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Notice

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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Information about this document as published in the Federal Register.

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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 August 30, 2007 to September 12, 2007. The last biweekly notice was published on September 11, 2007 (72 FR 51852).

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 Start Printed Page 54472at 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.

Detroit Edison Company, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: July 20, 2007.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.7.5 to add an Action Statement for two inoperable control building chiller (CBC) subsystems. The proposed new Action Statement would allow 72 hours to restore one CBC subsystem to operable status and require verification once every 4 hours that control room temperature remains less than 90 °F. The proposed changes are consistent, with certain variations, with TS Task Force (TSTF) Change Traveler TSTF-477, Revision 3, “Adding an Action Statement for Two Inoperable Control Room Air Conditioning Subsystems.”

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 December 18, 2006 (71 FR 75774), which 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 is described in Technical Specification Task Force (TSTF) Standard TS Change Traveler TSTF-477 adds an action statement for two inoperable control room subsystems.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed changes add an action statement for two inoperable control room subsystems. The equipment qualification temperature of the control room equipment is not affected. Future changes to the Bases or licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, test 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.

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 consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures.

Therefore, the changes do not involve a significant increase in the probability or consequences of any 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 changes add an action statement for two inoperable control room subsystems. The changes do not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal p[l]ant operation. The requirements in the TS continue to require maintaining the control room temperature within the design limits.

Therefore, the changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed changes add an action statement for two inoperable control room subsystems. Instituting the proposed changes Start Printed Page 54473will continue to maintain the control room temperature within design limits. Changes to the Bases or license[e-] controlled document are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that the control room temperature will be maintained within design limits.

The proposed changes maintain 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: Marjan Mashhadi, Florida Power & Light Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.

NRC Acting Branch Chief: Travis L. Tate.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: July 17, 2007, as supplemented by letter dated August 7, 2007.

Description of amendment request: The proposed amendment would revise the facility operating license (FOL), paragraph 2.C, and technical specifications (TS) 3.7.2 and TS 5.5 for Grand Gulf Nuclear Station, Unit 1.

The NRC staff issued a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments to revise the plant specific TS, to strengthen TS requirements regarding control room envelope (CRE) habitability by changing the action and surveillance requirements associated with the limiting condition for operation operability requirements for the CRE emergency ventilation system, and by adding a new TS administrative controls program on CRE habitability, 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 January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the model NSHC determination in its application dated July 17, 2007, as supplemented by letter dated August 7, 2007.

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 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves NSHC.

Attorney for licensee: Terence A. Burke, Associate General Council—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: August 16, 2007.

Description of amendment requests: A change is proposed to the Waterford 3 Control

Room Emergency Air Filtration System technical specifications (TSs) using the Nuclear Regulatory Commission (NRC) notice of availability regarding Control Room Envelope (CRE) Habitability using the Consolidated Line Item Improvement Process. The proposed amendment is consistent with the NRC approved Industry/Technical Specification Task Force (TSTF) change to the Standard Technical Specifications (STS), TSTF-448, Revision 3, “Control Room Habitability.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments adopting TSTF-448, 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 January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the following NSHC determination in its application dated August 16, 2007.Start Printed Page 54474

TSTF-448, Revision 3 is formatted to the Improved Technical Specification (ITS) plants while the Waterford 3 TSs are based on the CE standard technical specifications. Therefore, the information contained in TSTF-448, Revision 3 has been modified to the Waterford 3 TS format.

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 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the 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: Douglas K. Porter, Esquire, Southern California Edison Company,2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: June 29, 2007.

Description of amendment request: The proposed amendment would change Technical Specifications (TS) sections 3.7.4 and 5.5.13 to strengthen TS requirements regarding control building envelope (CBE) habitability. The proposed amendment would change the action and surveillance requirements associated with the limiting condition for operation operability requirements for the CBE standby filter unit and add a new TS administrative controls program on CBE habitability. The proposed changes to the TS and associated Bases are consistent with certain exceptions with standard technical specifications (STS) as revised by TS Task Force (TSTF) change traveler TSTF-448, Revision 3, “Control Room Envelope Habitability” to the extent that the amendment request adopts by reference certain model TSTF-448 content, where applicable.

The NRC staff issued a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments adopting TSTF-448, 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 January 17, 2007 (72 FR 2022).

The licensee affirmed the applicability of the following NSHC determination in its application dated June 29, 2007.

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 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.Start Printed Page 54475

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

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the 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: Marjan Mashhadi, Florida Power & Light Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.

NRC Acting Branch Chief: Travis L. Tate.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment request: July 30, 2007.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) by adding a new Surveillance Requirement (SR) 3.8.2.2 that would be applicable when onsite electrical power is supplied to a unit via backfeed through the main transformer, and the unit is in either Mode 5 or Mode 6, or during movement of irradiated fuel. The proposed SR would correct a non-conservatism in the TS and will assure the capability to transfer the required safety-related loads from the backfeed source to the qualified offsite circuit.

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 proposed change will add a new Technical Specification Surveillance Requirement applicable during shutdown conditions when a backfeed configuration is used to provide power from the offsite transmission network to required safety equipment via the main transformer. The new Surveillance Requirement will require that portions of an existing Surveillance Requirement be met. If not met, the existing Surveillance Requirement must be performed before establishing a backfeed configuration. It is highly unlikely that the proposed change will necessitate performance of the existing Surveillance Requirement more frequently than is currently required. Even if more frequent performance of the existing Surveillance Requirement were required, its performance would not significantly increase the probability of a loss of offsite power. Consequently, there is no significant change in the likelihood of any accident associated with verifying the existing Surveillance Requirement has been met. Therefore, the probability of occurrence of a previously evaluated accident will not be significantly increased.

The verifications required by the new Surveillance Requirement will assure that a unit's required safety-related equipment can be transferred to a qualified offsite circuit while the equipment is being provided power from the offsite transmission network using a backfeed configuration while the unit is shutdown or while irradiated fuel is [being] moved. This will provide assurance that the systems needed to mitigate the consequences of the accidents in these conditions will be provided with electrical power if the systems are needed to perform their specified safety function. Therefore, the consequences of a previously evaluated accident will not be significantly increased.

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

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

Response: No.

The addition of a new Technical Specification Surveillance Requirement to verify that an existing Surveillance Requirement has been met, or to perform that Surveillance Requirement if not met, would not create the possibility of a new or different kind of accident because the Surveillance Requirement has previously existed and previously been performed. Therefore, the proposed change does not involve any new systems, structures, or components, or any different mode of operation of any existing systems, structures, or components.

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

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

Response: No.

The margin of safety associated with the proposed change involves the availability of offsite electrical power to support required safety equipment when a unit is shut down or during the movement of irradiated fuel. The proposed change provides assurance that the single required qualified offsite circuit from the transmission network remains available while the required safety equipment is powered by a different circuit from that network. Consequently, the proposed change does not reduce the margin of safety provided by the required qualified offsite circuit, and enhances the margin of safety by acknowledging use of an additional offsite circuit.

Therefore, the proposed change does not involve a significant reduction in the 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 Acting Branch Chief: Travis Tate.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: August 10, 2007.

Description of amendment request: The proposed change to Technical Specification 2.1.1.2 will revise two recirculation loop and single recirculation loop safety limit minimum critical power ratio (SLMCPR) values to reflect results of a cycle-specific calculation.

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:

Start Printed Page 54476

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

Response: No.

Four accidents have been evaluated previously as reflected in the CNS [Cooper Nuclear Station] Updated Safety Analysis Report (USAR). These four accidents are (1) loss-of-coolant, (2) control rod drop, (3) main steamline break, and (4) fuel handling. The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications to the plant or any components, nor does it require a change in plant operation. Therefore, no individual precursors of an accident are affected.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. This proposed change makes no modification to the design or operation of the systems that are used in mitigation of accidents. Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change to the value of the SLMCPR continues to conservatively establish this safety limit such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

Based on the above NPPD [Nebraska Public Power District] concludes that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Creation of the possibility of a new or different kind of accident from an accident previously evaluated would require creation of precursors of that accident. New accident precursors may be created by modification of the plant configuration or changes in how the plant is operated. The proposed change does not involve a modification of the plant configuration or in how the plant is operated. The proposed change to the SLMCPR assures that safety criteria are maintained.

Based on the above, NPPD concludes that the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of the rods are expected to be in boiling transition if the Minimum Critical Power Ratio limit is not violated. The proposed change will ensure the appropriate level of fuel protection is maintained. Additionally, operational limits are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.

Based on the above, NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Thomas G. Hiltz.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: August 16, 2007.

Description of amendment request: The proposed amendment revises Technical Specification 5.5.6, “Inservice Testing Program,” to allow a one-time extension of the five-year frequency requirement for setpoint testing of safety valve MS-RV-70ARV.

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

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

Response: No.

The function of SRVs [safety relief valves] and SVs [safety valves] is to prevent overpressurization of the reactor coolant system (RCS) during transients and abnormal operation that could cause increases in RCS pressure. They are also used to depressurize the RCS when needed to allow injection of water from the high-volume, low-pressure Emergency Core Cooling System (ECCS) Low Pressure Coolant Injection mode of the Residual Heat Removal System into the reactor pressure vessel (RPV) as part of mitigation of an accident. Actuation or failure to actuate of a SRV or SV is not an initiator of any accident previously evaluated. Thus, this proposed amendment would not result in a significant increase in the probability of an accident previously evaluated.

A range or tolerance of plus-or-minus three percent of the setpoint pressure is acceptable for the results of setpoint testing. A 90-day extension of the interval for setpoint testing of one SV is not expected to result in actuation of the SV outside of its acceptable setpoint range. However, even if the single SV whose test interval is being extended did actuate outside of its acceptable range, it is not expected that this would result in a significant degradation in the ability of the Nuclear System Pressure Relief System to perform its safety function, since the remaining eight SRVs and two other SVs would be unaffected by the proposed extension of the testing interval for the single SV. The proposed change does not modify the design of or alter the operation of systems or components used in mitigating design basis accidents. Thus, this proposed amendment would not result in a significant increase in the consequences of any accident previously evaluated.

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

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

Response: No.

A new or different kind of accident from any previously evaluated might result from a modification of the plant design by either addition of a new system or removal of an existing system, or a change in how any of the plant systems function during the operation of the plant. The proposed change does not modify the plant design, nor does it alter the operation of the plant or equipment involved in either routine plant operation or in the mitigation of the design basis accidents.

Based on the above, it is concluded that 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 amendment involve a significant reduction in a margin of safety?

Response: No.

The margin of safety applicable to this issue would be the margin between the pressure at which the SRVs and SVs would actuate and the allowable ASME [American Society of Mechanical Engineers] Code overpressure limit of 1,375 psig [pounds per square inch gauge] (110 percent of vessel design pressure, 1250 psig). This margin would be impacted if the setpoint at which the applicable SV actuated experienced drift greater than the allowable plus-or-minus three percent of the setpoint pressure. This is not expected to occur based on the results demonstrated by the setpoint testing conducted over the last ten years. Those results were two actuations of the SV at a pressure below the nameplate rating with less than two percent deviation, and one actuation at a pressure above the nameplate rating with less than one percent deviation. However, even if this one SV did experience setpoint drift greater than the allowable plus-or-minus three percent, there would not be a significant reduction in the margin since it is expected that the remaining eight SRVs and the two other SVs would actuate within the allowable setpoint tolerance and begin to reduce RCS pressure as needed. Furthermore, the proposed extension will not result in a change to the steam discharge capacity and characteristics of the applicable SV.Start Printed Page 54477

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

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: July 23, 2007.

Description of amendment request: The proposed amendment would modify a footnote in NMP2 Technical Specification (TS) Table 3.3.2.1-1, “Control Rod Block Instrumentation,” such that a new banked position withdrawal sequence (BPWS) shutdown sequence could be utilized. The proposed change is consistent with TS Task Force (TSTF) change TSTF-476, Revision 1, “Improved BPWS Control Rod Insertion Process (NEDO-33091).” The availability of the TS change was published in the Federal Register on May 23, 2007 (72 FR 29004) as part of the consolidated line item improvement process. The licensee affirmed the applicability of the model no significant hazards consideration 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 changes modify the TS to allow the use of the improved banked position withdrawal sequence (BPWS) during shutdowns if the conditions of NEDO-33091-A, Revision 2, ”Improved BPWS Control Rod Insertion Process,” July 2004, have been satisfied. The [NRC] staff finds that the licensee's justifications to support the specific TS changes are consistent with the approved topical report and TSTF-476, Revision 1. Since the change only involves changes in control rod sequencing, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting TSTF-476 are no different than the consequences of an accident prior to adopting TSTF-476. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. 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 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 control rod drop accident (CRDA) is the design basis accident for the subject TS changes. This change does not create the possibility of a new or different kind 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 proposed change, TSTF-476, Revision 1, incorporates the improved BPWS, previously approved in NEDO-33091-A, into the improved TS. The control rod drop accident (CRDA) is the design basis accident for the subject TS changes. In order to minimize the impact of a CRDA, the BPWS process was developed to minimize control rod reactivity worth for BWR plants. The proposed improved BPWS further simplifies the control rod insertion process, and in order to evaluate it, the [NRC] staff followed the guidelines of Standard Review Plan Section 15.4.9, and referred to General Design Criterion 28 of Appendix A to 10 CFR Part 50 as its regulatory requirement. The TSTF stated the improved BPWS provides the following benefits: (1) Allows the plant to reach the all-rods-in condition prior to significant reactor cool down, which reduces the potential for re-criticality as the reactor cools down; (2) reduces the potential for an operator reactivity control error by reducing the total number of control rod manipulations; (3) minimizes the need for manual scrams during plant shutdowns, resulting in less wear on control rod drive (CRD) system components and CRD mechanisms; and (4) eliminates unnecessary control rod manipulations at low power, resulting in less wear on reactor manual control and CRD system components. The addition of procedural requirements and verifications specified in NEDO-33091-A, along with the proper use of the BPWS will prevent a control rod drop accident (CRDA) from occurring while power is below the low power setpoint (LPSP). 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: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: July 30, 2007.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) 3.7.3, “Control Room Envelope Air Conditioning (AC) System,” by adding an Action Statement to the Limiting Conditions for Operation. The new Action Statement allows a finite time to restore one control room envelope AC subsystem to operable status and requires verification that the control room temperature remains < 90 °F every 4 hours. The proposed changes are consistent with Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) TSTF-477, Revision 3, “Adding an Action Statement for Two Inoperable Control Room Air Conditioning Subsystems.” The availability of this TS improvement was published in the Federal Register on March 26, 2007 (72 FR 14143) as part of the consolidated line item improvement process. The licensee affirmed the applicability of the model no significant hazards consideration 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 is described in Technical Specification Task Force (TSTF) Standard TS Change Traveler TSTF-477 [and] adds an action statement for two inoperable control room subsystems. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed changes add an action statement for two inoperable control room subsystems. The equipment qualification temperature of the control room equipment is not affected. Future changes to the Bases or licensee controlled documents will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, test 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. 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 Start Printed Page 54478consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures. Therefore, the changes do 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 changes add an action statement for two inoperable control room subsystems. The changes do not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal plant operation. The requirements in the TS continue to require maintaining the control room temperature within the design limits. Therefore, the changes do not create the possibility of a new or different kind 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 proposed changes add an action statement for two inoperable control room subsystems. Instituting the proposed changes will continue to maintain the control room temperature within design limits. Changes to the Bases or license[e-]controlled document[s] are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that the control room temperature will be maintained within design limits. The proposed changes maintain sufficient controls to preserve the current margins of safety.

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

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Mark G. Kowal.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: June 7, 2007.

Description of amendment request: The proposed amendment would delete the license conditions that require reporting of violations of other requirements (e.g., conditions listed in Sections 2.C and 2.F for Unit 1 and Section 2.C for Unit 2) in the operating licenses. This change is in accordance with Nuclear Regulatory Commission (NRC)-approved Technical Specification (TS) Task Force (TSTF) change traveler TSTF-372, Revision 4. The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination in the Federal Register on August 29, 2005 (70 FR 51098). The notice included a model safety evaluation, a model NSHC determination, and a model license amendment request. In its application dated June 7, 2007, the licensee affirmed the applicability of the model NSHC determination which is presented below.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC adopted by the licensee 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 has reviewed the analysis adopted by the licensee and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves NSHC.

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRCBranch Chief: Mark G. Kowal.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: June 8, 2007.

Description of amendment request: The proposed amendment would revise Limiting Condition for Operation (LCO) 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursion greater than 200 degrees Fahrenheit (°F) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4 for SSES 1 and 2. This change is in accordance with Nuclear Regulatory Commission (NRC)-approved Technical Specification (TS) Task Force (TSTF) change traveler TSTF-484, “Use of TS 3.10.1 for Scram Time Testing Activities.” The NRC staff issued a notice of opportunity to comment and notice of availability of a model no significant hazards consideration (NSHC) determination in the Federal Register on August 21, 2006 (71 FR 48561) and October 27, 2006 (71 FR 63050), respectively. The notices included a model safety evaluation, a model NSHC determination, and a model license amendment request. In its application dated June 8, 2007, the licensee affirmed the applicability of the model NSHC determination which is presented below.

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

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

Technical Specifications currently allow for operation at greater than 200 degrees Fahrenheit (°F) while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Technical Specifications currently allow for operation at greater than 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. 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. In addition, the changes do not impose any new or different requirements or Start Printed Page 54479eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Technical Specifications currently allow for operation at greater than 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the analysis adopted by the licensee and, based on this review, it appears that the standards of 10 CFR 50.92 (c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves NSHC.

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief: Mark G. Kowal.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: August 14, 2007.

Description of amendment request: The proposed amendments would add a new license condition to the SSES 1 and 2 Operating Licenses to permit the valves in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J leakage test program to be tested at the higher pressure during the next scheduled test rather than requiring all of the valves to be tested at the higher pressure prior to the implementation of the constant pressure power uprate license amendment.

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 License Condition change does not involve any physical change to structures, systems, or components (SSCs) and does not alter the method of operation or control of SSCs. The current assumptions in the safety analysis regarding accident initiators and mitigation of accidents are unaffected by this change. No additional failure modes or mechanisms are being introduced and the likelihood of previously analyzed failures remains unchanged.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced and installed equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the FSAR [final safety analysis report]. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis.

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change is acceptable because of the satisfactory performance of the Primary Containment Integrated Leak Rate Tests on both Unit 1 and Unit 2 at the new calculated pressure and the substantial margin to leakage rate acceptance limits based upon the Integrated Leak Rate Test and the current LLRT [local leak rate tests] results. Therefore, the plant response to analyzed events will continue to provide the margin of safety assumed by the analysis.

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: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief: Mark G. Kowal.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: April 27, 2007.

Description of amendment request: The proposed amendment revises the Joseph M. Farley Nuclear Plant, Units 1 and 2 Technical Specifications (TS) for Limiting Condition for Operation 3.9.3 “Containment Penetrations,” to allow the containment personnel air locks that provide direct access from the containment atmosphere to the auxiliary building to be open during refueling activities if appropriate administrative controls are established.

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

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

Response: No.

The proposed change would allow the personnel air lock doors, and emergency air lock doors to remain open during fuel movement and core alterations. These doors are normally closed during this time period in order to prevent the release of radioactive material in the event of a fuel handling accident (FHA) inside containment. These doors are not initiators of any accident. The probability of a FHA is unaffected by the operational status of these doors.

The new FHA analysis with open containment personnel air locks demonstrates that maximum offsite dose is within the acceptance limits specified in RG [Regulatory Guide] 1.195. The FHA analysis results in maximum offsite doses of 68.5 rem [roentgen equivalent man] to the thyroid and 0.2 rem to the whole body. The calculated control room dose is also within the acceptance criteria specified in GDC [General Design Criteria] 19. The analysis results in thyroid and whole body doses to the control room operator of 39.6 rem and < 0.1 rem, respectively.

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

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

Response: No.

The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change will not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a TS change that Start Printed Page 54480will allow the air lock doors to be open during core alterations and fuel movement inside containment. Open doors and penetrations do not create the possibility of a new accident. Administrative controls will be implemented to ensure the capability to close the containment in the event of a FHA.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change has the potential to increase the post-FHA dose at the Site Boundary, Low Population Zone and in the control room. However, a revised FHA analysis demonstrates that the dose consequences at both locations remains within regulatory acceptance limits and the margin of safety as defined by 10 CFR 100 and GDC 19 has not been significantly reduced. To ensure a bounding calculation, the revised FHA was performed with conservative assumptions. For example, it assumes the unfiltered release to the outside atmosphere of all airborne activity reaching the containment. Additional margin will be established through administrative procedures to require that the equipment hatch and at least one door in each air lock be closed following an evacuation of containment.

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: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Branch Chief: Evangelos C. Marinos.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: June 5, 2007.

Description of amendment request: The proposed amendments is for a new technical specification (TS) to address the operation of Engineered Safety Feature (ESF) Room Coolers required to support ESF TS equipment. This amendment includes surveillance requirements and will establish a Completion Time of 72 hours to allow adequate time to complete maintenance activities on the ESF Room Coolers and thus reduce the need for unnecessary plant shutdowns.

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 addition of Technical Specification (TS) 3.7.19 creates a Limiting Condition for Operation (LCO) for the Engineering Safety Feature (ESF) Room Coolers required to support ESF TS equipment. The Completion Time presented in the new TS is consistent with other ESF mechanical system Completion Times and is supported by the inputs used in the current analysis. The possibility of a loss of off site power (LOSP) is actually reduced by continuing power operation of the Unit. The radiological consequences of any associated accidents are not impacted by the proposed amendment.

Therefore, it is concluded that this change does not significantly increase 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 change in the methods governing normal operation of the plant. No new accident scenarios, failure mechanisms or limiting single failures are introduced as result of the proposed change. The change has no adverse effects on any safety-related system.

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 does not impact accident offsite dose, containment pressure or temperature, emergency core cooling system (ECCS) or reactor protection system (RPS) settings or any other parameter that could affect a margin of safety.

Therefore, it is concluded that this 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: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Branch Chief: Evangelos C. Marinos.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: July 17, 2007.

Description of amendment request: The proposed amendments would revise the current Joseph M. Farley Nuclear Plant, Units 1 and 2 technical specification (TS) requirement for the Plant Manager or the Operations Manager to hold a Senior Reactor Operator (SRO) license.

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

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not impact any accident initiators or analyzed events. It does not impact any assumed mitigation capability for any accident or transient event. The change does not involve the addition or removal of any equipment or any design changes to the facility. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not involve any physical modifications to plant structures, systems, or components (SSCs), or Start Printed Page 54481the manner in which SSCs are operated, maintained, modified, tested, or inspected. In addition, there is no change in the types or increases in the amounts of effluents that may be released offsite, and there is no increase in individual or cumulative occupational radiation exposure. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. The subject Operations Superintendent will be qualified to fill the Operations Manager position and have the same management authority over licensed operators as the Operations Manager.

In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. Administrative procedures will ensure that there is always an individual holding a current SRO license within Operations management. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1.

This change does not involve any physical modifications to SSCs, or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The setpoints at which protective actions are initiated are not altered by the change. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: July 17, 2007.

Description of amendment request: The proposed amendments would revise the current HNP Technical Specification requirement for the Operations Manager to hold an active or inactive Senior Reactor Opeator (SRO) license.

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

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not impact any accident initiators or analyzed events. It does not impact any assumed mitigation capability for any accident or transient event. The change does not involve the addition or removal of any equipment or any design changes to the facility. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not involve any physical modifications to plant structures, systems, or components (SSCs), or the manner in which SSCs are operated, maintained, modified, tested, or inspected. In addition, there is no change in the types or increases in the amounts of effluents that may be released offsite, and there is no increase in individual or cumulative occupational radiation exposure. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. The subject Operations Superintendent will be qualified to fill the Operations Manager position and have the same management authority over licensed operators as the Operations Manager. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. Administrative procedures will ensure that there is always an individual holding a current SRO license within Operations management. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1.

This change does not involve any physical modifications to SSCs, or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The setpoints at which protective actions are initiated are not altered by the change. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Evangelos C. Marinos.Start Printed Page 54482

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

Date of amendment request: July 17, 2007.

Description of amendment request: The proposed amendments would revise the current VEGP Technical Specification requirement for the Operation Manager to hold a Senior Reactor Operator (SRO) license.

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

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not impact any accident initiators or analyzed events. It does not impact any assumed mitigation capability for any accident or transient event. The change does not involve the addition or removal of any equipment or any design changes to the facility. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1. This change does not involve any physical modifications to plant structures, systems, or components (SSCs), or the manner in which SSCs are operated, maintained, modified, tested, or inspected. In addition, there is no change in the types or increases in the amounts of effluents that may be released offsite, and there is no increase in individual or cumulative occupational radiation exposure.

As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change to TS 5.2 revises the requirement concerning the Operations management position that must hold an SRO license. At least one Operations Superintendent or the Operations Manager will continue to maintain an SRO license. The subject Operations Superintendent will be qualified to fill the Operations Manager position and have the same management authority over licensed operators as the Operations Manager. In addition, a requirement was added that if not currently licensed, the Operations Manager shall have previously held an SRO license. Administrative procedures will ensure that there is always an individual holding a current SRO license within Operations management. The training, qualification and experience requirements for Operations management personnel will continue to satisfy the Unit Staff Qualifications as described in the applicable TS 5.3.1.

This change does not involve any physical modifications to SSCs, or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The setpoints at which protective actions are initiated are not altered by the change. As the proposed change is administrative in nature, operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: August 16, 2007.

Brief description of amendments: The proposed amendments would revise Technical Specifications (TS) 3.1.4, “Rod Group Alignment Limits,” Table 3.3.1-1, “Reactor Trip System Instrumentation,” Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” TS 3.4.10, “Pressurizer Safety Valves,” TS 3.7.1, “Main Steam Safety Valves (MSSVs),” and Table 3.7.1-1, “Operable Main Steam Safety Valves Versus Maximum Allowable Power.” The proposed change is a request to revise TSs for Comanche Peak Steam Electric Station, Units 1 and 2, to reflect cycle-specific safety analysis assumptions and results associated with the adoption of Westinghouse accident analyses methodologies.

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

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

Response: No.

The proposed changes only affect the transient and accident mitigation capability of the plant. The proposed changes to the pressurizer safety valve set pressure and as-found tolerance do not overlap with the pressurizer control system operation nor with the reactor trip setpoint. Therefore, the proposed changes do affect the probability of an accident previously evaluated.

The revised Reactor Trip System and Engineered Safety Features Actuation System setpoints have been shown, using NRC-approved analysis methodologies [the licensee's submittal for incorporating standard Westinghouse-developed analytical methods at Comanche Peak Steam Electric Station is under review by NRC], to meet all relevant event acceptance criteria. Similarly, the change to the nominal set pressure of the pressurizer safety valve, when evaluated using NRC-approved analysis methodologies, has been shown to meet the relevant event acceptance criteria. The proposed reduction to maximum allowable power level for operation in inoperable MSSVs has been previously shown to be very conservative. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes are based on analyses and evaluations performed in accordance with NRC-approved methodologies shown to be applicable [to] CPNPP [Comanche Peak Nuclear Power Plant] and to be conservatively applied to CPNPP [Comanche Peak Steam Electric Station herein referred to as CPNPP]. None of Start Printed Page 54483the proposed changes can result in plant operation outside the limits previously considered, nor allow the progression of transient or accident in a manner different that previously considered. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes are based on analyses and evaluations performed in accordance with NRC-approved methodologies shown to be applicable to CPNPP and to be conservatively applied to CPNPP. All relevant event acceptance criteria were found to be satisfied. Therefore the proposed change does not involve a reduction in a margin of safety.

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

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

NRC Branch Chief: Thomas G. Hiltz.

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

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

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

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: January 22, 2007.

Description of amendments request: The proposed amendments change the Technical Specifications related to the fuel design description and the fuel criticality methods to accommodate the transition to AREVA fuel.

Date of publication of individual notice in the Federal Register: August 29, 2007 (72 FR 49742).

Expiration dates of individual notice: September 28, 2007 (Public comments) and October 29, 2007 (Hearing requests).

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: July 10, 2007.

Brief description of amendment request: The proposed amendment would revise the values of the safety limit minimum critical power ratio in Technical Specification Section 2.1.1, “Reactor Core SLs.”

Date of publication of individual notice in Federal Register: September 5, 2007 (72 FR 50986).

Expiration date of individual notice: November 5, 2007.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(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.

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: November 3, 2005, as supplemented March 22 and July 17, 2007.

Brief description of amendments: The amendments implement the alternative source term methodology for analyzing design basis accident radiological consequences, thereby replacing the existing accident radiological source term that is described in Technical Information Document TID-14844, “Calculation of Distance Factors for Power and Test Reactor Sites.”

Date of issuance: August 29, 2007.

Effective date: This license amendment is effective as of the date of its issuance and shall be implemented within 60 days following completion of the installation and testing of the plant modifications described in the licensee's letters dated November 3, 2005,March 22 and July 17, 2007.

Amendment Nos.: 281 and 258.

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

Date of initial notice in Federal Register: January 17, 2006 (71 FR 2589) The supplements dated March 22 and July 17, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated August 29, 2007.

No significant hazards consideration comments received: No.Start Printed Page 54484

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of application for amendment: August 2, 2007 as supplemented by letters dated March 9 and May 8, 2007.

Brief description of amendment: This amendment revises Technical Specification 2.2.1 and 3/4.3.2 to modify the statistical summation error term “Z” and one of the allowable values for certain steam generator water level trip setpoints used in the Reactor Trip system and Engineered Safety Feature Actuation System instrumentation.

Date of issuance: August 31, 2007.

Effective date: 60 days from the date of issuance.

Amendment No. 126.

Facility Operating License No. NPF-63: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: February 27, 2007 (72 CR 8801). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 31, 2007. 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. No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 16, 2006, as supplemented by letter dated July 30, 2007.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to add a topical report to the analytical methods referenced in TS 5.6.5.b, “Core Operating Limits Report (COLR),” previously approved by U.S. Nuclear Regulatory Commission. The current method of performing the loss-of-coolant accident analyses was replaced by an updated method described in AREVA NP (formerly known as Framatome or Siemens) topical report, “EXEM BWR-2000 [Boiling-Water Reactor-2000] ECCS [Emergency Core Cooling System] Evaluation Model.”

Date of issuance: August 30, 2007.

Effective date: As of the date of issuance and shall be implemented prior to Cycle 15 operation.

Amendment No.: 153.

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

Date of initial notice in Federal Register: November 7, 2006 (71 FR 65141). The supplemental letter dated July 30, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 30, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 19, 2006, as supplemented June 7, 2007.

Brief description of amendments: Amendments revise Technical Specification 4.6.2.1.d. to change the frequency of air or smoke flow testing of the containment spray nozzles.

Date of Issuance: September 4, 2007.

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

Amendment Nos.: 201 and 148.

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

Date of initial notice in Federal Register: January 3, 2007 (72 FR 152). The supplement dated June 7, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 4, 2007.

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 application for amendments: April 26, 2007.

Brief description of amendments: The amendments revised the technical specifications (TSs) to add new Limiting Condition for Operation 3.0.6.

Date of issuance: September 5, 2007.

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

Amendment Nos: 235 and 230.

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

Date of initial notice in Federal Register: July 3, 2007 (72 FR 36522).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 5, 2007.

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: October 17, 2006, as supplemented by letters dated February 7, April 17, May 4, and July 26, 2007.

Brief description of amendment: The amendment revised Technical Specification (TS) 4.3.1.1.c, “Criticality,” by adding a new nominal center-to-center distance between fuel assemblies for two new storage racks, and TS 4.3.3, “Capacity,” by increasing the capacity of the spent fuel storage pool from 2366 assemblies to 2651 assemblies.

Date of issuance: September 6, 2007.

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

Amendment No.: 227.

Facility Operating License No. DPR-46: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: December 5, 2006 (71 FR 70561) and January 19, 2007 (72 FR 2560).

The supplements dated February 7, April 17, May 4, and July 26, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 6, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 4, 2007, as supplemented by letters dated April 27, 2007, May 22, 2007, and July 23, 2007.Start Printed Page 54485

Brief description of amendment: The amendment revises Technical Specification (TS) 3.7.1, “Service Water (SW) System and Ultimate Heat Sink (UHS),” as follows: revises the existing Limiting Condition for Operation (LCO) statement to require four operable SW pumps to be in operation when SW subsystem supply header water temperature is ≤82 °F; adds a requirement that five operable SW pumps be in operation when SW subsystem supply header water temperature is >82 °F and ≤84 °F; deletes Condition G and the associated Required Actions and Completion Times; revises Surveillance Requirement 3.7.1.3 to increase the maximum allowed SW subsystem supply header water temperature from 82 °F to 84 °F; and modifies the requirements for increasing the surveillance frequency as the temperature approaches the limit.

Date of issuance: September 4, 2007.

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

Amendment No.: 119.

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

Date of initial notice in Federal Register: March 13, 2007 (72 FR 11390).

The supplemental letters dated April 27, 2007, May 22, 2007, and July 23, 2007, 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 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 4, 2007.

No significant hazards consideration comments received: No

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

Date of application for amendments: November 16, 2006, as supplemented on March 29 and July 31, 2007.

Brief Description of amendments: These amendments added a reference in Technical Specification (TS) Section 6.2.C, “Core Operating Limits Report (COLR),” to permit the use of the Westinghouse Best-Estimate Large Break Loss-of-Coolant Accident (BE-LBLOCA) analysis methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis of LBLOCA.

Date of issuance: September 6, 2007.

Effective date: As of date of issuance and shall be implemented at the completion of Unit 1 fall 2007 refueling outage.

Amendment Nos.: 254 and 253.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments changed the licenses and the technical specifications.

Date of initial notice in Federal Register: December 5, 2006 (71 FR 70564). The supplements dated March 29 and July 31, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 6, 2007.

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: March 14, 2007.

Brief description of amendment: The amendment revised Surveillance Requirements 3.7.2.1 and 3.7.3.1 for the main steam isolation valves and main feedwater isolation valves, respectively, to replace the isolation times by the phrase “within limits.” The valve closure times will be stated in the TS Bases, which is controlled by TS 5.5.14, “Technical Specification (TS) Bases Control Program.” This amendment is consistent with the NRC-approved Technical Specification Task Force Traveler 491, Revision 2, “Removal of Main Steam and Main Feedwater Isolation Times.”

There are other proposed changes to the TSs in the application dated March 14, 2007, that are not being addressed in this amendment. These will be addressed in future letters to the licensee.

Date of issuance: August 28, 2007.

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

Amendment No.: 174.

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

Date of initial notice in Federal Register: June 19, 2007 (72 FR 33785).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 28, 2007.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 17th day of September, 2007.

For the Nuclear Regulatory Commission.

Catherine Haney,

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

End Signature End Preamble

[FR Doc. E7-18634 Filed 9-24-07; 8:45 am]

BILLING CODE 7590-01-P