 3.4.16, “RCS [Reactor Coolant System] Specific Activity.” The revisions replaced the current Limiting Condition for Operation (LCO) 3.4.16 limit on RCS gross-specific activity with limits on RCS Dose Equivalent I-131 and Dose Equivalent Xe-133 (DEX). The conditions and required actions for LCO 3.4.16 not being met, as well as surveillance requirements for LCO 3.4.16, are revised. The modes of applicability for LCO 3.4.16 are extended. The current definition of Ē—Average Disintegration Energy in TS 1.1 is replaced by the definition of DEX. In addition, the current definition of Dose Equivalent I-131 in TS 1.1 is revised to allow alternate NRC-approved thyroid dose conversion factors.

Date of issuance: December 18, 2006.

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

Amendment No.: 178.

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

Date of initial notice in Federal Register: June 20, 2006 (71 FR 35461).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 18, 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: February 1, 2006, as supplemented by letter dated May 24, 2006.

Brief description of amendment: The amendment revised the Inservice Testing Program in Section 5.5.8 of the Technical Specifications, “Administrative Controls, Programs and Manuals,” to adopt the Commission-approved Technical Specification Task Force (TSTF)-479, Revision 0, “Changes to Reflect Revision of 10CFR50.55a.”

Date of issuance: November 15, 2006.

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

Amendment No.: 172.

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

Date of initial notice in Federal Register: February 28, 2006 (71 FR 10079).

The supplemental letter dated May 24, 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 published in the Federal Register.

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

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this December 26, 2006.

For the Nuclear Regulatory Commission.

Timothy J. McGinty,

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