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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 November 22, 2007, to December 5, 2007. The last biweekly notice was published on December 4, 2007 (72 FR 68206).

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, person(s) 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 via electronic submission through the NRC E-Filing system 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 person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed 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 Start Printed Page 71704fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for hearing or a petition for leave to intervene must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated in August 28, 2007, (72 FR 49139). The E-Filing process requires participants to submit and serve documents over the internet or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM is free and is available at http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html. Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html or by calling the NRC technical help line, which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, Monday through Friday. The help line number is (800) 397-4209 or locally, (301) 415-4737.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First-class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later than 11:59 p.m. Eastern Time on the due date.

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a Presiding Officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. Start Printed Page 71705

For further details with respect to this amendment 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.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No.1 (CPS), DeWitt County, Illinois

Date of amendment request: September 27, 2007.

Description of amendment request: The proposed amendment would modify technical specification (TS) by relocating references to specific American Society for Testing and Materials (ASTM) standards for fuel oil testing to licensee-controlled documents. In the referenced letter, AmerGen (the licensee) previously received approval for a change to the Unit No. 1, CPS TS that added the water and sediment content test as alternative criteria to the “clear and bright” acceptance test for new fuel oil.

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

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

Response: No.

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

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

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

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

Response: No.

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

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

Response: No.

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

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

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

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell A. Gibbs.

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

Date of amendments request: November 8, 2007.

Description of amendments request: The amendment would clarify the Technical Specification definitions for Channel Calibration and Channel Functional Test. The proposed amendments would incorporate Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-205-A, “Revision of Channel Calibration, Channel Functional Test, and Related Definitions,” Revision 3, dated July 31, 2003.

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

1. Would not involve a significant increase in the probability or consequences of any accident previously evaluated.

The proposed change clarifies the Technical Specification requirements for performance of channel calibrations and channel functional tests. Specifically, the proposed change incorporates the Nuclear Regulatory Commission-approved Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-205-A, “Revision of Channel Calibration, Channel Functional Test, and Related Definitions,” Revision 3, dated July 31, 2003. The change does not adversely affect the performance or effectiveness of required testing, as testing appropriate to the associated Surveillance Requirements will continue to be performed. The proposed change does not have a detrimental impact on the condition or performance of any plant structure, system, or component that could initiate an analyzed Start Printed Page 71706event. Therefore, the probability of an accident previously evaluated is not significantly increased.

The equipment being calibrated or tested is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected.

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

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

The scope of the proposed change is limited to the clarification of existing calibration and test requirements. As such, the proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

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

3. The proposed change will not involve a significant reduction in [a] margin of safety.

The margin of safety in this case is the verification of instrument channel operability. The proposed change clarifies requirements for the performance of channel calibrations and channel functional tests. Specifically, the proposed change incorporates the Nuclear Regulatory Commission-approved Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-205-A, “Revision of Channel Calibration, Channel Functional Test, and Related Definitions,” Revision 3, dated July 31, 2003. No changes of setpoints to plant process limits are involved. The surveillance requirements, as revised, will continue to ensure that affected equipment is tested in a manner that gives confidence that the equipment can perform its appropriate safety function.

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

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

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

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: November 9, 2007.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 3.8.a.7 related to the movement of heavy loads over and in the spent fuel pools and would relocate the modified requirements to a licensee-controlled document, the Kewaunee Power Station Technical Requirements Manual (TRM). The proposed amendment is needed to facilitate future spent fuel cask handling activities associated with dry cask spent fuel storage. The proposed amendment would incorporate the use of a single-failure-proof lifting system for handling of necessary heavy loads over or in the spent fuel pool with irradiated fuel in either the fuel storage racks or in the just-loaded spent fuel canister in the spent fuel pool. The proposed modified TS 3.8.a.7 would then be relocated to the TRM.

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 amendment revises Kewaunee Power Station (KPS) heavy load handling Technical Specification (TS) 3.8.a.7 requirements consistent with modifications to the Auxiliary Building (AB) crane and the NRC's [Nuclear Regulatory Commission] current guidance for single-failure-proof lifting systems. The proposed amendment also relocates the affected heavy load handling-related TS to a licensee-controlled document, consistent with the NRC's regulations.

The proposed change to TS 3.8.a.7 permits spent fuel cask handling in the spent fuel pool, which is required for loading spent fuel for dry storage at the on-site Independent Spent Fuel Storage Installation (ISFSI). Proposed TS 3.8.a.7 includes a new requirement that the AB crane and associated lifting devices meet the applicable single-failure-proof criteria.

Heavy load handling will continue to be conducted in accordance with the KPS heavy load handling program, which meets the NRC's guidance in NUREG-0612, as described in this LAR, and as augmented by Regulatory Information Summary 2005-25. With the upgrade of the AB crane load handling system, drops of heavy loads will not be considered credible. Notwithstanding the AB crane upgrade, heavy loads will still be prohibited from being suspended over irradiated fuel in the spent fuel pool storage racks under the revised requirements.

The previously evaluated cask drop accident is not considered credible with the upgraded AB crane because the crane trolley is being upgraded to a single-failure-proof design, consistent with applicable NRC-endorsed guidance. Lifting devices and interfacing lifting points associated with spent fuel cask handling will also be designed in accordance with applicable NRC guidance pertaining to single-failure-proof lifting systems. The result of these design upgrades is that the AB crane will retain the lifted load in the event of a single failure in the load path, including a failure of a wire rope. In addition, the crane will hold the load and the trolley and bridge will be designed to stay on their respective rails during a design basis seismic event.

The relocation of TS 3.8.a.7 to the KPS Technical Requirements Manual (TRM) is an administrative change that does not affect plant operation or heavy load handling.

Revised TS 3.8.a.7 and its associated Bases will be relocated to the TRM after approval of this amendment request. Changes to the KPS TRM are controlled by 10 CFR 50.59. Regulation 10 CFR 50.59 requires that NRC approval be obtained prior to any change that would result in more than a minimal increase in (1) the frequency of occurrence of an accident previously evaluated, (2) likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated, or (3) consequences of a malfunction of a SSC important to safety previously evaluated. Accordingly, upon relocation of the requirements of TS 3.8.a.7 and associated Bases to the TRM, appropriate control of changes will be maintained, based on the criteria in 10 CFR 50.59. Administrative relocation of the requirements of TS 3.8.a.7 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of KPS or the manner in which it is operated.

Therefore, the proposed change does not significantly increase 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.

Heavy load handling will continue to be conducted in accordance with the KPS heavy load handling program, which meets the NRC's guidance in NUREG-0612, as approved for KPS. Drops of heavy loads will continue to be very improbable events and the upgrade of the KPS AB crane lifting system to a single-failure-proof design provides additional defense-in-depth against such events. Notwithstanding the AB crane upgrade, heavy loads will still be prohibited from being suspended over irradiated fuel in the spent fuel pool storage racks under the revised requirements.

Heavy load handling operations at KPS will continue to be conducted as they currently are and no new heavy load handling operations are required as a result of this amendment. The previously evaluated cask drop accident is not considered credible with the upgraded AB crane because the crane trolley is being upgraded to a single-failure-proof design, consistent with applicable NRC-endorsed guidance. Lifting devices and interfacing lifting points associated with spent fuel cask handling will also be designed in accordance with applicable NRC guidance pertaining to single-failure-proof lifting systems. The result Start Printed Page 71707of these design upgrades is that the AB crane will retain the lifted load in the event of a single failure in the load path, including a failure of a wire rope. In addition, the crane will hold the load and the trolley and bridge will be designed to stay on their respective rails during a design basis seismic event.

The relocation of TS 3.8.a.7 to the KPS Technical Requirements Manual (TRM) is an administrative change that does not affect plant operation or heavy load handling.

Accordingly, upon relocation of the requirements of TS 3.8.a.7 and associated Bases to the TRM, appropriate control of changes will be maintained, based on the criteria in 10 CFR 50.59. Modification of the requirements of TS 3.8.a.7 does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of KPS or the manner in which it is operated.

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

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

Response: No.

The proposed amendment revises KPS heavy load handling TS 3.8.a.7 requirements consistent with modifications to the AB crane and the NRC's current guidance for single-failure-proof lifting systems.

Heavy load handling will continue to be conducted in accordance with the KPS heavy load handling program, which meets the NRC's guidance in NUREG-0612, as approved for KPS. Drops of heavy loads will continue to be very improbable events and the upgrade of the KPS AB crane lifting system to a single-failure-proof design provides additional defense-in-depth against such events and an increase in overall design margin. Notwithstanding the AB crane upgrade, heavy loads will still be prohibited from being suspended over irradiated fuel in the spent fuel pool storage racks under the revised requirements.

Further, the relocation of TS 3.8.a.7 to the KPS Technical Requirements Manual (TRM) is an administrative change that does not affect plant operation or heavy load handling.

Heavy load handling operations at KPS will continue to be conducted as they currently are and no new heavy load handling operations are required as a result of this amendment. The previously evaluated cask drop accident is less probable with the upgraded AB crane because the crane trolley is being upgraded to a single-failure-proof design, consistent with applicable NRC-endorsed guidance. Lifting devices and interfacing lifting points associated with spent fuel cask handling will also be designed in accordance with applicable NRC guidance pertaining to single-failure-proof lifting systems. The result of these design upgrades is that the AB crane will retain the lifted load in the event of a single failure in the load path, including a failure of a wire rope. In addition, the crane will hold the load and the trolley and bridge will be designed to stay on their respective rails during a design basis seismic event.

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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Acting Branch Chief: Cliff Munson.

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

Date of amendment request: November 9, 2007.

Description of amendment request: The proposed amendment would revise the Kewaunee Power Station (KPS) Updated Safety Analysis Report (USAR) to modify the design and licensing basis for the auxiliary building (AB) crane. The proposed amendment would allow the use of a methodology for performing the seismic qualification analysis of the upgraded crane. The crane is being upgraded to become a single-failure-proof design. The new methodology includes rolling of the crane bridge and trolley wheels during a seismic event.

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

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

Response: No.

This amendment request pertains solely to an analysis method supporting the upgrade of the KPS AB crane from a non-single-failure-proof design to a single-failure-proof design. The AB crane is used to lift and handle loads in the KPS spent fuel pool and truck bay areas. The AB crane does not interface with operating plant equipment. The design rated load of the AB crane remains the same as previously approved. The proposed amendment does not change the current heavy load handling practices that are in use at KPS. Upgrading the AB crane to a single-failure-proof design will reduce the probability of a heavy load drop in the areas where the AB crane lifts and handles loads.

The seismic analysis method proposed for use recognizes the inherent propensity for structures not fixed to one another (e.g., steel wheels on steel rails) to roll if sufficient lateral force is applied to either object. This seismic analysis method is proposed for use solely on the AB crane upgrade and not for any other plant structures, systems, or components. The recognition of wheel rolling between the AB crane trolley and bridge and their respective rails reflects the true nature of the installed equipment and its response to horizontal forces generated by a seismic event. Consideration of rolling reduces the projected analyzed loads on the crane and building structures and eliminates the need for unnecessary modifications to both.

Therefore, 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.

This amendment request pertains to an analysis method supporting the upgrade of an existing plant component. Specifically, the existing AB crane trolley is being replaced with a state-of-the-art design that is single-failure-proof. The AB crane does not interface with operating plant equipment. This seismic analysis method is proposed for use solely on the AB crane upgrade and not for any other plant structures, systems, or components.

The design rated load of the AB crane remains the same at 125 tons. This load controls the design and supporting analysis. The auxiliary hook design rated load is being increased from 10 tons to 15 tons. The proposed amendment does not change the currently acceptable heavy load handling practices in use at KPS. The number and types of lifts made using this crane in support of KPS plant operations are not significantly changed from that contemplated during original plant licensing. Furthermore, the basic operations of the crane (i.e., hoisting and horizontal travel) remain the same, although the electronic controls will be upgraded to current standards.

Therefore, the proposed amendment does not create a new or different kind of accident from any accident previously evaluated in the KPS licensing basis.

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

Response: No.

Although the proposed change is made specifically to support the upgrade of the KPS AB crane from a non-single-failure-proof to a single-failure-proof design, the margin of safety under consideration in this evaluation is mainly based on that contained within the safety analysis (seismic analysis).

The purpose of this methodology is to determine the stress placed on the AB cranes' structural components. The stresses determined by this methodology are then compared to the yield strength values contained in CMAA-70. If the stresses the structural component are analyzed to receive during a postulated seismic event are less than the values contained in CMAA-70 the structural integrity of the crane is maintained and a suspended load will remain suspended during a seismic event. Additional margin has been added by reducing the analysis acceptance criteria to 90% of the acceptance criteria values contained in CMAA-70, modifying the crane support structure Start Printed Page 71708through additional welds and material, and confirming the bolts are of the proper material.

DEK [Dominion Energy Kewaunee] is modeling the AB crane to roll during a seismic event when the postulated forces exceed the brake holding force. This provides a more realistic approach because the crane trolley is not fixed to the bridge rails. DEK has provided additional conservatisms by doubling the calculated force needed to overcome the brake holding force.

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

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Acting Branch Chief: Cliff Munson.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: October 22, 2007.

Description of amendment request: The proposed amendment would modify the Technical Specification (TS) to establish more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope (CRE) in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard Technical Specification change traveler TSTF-448, Revision 3, “Control Room Habitability.” Specifically, the proposed amendment would modify TS 3.7.9, “Control Room Emergency Ventilation System (CREVS),” and would establish a CRE habitability (CREH) program in TS Section 5.5, “Administrative Controls—Programs and Manuals.” The NRC staff issued a “Notice of Availability of Technical Specification Improvement to Modify Requirements Regarding Control Room Envelope Habitability Using the Consolidated Line Item Improvement Process” associated with TSTF-448, Revision 3, in the Federal Register on January 17, 2007 (72 FR 2022). The notice included a model safety evaluation, a model no significant hazards consideration (NSHC) determination, and a model license amendment request. In its application dated October 22, 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

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 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: 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 Nos. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: October 22, 2007.

Description of amendment request: The proposed amendment would modify requirements of Technical Specification (TS) 3.4.12, “RCS Specific Activity,” and TS 3.7.4, “Secondary Specific Activity,” as related to the use of an alternate source term (AST) associated with accident offsite and control room dose consequences. Implementation of AST supports adoption of the control room envelope habitability controls in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard Technical Specification change traveler TSTF-448, Revision 3, “Control Room Habitability.”

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. Start Printed Page 71709

The use of an AST is recognized in 10 CFR 50.67 and guidance for its implementation is provided in RG [Regulatory Guide] 1.183. The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. As such, the AST cannot affect the probability of occurrence of a previously evaluated accident. In addition, the reduction is specific activity limits within the TSs is unrelated to accident initiators. No facility equipment, procedure, or process changes are required in conjunction with implementing the AST that could increase the likelihood of a previously analyzed accident. The proposed changes in the source term and the methodology for the dose consequence analyses follow the guidance of RG 1.183. As a result, there is no increase in the likelihood of existing event initiators.

Regarding accident consequences, the reduction in specific activity limits within the TSs is more restrictive (more conservative) and acts to support the analysis results given the application of an AST. The results of accident dose analyses using the AST are compared to TEDE [total effective dose equivalent] acceptance criteria that account for the sum of deep dose equivalent (for external exposure) and committed effective dose equivalent (for internal exposure). Dose results were previously compared to separate limits on whole body, thyroid, and skin doses as appropriate for the particular accident analyzed. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. Implementing the AST involves no facility equipment, procedure, or process changes that could affect the radioactive material actually released during an event. Consequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any of the events being 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 AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. As such, the AST cannot create the possibility of a new or different kind of accident. In addition, the reduction is specific activity limits within the TSs is unrelated to accident initiators. No facility equipment, procedure, or process changes have been made in conjunction with implementing the AST that could initiate or substantially alter the progression of an accident.

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

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

Response: No.

Implementing the AST is relevant only to calculated accident dose consequences. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. In addition, the reduction is specific activity limits within the TSs is unrelated to accident initiators. No facility equipment, procedure, or process changes are required in conjunction with implementing the AST that could increase the exposure of control room or offsite individuals to radioactive material. The AST does not affect the transient behavior of non-radiological parameters (e.g., Reactor Coolant System pressure, Containment pressure) that are pertinent to a margin of safety.

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

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

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

NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: October 22, 2007.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) requirements for mode change limitations in Limiting Condition for Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 3.0.4. The proposed TS changes are consistent with Revision 9 of Nuclear Regulatory Commission (NRC)-approved Industry TS Task Force (TSTF) Standard TS (STS) change traveler, TSTF-359, “Increase Flexibility in Mode Restraints.” The amendment would also modify other TSs to reflect the revisions to LCO 3.0.4. The spelling of the word “not” is corrected in Section 1.4 of the TSs.

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), as part of the Consolidated Line Item Improvement Process (CLIIP), on possible amendments to revise the plant-specific TS to modify requirements for model change limitations in LCO 3.0.4 and SR 3.0.4.

The NRC staff subsequently issued a notice of availability of the models for Safety Evaluation and No Significant Hazards Consideration Determination for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the CLIIP, including the model No Significant Hazards Consideration Determination, in its application dated October 22, 2007.

The proposed TS changes are consistent with NRC-approved Industry TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579. TSTF-359, Revision 8, was subsequently revised to incorporate the modifications discussed in the April 4, 2003, Federal Register notice and other minor changes. TSTF-359, Revision 9, was subsequently submitted to the NRC on April 28, 2003, and was approved by the NRC on May 9, 2003.

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

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, 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 Start Printed Page 71710accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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 Nos. 50-368, Arkansas Nuclear One, Unit 2*, Pope County, Arkansas

Date of amendment request: October 22, 2007.

Description of amendment request: The proposed amendment would modify the Technical Specification (TS) to establish more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope (CRE) in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard Technical Specification change traveler TSTF-448, Revision 3, “Control Room Habitability.” Specifically, the proposed amendment would modify TS 3.7.6.1, “Control Room Emergency Ventilation and Air Condition System,” and would establish a CRE habitability (CREH) program in TS Section 6.5, “Administrative Controls—Programs and Manuals.” The NRC staff issued a “Notice of Availability of Technical Specification Improvement to Modify Requirements Regarding Control Room Envelope Habitability Using the Consolidated Line Item Improvement Process” associated with TSTF-448, Revision 3, in the Federal Register on January 17, 2007 (72 FR 2022). The notice included a model safety evaluation, a model no significant hazards consideration (NSHC) determination, and a model license amendment request. In its application dated October 22, 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

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 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: 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 Nos. 50-368, Arkansas Nuclear One, Unit 2, Pope County, Arkansas

Date of amendment request: October 22, 2007.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) requirements for mode change limitations in Limiting Condition for Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 4.0.4. The proposed TS changes are consistent with Revision Start Printed Page 717119 of Nuclear Regulatory Commission (NRC)-approved Industry TS Task Force (TSTF) Standard TS (STS) change traveler, TSTF-359, “Increase Flexibility in Mode Restraints.” The amendment would also modify other TSs to reflect the revisions to LCO 3.0.4. In addition, a change to TS 3.4.3 was made which was determined to be equivalent to the TSTF-359 changes.

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), as part of the Consolidated Line Item Improvement Process (CLIIP), on possible amendments to revise the plant-specific TS to modify requirements for model change limitations in LCO 3.0.4 and SR 4.0.4.

The NRC staff subsequently issued a notice of availability of the models for Safety Evaluation and No Significant Hazards Consideration Determination for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the CLIIP, including the model No Significant Hazards Consideration Determination, in its application dated October 22, 2007.

The proposed TS changes are consistent with NRC-approved Industry TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579. TSTF-359, Revision 8, was subsequently revised to incorporate the modifications discussed in the April 4, 2003, Federal Register notice and other minor changes. TSTF-359, Revision 9, was subsequently submitted to the NRC on April 28, 2003, and was approved by the NRC on May 9, 2003.

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

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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 Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 18, 2007.

Description of amendment request: The proposed amendment would revise the Technical Specifications to change requirements related to Emergency Diesel Generator (EDG) fuel oil tank volume, EDG fuel oil testing and Reactor Building crane inspections.

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 changes do not impact the operability of any Structure, System or Component that affects the probability of an accident or that supports mitigation of an accident previously evaluated. The proposed change does not affect reactor operations or accident analysis and has no radiological consequences. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. 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 specified margin for onsite fuel oil storage is maintained and the applicable testing standards and methods remain unchanged. These changes do not change any existing requirements, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Start Printed Page 71712

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. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 18, 2007.

Description of amendment request: The proposed amendment would revise the Technical Specifications applicability requirements related to primary containment oxygen concentration and drywell-to-suppression chamber differential pressure limits. The associated actions would also be revised to be consistent with exiting the applicability for each specification.

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

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

Response: No. The proposed change does not increase the probability of an accident since it does not involve the modification of any plant equipment or affect how plant systems or components are operated, it only changes the requirements for when inerting and differential pressure need to be established. Whether the containment is inerted or differential pressure is established does not impact the likelihood of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. The technical limits (i.e., oxygen concentration and differential pressure) imposed by the associated Technical Specifications remain unchanged. Brief periods where the requirements for maintaining these technical limits are relaxed are currently considered in the Technical Specifications and associated licensing basis. The proposed change clarifies the definition of these periods however, any changes are not considered significant and are supported by remaining consistent with the recommended allowances of NUREG 1433, Revision 3. The consequences of analyzed events are therefore not affected. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. 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 involve the modification of any plant equipment or affect basic plant operation. Additionally, the associated limitations remain unchanged. These changes do not negate any existing requirement, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change.

The revised plant conditions reflecting the applicability and the duration allowed to restore limits are not credited in any design basis event. These changes do not reflect any significant adverse impact to the overall risk of operating during brief periods without the required primary containment oxygen concentration or differential pressure since the total time for any occurrence is only marginally extended and reflects times consistent with NUREG-1433, Revision 3. 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: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: November 20, 2007.

Description of amendment request: The proposed amendment would revise the values of the safety limit minimum critical power ratio (SLMCPR) in Technical Specification (TS) Section 2.1.1, “Reactor Core SLs.” Specifically, the proposed change would delete the Quad Cities Nuclear Power Station (QCNPS) Unit 2 fuel-specific SLMCPR requirements for Global Nuclear Fuel (GNF) GE14 fuel and consolidate the Unit 1 and Unit 2 SLMCPR requirements into a bounding dual-unit requirement. This change is needed to support the next cycle of Unit 2 operation.

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

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. 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 delete the QCNPS Unit 2 fuel-specific SLMCPR requirements for Global Nuclear Fuel (GNF) GE14 fuel conservatively establishes the SLMCPR for QCNPS, Unit 2, Cycle 20 at the SLMCPR value for the co-resident Westinghouse SVEA-96 Optima2 fuel, such that the fuel is protected during normal operation and during plant transients or anticipated operational occurrences (AOOs).

The proposed change to delete the GE14 SLMCPR and establish the requirement at the SLMCPR value for the co-resident Westinghouse SVEA-96 Optimal fuel does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The proposed change to delete the GE14 SLMCPR and establish the requirement at the SLMCPR value for the co-resident Westinghouse SVEA-96 Optimal fuel revises the QCNPS Unit 2 SLMCPR requirement to protect the fuel during normal operation as well as during plant transients or AOOs. Operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated. This will ensure that the fuel design safety criterion (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs) is met. Since the proposed change does not affect operability of plant systems designed to mitigate any consequences of accidents, the consequences Start Printed Page 71713of an accident previously evaluated will not increase.

The proposed consolidation of the Unit 1 and Unit 2 SLMCPR requirements into a bounding dual-unit requirement is administrative. As such, the proposed consolidation does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed changes do 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?

Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The proposed changes do not involve any plant configuration modifications or changes to allowable modes of operation. The proposed change to delete the GE14 SLMCPR and establish the requirement at the SLMCPR value for the co-resident Westinghouse SVEA-96 Optimal fuel assures that safety criteria are maintained for QCNPS, Unit 2, Cycle 20. The proposed consolidation of the Unit 1 and Unit 2 SLMCPR requirements into a bounding dual-unit requirement is administrative.

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?

The SLMCPR provides a margin of safety by ensuring that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs if the SLMCPR limit is not violated. The proposed change will ensure the current level of fuel protection is maintained by continuing to ensure that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs if the SLMCPR limit is not violated. The proposed SLMCPR values were developed using NRC-approved methods. Additionally, operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated. This will ensure that the fuel design safety criterion (i.e., that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limit is not violated) is met.

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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: October 29, 2007.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2 Surveillance Requirement (SR) 3.8.1.9, to require that the test is performed at or below a power factor of 0.85. The proposed amendments fulfill the commitment made in Amendments 178 to Unit 1, and 168 to Unit 2, issued on May 30, 2007 (Agency wide Documents Access and Management System (ADAMS) Accession No. ML071310023).

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

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

Response: No.

This license amendment request proposes more restrictive changes to the Technical Specification Surveillance Requirements for the emergency diesel generators which will require testing at a specified power factor, grid conditions permitting.

The emergency diesel generators are not accident initiators and therefore, these changes do not involve a significant increase in the probability of an accident. The proposed changes increase the load testing requirements, are consistent with the intent of current regulatory guidance for testing emergency diesel generators, and will continue to assure that this equipment performs its design function. Thus these changes do not involve a significant increase in the consequences of an accident.

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

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

Response: No.

This license amendment request proposes more restrictive changes to the Technical Specification Surveillance Requirements for the emergency diesel generators which will require testing at a specified power factor, grid conditions permitting.

The changes proposed for the emergency diesel generators do not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional. These changes do not create new failure modes or mechanisms which are not identifiable during testing and no new accident precursors are generated.

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

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

Response: No.

This license amendment request proposes more restrictive changes to the Technical Specification Surveillance Requirements for the emergency diesel generators which will require testing at a specified power factor, grid conditions permitting.

The current Technical Specification Surveillance Requirements do not specify testing at any power factor. The Technical Specification Surveillance Requirements proposed in this license amendment request are thus more restrictive in that they place additional restraints on the test conditions. These changes may make the testing more rigorous and thus more difficult for the emergency diesel generators to meet the test acceptance criteria. The addition of a power factor is consistent with the intent of current regulatory guidance for testing emergency diesel generators. Since these changes are an increase in the test requirements and are consistent with the intent of current regulatory guidance, these changes do not involve a significant reduction in a margin of safety.

Therefore, 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 requests involve no significant hazards consideration.

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Acting Branch Chief: Cliff Munson.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: November 19, 2007.

Description of amendment request: The proposed amendments would revise Technical Specifications for the PINGP, Units 1 and 2, to replace the Start Printed Page 71714current fixed Frequency for testing the containment spray nozzles in Surveillance Requirement 3.6.5.8 with a maintenance or event based Frequency.

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

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

Response: No.

This license amendment request proposes Technical Specification Surveillance Requirement changes which will require verification that the containment spray system spray nozzles are unobstructed following maintenance which could result in nozzle blockage.

The containment spray system and its spray nozzles are not accident initiators and therefore, these changes do not involve a significant increase the probability of an accident. The revised surveillance requirement will require event based verification in lieu of fixed Frequency verification which may require either fewer or more verifications of operability. The proposed changes to verify system operability following maintenance is considered adequate to ensure operability of the containment spray system. Since the system continues to be available to perform its accident mitigation function, the consequences of accidents previously evaluated are not significantly increased.

Therefore, the proposed changes do 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.

This license amendment request proposes Technical Specification Surveillance Requirement changes which will require verification that the containment spray system spray nozzles are unobstructed following maintenance which could result in nozzle blockage.

The proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant. The change does not introduce new accident initiators or impact the assumption made in the safety analysis. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional.

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

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

Response: No.

This license amendment request proposes Technical Specification Surveillance Requirement changes which will require verification that the containment spray system spray nozzles are unobstructed following maintenance which could result in nozzle blockage.

The containment spray system is not susceptible to corrosion-induced obstruction or obstruction from sources external to the system. Maintenance activities that could introduce foreign material into the system would require subsequent verification to ensure there is no spray nozzle blockage. The spray header nozzles are expected to remain unblocked and available in the event that the safety function is required. Therefore, the capacity of the system would remain unaffected.

Therefore, 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 requests involve no significant hazards consideration.

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Acting Branch Chief: Clifford G. Munson.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power Plant (HBPP), Unit 3 Humboldt County, California J00336

Date of amendment request: November 5, 2007.

Description of amendment request: The licensee has proposed amending the technical specifications (TS) to delete many operational and administrative requirements upon transfer of spent nuclear fuel assemblies and fuel fragment containers from the Spent Fuel Pool (SFP) to the Humboldt Bay Independent Spent Fuel Storage Installation (ISFSI). Some TS requirements will be relocated to the HBPP Quality Assurance Plan.

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

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

Response: No.

The proposed changes reflect the transfer of spent fuel from the Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel Storage Installation. Design basis accidents related to the SFP are discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety Analysis Report (DSAR). These postulated accidents are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no important-to-safety systems, structures or components required to function or to be monitored. In addition, there are no remaining credible accidents involving spent fuel or the SFP that require actions of a Certified Fuel Handler or Noncertified Fuel Handler to prevent occurrence or to mitigate consequences. The proposed change to the Design Features section of the Technical Specifications (TS) clarifies that the spent fuel is being stored in dry casks within an ISFSI. The probability or consequences of accidents at the ISFSI are evaluated in the HB ISFSI Final Safety Analysis Report (FSAR) and are independent of the accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed changes will not involve a significant increase in 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 evaluated?

Response: No.

The proposed changes reflect the reduced operational risks as a result of the spent fuel being transferred to dry casks within an ISFSI. The proposed changes do not modify any systems, structures or components. The plant conditions for which the HBPP Unit 3 DSAR design basis accidents relating to spent fuel and the SFP have been evaluated are no longer applicable. The aforementioned proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of an accident. Design basis accidents associated with the dry cask storage of spent fuel are already considered in the HB ISFSI FSAR. No new accident scenarios are created as a result of deleting nonapplicable operational and administrative requirements. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from those previously evaluated.

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

Response: No.

The proposed changes reflect the reduced operational risks as a result of the spent fuel being transferred to dry casks within an ISFSI. The design basis and accident assumptions within the HBPP Unit 3 DSAR and the TS relating to spent fuel are no longer applicable. The proposed changes do not affect remaining plant operations, nor structures, systems, or components supporting decommissioning activities. In addition, the proposed changes do not result in a change in initial conditions, system response time, or in any other parameter affecting the course of a decommissioning activity accident analysis. Therefore, the proposed changes will not involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the Start Printed Page 71715amendment request involves no significant hazards consideration.

Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and Electric Company, 77 Beale Street, B30A, San Francisco, CA.

NRC Branch Chief: Andrew Persinko.

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

Date of amendment request: November 30, 2007.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) Sections TS 5.5.9, “Steam Generator (SG) Program” and TS 5.6.10, “Steam Generator Tube Inspection Report.” The proposed changes to TS 5.5.9 modify the inspection and plugging requirements for portions of SG tubes within the hot leg side of the tubesheet region of the SGs only. The proposed changes to TS 5.6.10 will add requirements to report specific data related to indications, leakage detected, and calculated accident leakage.

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

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

No. The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed changes that alter the SG inspection criteria do not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed changes will not alter the operation of, or otherwise increase the failure probability of, any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed changes to the SG tube inspection criteria, are the SG tube rupture (SGTR) event and the steam line break (SLB) accident.

During the SGTR event, the required structural integrity margins of the SG tubes will be maintained by the presence of the SG tubesheet. SG tubes are hydraulically expanded in the tubesheet area. Tube rupture in tubes with cracks in the tubesheet is precluded by the constraint provided by the tubesheet. This constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet and from the differential pressure between the primary and secondary side. Based on this design, the structural margins against burst discussed in RG 1.121 (Reference 4) [Regulatory Guide 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes,” dated August 1976], are maintained for both normal and postulated accident conditions.

The proposed changes do not affect other systems, structures, components or operational features. Therefore, the proposed changes result in no significant increase in the probability of the occurrence of a SGTR accident.

At normal operating pressures, leakage from primary water stress corrosion cracking (PWSCC) below the proposed limited inspection depth is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region. The consequences of a SGTR event are affected by the primary-to-secondary leakage flow during the event. Primary-to-secondary leakage flow through a postulated broken tube is not affected by the proposed change since the tubesheet enhances the tube integrity in the region of the hydraulic expansion by precluding tube deformation beyond its initial hydraulically expanded outside diameter.

The probability of a SLB is unaffected by the potential failure of a SG tube, since this failure is not an initiator for a SLB.

The consequences of a SLB are also not significantly affected by the proposed changes. During a SLB accident, the reduction in pressure above the tubesheet on the shell side of the SG creates an axially uniformly distributed load on the tubesheet due to the reactor coolant system pressure on the underside of the tubesheet. The resulting bending action constrains the tubes in the tubesheet, thereby restricting primary-to-secondary leakage below the midplane.

The purpose of the tube-end weld is to ensure the hydraulically expanded tube-to-tubesheet joints in Model F SGs are leak-tight. Considerations were also made with regard to the potential for primary-to-secondary leakage during postulated faulted conditions. However, the leak rate during postulated accident conditions would be expected to be less than that during normal operation for indications near the bottom of the tubesheet based on the evaluation (Reference 1) [Westinghouse Electric Company WCAP-16794-P, “Steam Generator Tube Alternate Repair Criteria for the Portion of the Tube Within the Tubesheet at the Vogtle 1 & 2 Electric Generating Plants,” dated October 2007] which shows that while the driving pressure increases by about a factor of almost two, the flow resistance increases, because the tube-to-tubesheet contact pressure also increases. Depending on the depth within the tubesheet, the relative increase in resistance could easily be larger than that of the pressure potential. Therefore, the leak rate under normal operating conditions could exceed its allowed value before the accident condition leak rate would be expected to exceed its allowed value. This approach is termed an application of the “bellwether principle.” While such a decrease in the leak rate is expected, the postulated accident leak rate could conservatively be taken to be bounded by twice the normal operating leak rate if the increase in contact pressure is ignored.

Since normal operating leakage is limited by VEGP TS 3.4.13 and by NEI 97-06 (Reference 3) [NEI 97-06, “Steam Generator Program Guidelines,” Revision 2, dated May 2, 2005] to less than 150 gpd throughout one SG in the VEGP Units 1 and 2 SGs, the attendant accident condition leak rate, assuming all leakage to be from lower tubesheet indications, would be bounded by 0.20 gpm in the faulted SG which is less than the accident analysis assumption of 0.35 gpm to the affected SG included in Section 15.1.5 of the VEGP FSAR. Hence, it is reasonable to omit any consideration of inspection of the tube, tube end weld, bulges/overexpansions or other anomalies below 17 inches from the top of the hot leg tubesheet.

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

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

No. The proposed changes do not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases.

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

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

No. The proposed changes maintain the required structural margins of the SG tubes for both normal and accident conditions. NEI 97-06 (Reference 3) and RG 1.121 (Reference 4), are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. RG 1.121 (Reference 4) describes a method acceptable to the NRC for meeting the following General Design Criteria (GDC).

  • GDC 14, “Reactor coolant pressure boundary,”
  • GDC 15, “Reactor coolant system design,”
  • GDC 31, “Fracture prevention of reactor coolant pressure boundary,” and,
  • GDC 32, “Inspection of reactor coolant pressure boundary.”

RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME Code [American Society of Mechanical Engineers, Boiler and Pressure Vessel Code].

Application of the limited tubesheet inspection depth criteria will preclude Start Printed Page 71716unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the proposed limited tubesheet inspection depth criteria.

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, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Branch Chief: Evangelos C. Marinos.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: October 23, 2007.

Description of amendment request: The amendments will relocate the surveillance test intervals of various Technical Specifications (TSs) to a licensee-controlled program (risk-informed Initiative 5(b)) in accordance with the Surveillance Frequency Control Program, which is being added to the Administrative Controls section of the TS.

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

Response: No.

The proposed change[s] [involve] the relocation of various surveillance test intervals from Technical Specifications (TS) to a licensee-controlled program. The proposed change[s] [do] not involve the modification of any plant equipment or affect basic plant operation. The proposed change[s] will have no impact on the design or function of any safety related structures, systems or components. Surveillance test intervals are not assumed to be an initiator of any analyzed event, nor are they assumed in the mitigation of consequences of accidents. The surveillance requirements themselves will be maintained in the TS along with the applicable Limiting Conditions for Operation (LCOs) and Action statements. The surveillances performed at the intervals specified in the licensee-controlled program will assure that the affected system or component function is maintained, that the facility operation is within the Safety Limits, and that the LCOs are met.

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

2. [Do] the proposed change[s] create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change[s] [do] not involve any physical alteration of plant equipment and does not change the method by which any safety-related structure, system, or component performs its function or is tested. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged.

The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The proposed change[s] [do] not negate any existing requirement, and [do] not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by relocation of the surveillance test intervals to a licensee-controlled program.

Therefore, the proposed change[s] [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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Branch Chief: Thomas G. Hiltz.

U.S. Department of Transportation (USDOT), United States Maritime Administration (MARAD), License No. NS-1, Docket No. 50-238, Nuclear Ship Savannah (NSS)

Date of amendment request: October 9, 2007.

Description of amendment request:

The proposed license amendment would modify the Technical Specification (TS) requirements to clarify the TS and make the requirements commensurate with the current ship status and decommissioning schedule. Thirty-nine TS changes are proposed. The proposed changes modify the TS as follows:

  • Delete requirements more appropriate for the Final Safety Analysis Report;
  • Provide consistent titles and phrases;
  • Delete duplicate requirements;
  • Organize similar requirements into single locations;
  • Remove requirements that can be implemented through current regulations;
  • Delete archaic requirements;
  • Invoke requirements commensurate with current ship status and decommissioning schedule;
  • Format and renumber, as appropriate;
  • Revise requirements to reflect historical practices;
  • Revise TS to be consistent with the Decommissioning Quality Assurance Plan; and
  • Correct errors introduced in License Amendment 13, Reference (a).

The application for license amendment is available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/​reading-rm/​adams.html. From this site, you can access the NRC's Agencywide Document Access and Management System (ADAMS), which provides text and image files of NRC's public documents. The ADAMS accession number for the October 9, 2007, request is ML072880143.

If you do not have access to ADAMS, or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. These documents may also be viewed electronically on the public computers located at the NRC's PDR, O 1 F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction contractor will copy documents for a fee.

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

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

Response: No.

The proposed changes are administrative and do not involve modification of any plant equipment or affect basic plant operation. The NSS's reactor is not operational and the level of radioactivity in the NSS has significantly decreased from the levels that Start Printed Page 71717existed when the 1976 Possession-only License was issued. No aspect of any of proposed changes is and initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased.

Therefore, the proposed changes no not involve a significant increase in 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 evaluated?

Response: No.

All of the proposed changes are administrative and do not involve physical alteration of plant equipment that was not previously allowed by Technical Specifications. These proposed changes do not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

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

Response: No.

All of the proposed changes are administrative in nature. No margins of safety exist that are relevant to the ship's defueled and partially dismantled reactor. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed changes. The proposed changes involve movement of the ship, changes in the performance of responsibilities and reflect significantly improved radiological conditions since 1976.

Therefore, 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 upon the staff's review of the licensee's analysis, as well as the staff's own evaluation, the staff concludes 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.

Senior Technical Advisor, N.S. Savannah: Erhard W. Koehler, MARAD, Office of Ship Disposal Programs.

NRC Branch Chief: Andrew Persinko.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendment request: October 24, 2007.

Description of amendment request: The amendments would revise the Technical Specifications (TS) Limiting Condition for Operations (LCO) 3.8.7 and 3.8.9, pertaining to electrical power systems and distribution associated with the 120 Volt AC vital bus inverters. The TS changes are intended to support operability of components shared between Unit 1 and Unit 2. The proposed changes will add new Conditions, Required Action statements and Completion Times for LCO 3.8.7 and LCO 3.8.9 to address shared components.

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?

The proposed amendment does not involve a significant increase in the probability or consequence of an accident previously analyzed. There is no change to how or under what conditions the inverters or 120 VAC vital buses are operated, nor are there any changes to acceptable operating parameters. Operability requirements, which are consistent with current operation of the inverters and vital buses, are being established for the inverters and vital buses associated with shared systems. The proposed change will ensure that there is an operable electrical control circuit for the Auxiliary Building Central Exhaust subsystem filter and bypass dampers for each train of the [Emergency Core Cooling System Pump Room Exhaust Air Cleanup System] ECCS PREACS which will ensure that the evaluated dose consequences for [design basis accidents] DBAs will not be exceeded.

Therefore, the proposed changes do 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?

The implementation of the proposed changes does not create the possibility of an accident of a different type than was previously evaluated in the [Updated Final Safety Analysis Report] UFSAR. There is no change to how or under what conditions the inverters or 120 VAC vital buses are operated nor are there any changes to acceptable operating parameters. The proposed operability requirements, which are consistent with current operation of the inverters and vital buses, are being established for the inverters and vital buses associated with shared systems. The proposed changes ensure vital 120 VAC power is available to support operation of the Auxiliary Building Central Exhaust subsystems. These changes do not alter the nature of events postulated in the UFSAR nor do they introduce any unique precursor mechanisms.

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

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

The implementation of the proposed changes does not reduce the margin of safety. The proposed changes for the 120 VAC Vital Bus System and Inverters do not affect the ability of these systems or components to perform their intended safety functions to provide power to required safety and monitoring systems or components. Operability requirements, which are consistent with current operation of the inverters and vital buses, are being established for the inverters and vital buses associated with shared systems. These changes provide additional assurance that the Auxiliary Building Central Exhaust subsystems will operate to maintain the margin of safety.

Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Evangelos C. Marinos.

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 Start Printed Page 71718with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of application for amendments: January 22, 2007, as supplemented by letter dated September 28, 2007.

Brief Description of amendments: The amendments change the Technical Specifications (TSs) related to the fuel design description and the fuel criticality methods to accommodate the transition to AREVA NP fuel.

Date of issuance: November 27, 2007.

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

Amendment Nos.: 243 and 271.

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

Date of initial notice in Federal Register: August 29, 2007 (72 FR 49742). The supplement dated September 28, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 27, 2007.

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: August 17, 2007.

Brief description of amendment: The amendment revised the date for performing the “Type A test” in the River Bend Station, Unit 1, Technical Specification 5.5.13, “Primary Containment Leak Rate Testing Program,” from “prior to December 14, 2007,” to “prior to April 14, 2008.”

Date of issuance: December 3, 2007.

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

Amendment No.: 155.

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

Date of initial notice in Federal Register: September 11, 2007 (72 FR 51857). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 3, 2007.

No significant hazards consideration comments received: No.

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

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

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois.

EGC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3 (PBAPS), York and Lancaster Counties, Pennsylvania.

Date of application for amendments: December 15, 2006.

Brief description of amendments: The amendments modify the technical specifications (TSs) by replacing the term “plant-specific” with “generic” when discussing job titles in TS Section 5.2.1.a. This revision will ensure the TS description is consistent with the licensee Quality Assurance Topical Report (QATR). The proposed amendment will also revise the PBAPS TS Section 5.2.1.a to replace the reference to the Updated Final Safety Analysis Report with reference to the EGC QATR. This change aligns the PBAPS TS wording with the rest of the licensee fleet.

Date of issuance: November 19, 2007.

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

Amendment Nos.: 152, 152, 147, 147, 225, 217, 187, 174, 265, 269, 236, and 231.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, DPR-30, DRP-44, and DPR-56: The amendments revised the Technical Specifications and Operating Licenses.

Date of initial notice in Federal Register: March 13, 2007 (72 FR 11387).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 19, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 19, 2006.

Brief description of amendments: Amendments revise the requirements in Technical Specification (TS) 5.5.8, “Inservice Testing Program,” to update references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves, and address the applicability of Surveillance Requirement 3.0.2 to other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program.

Date of issuance: December 4, 2007.

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

Amendment Nos.: Unit 1-140; Unit 2-140.

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

Date of initial notice in Federal Register: May 22, 2007 (72 FR 28724). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 4, 2007.

No significant hazards consideration comments received: No. Start Printed Page 71719

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

Date of amendment request: August 16, 2007, as supplemented by letter dated November 5, 2007.

Brief description of amendment: The amendment revised Technical Specification 5.5.6, “Inservice Testing Program,” to allow a one-time extension of the 5-year frequency requirement for setpoint testing of safety valve MS-RV-70ARV.

Date of issuance: December 4, 2007.

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

Amendment No.: 228.

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

Date of initial notice in Federal Register: September 25, 2007 (72 FR 54476). The supplement dated November 5, 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 initially published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 4, 2007.

No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco Nuclear Generating Station, Sacramento County, California

Date of application for amendment: April 12, 2006, and supplemented November 21, 2006.

Brief description of amendment: The amendment incorporates the Nuclear Regulatory Commission (NRC) approved, License Termination Plan (LTP), and associated addendum, into the Rancho Seco license and specifies limits on the changes the licensee is allowed to make to the approved LTP without prior NRC review and approval.

Date of issuance: November 26, 2007.

Effective date: November 26, 2007.

Amendment No: 133.

Facility Operating License No. DPR-54: The amendment revised the License.

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6789).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 26, 2007.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of application for amendments: July 14, 2006, as supplemented by letters dated June 28, September 26, and November 2, 2007.

Brief description of amendments: The amendments incorporate a description of the parent tube inspection limitation adjacent to the nickel band portion of the lower sleeve joint and provide the basis for the structural and leakage integrity of the joint being ensured with the existing inspection of the parent tube adjacent to the nickel band region.

Date of issuance: November 29, 2007.

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

Amendment Nos.: Unit 2-215; Unit 3-207.

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

Date of initial notice in Federal Register: September 12, 2006 (71 FR 53720). The supplements dated June 28, September 26, and November 2, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 29, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 5, 2007, as supplemented June 11, 2007.

Brief description of amendments: The amendments revised the Technical Specifications testing frequency for surveillance requirement 3.1.4, “Control Rod Scram Times,” from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in MODE 1.”

Date of issuance: November 26, 2007.

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

Amendment Nos.: 254, 198.

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

Date of initial notice in Federal Register: July 17, 2007, (72 FR 39084).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 26, 2007.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 10th day of December 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. E7-24284 Filed 12-17-07; 8:45 am]

BILLING CODE 7590-01-P