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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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This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

Start Preamble

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 to 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 January 31 to February 13, 2008. The last biweekly notice was published on February 12, 2008 (73 FR 8068).

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.

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-Start Printed Page 10294day 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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

A request for 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. Start Printed Page 10295Information 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.

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.

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

Date of amendments request: August 13, 2007.

Description of amendments request: The amendment would revise Technical Specification (TS) Table 3.3.1.2-1, “Source Range Monitor [SRM] Instrumentation,” to add a note that specifies the required locations of SRMs in Mode 5 during core alterations, and also to make an administrative correction to Unit 1 TS Surveillance Requirement (SR) 3.3.1.2.2.

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

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

Response: No.

The proposed changes are administrative in nature. There are no requirements being added, deleted, or altered as a result of either of the proposed changes.

The change to Table 3.3.1.2-1 adds a footnote to Table 3.3.1.2-1 which duplicates the Mode 5 operable SRM location requirements currently specified in SR 3.3.1.2.2 and discussed in the LCO [limiting condition for operation] bases section for TS 3.3.1.2. The specific Mode 5 operable SRM location requirements are not being changed and are consistent with the requirements provided in the current version of NUREG-1433. This change is being done as an aid to Operations personnel, to help prevent inadvertently missing the requirements.

The change to SR 3.3.1.2.2 for Unit 1 corrects a typographical error to be consistent with other locations within the Unit 1 and Unit 2 TSs as well as the current version of NUREG 1433.

The proposed changes do not involve a physical change to the SRMs, nor do they alter the assumptions of the accident analyses. Therefore, the probability and the consequences of an accident previously evaluated are not affected.

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

Response: No.

The proposed changes do not involve a physical change to the SRMs, nor do they alter the assumptions of the accident analyses. The changes are purely administrative in nature. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes are administrative in nature, being done as an aid to Operations personnel, to help prevent inadvertently missing the Mode 5 operable SRM location requirements and to correct a typographical error. There are no requirements being added, deleted, or altered as a result of either Start Printed Page 10296of the proposed changes. As such, the proposed changes do not involve a reduction in a margin of safety.

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Thomas H. Boyce.

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

Date of amendment request: January 15, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) Surveillance Requirement (SR) frequency in TS 3.1.3, “Control Rod OPERABILITY” from “7 days after the control rod is withdrawn and THERMAL POWER is greater than the [Low Power Setpoint] LPSP of [Rod Worth Minimizer] RWM” to “31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM” and revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The proposed amendment does not adopt the clarification of Source Range Monitor (SRM) TS action for inserting control rods, which is applicable only to Boiling Water Reactor (BWR)/6 plants. Since Fermi 2 is a BWR/4 plant, this change in TSTF-475, Revision 1 is not applicable and therefore, not adopted.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration by a reference to a generic analysis published in the Federal Register on November 13, 2007 (72 FR 63935), which is presented below:

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

The proposed change generically implements TSTF-475, Revision 1, “Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.” TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, “Control Rod OPERABILITY”, [ ], and (3) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The consequences of an accident after adopting TSTF-475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously analyzed. Thus, 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.

TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 frequency in TS 3.1.3, “Control Rod OPERABILITY”, [ ], and (3) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The GE Nuclear Energy Report, “CRD Notching Surveillance Testing for Limerick Generating Station,” dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF-475, Revision 1 [ ] do not involve a significant reduction in a margin of safety.

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

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

NRC Acting Branch Chief: Patrick Milano.

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

Date of amendment request: July 30, 2007.

Description of amendment request: The amendments would revise the Technical Specifications to allow single header operation of the nuclear service water system (NSWS) for a time period of 35 days. The change will facilitate future maintenance of the NSWS headers.

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:

[First Standard]

Does operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed single supply header operation configuration for NSWS operation and the associated proposed TS and Bases changes have been evaluated to assess their impact on plant operation and to ensure that the design basis safety functions of safety related systems are not adversely impacted. During single supply header operation, the operating NSWS header will be able to supply all required NSWS flow to safety related components. It was demonstrated that proposed single failures would not cause the NSWS to be rendered incapable of performing its required safety related function under accident conditions.

The purpose of this amendment request is to ultimately facilitate inspection and maintenance of the NSWS supply headers. Therefore, NRC approval of this request will ultimately help to enhance the long-term structural integrity of the NSWS and will help to ensure the system's reliability for many years.

In general, the NSWS serves as an accident mitigation system and cannot by itself initiate an accident or transient situation. The only exception is that the NSWS piping can serve as a source of floodwater to safety related equipment in the auxiliary building or in the diesel generator buildings in the event of a leak or a break in the system piping. The probability of such an event is not significantly increased as a result of this proposed request. NSWS piping added in support of the proposed request will be tested and maintained in a manner consistent with that for comparable safety related piping in the NSWS.

The proposed 35 day TS Required Action Completion Time has been evaluated for risk significance and the results of this evaluation have been found acceptable. The probabilities of occurrence of accidents presented in the UFSAR will not increase as a result of implementation of this change. Because the PRA analysis supporting the proposed change yielded acceptable results, the NSWS will maintain its required availability in response to accident situations. Since NSWS availability is maintained, the response of the plant to accident situations will remain acceptable and the consequences of accidents presented in the UFSAR will not increase.

[Second Standard]

Does operation of the facility in accordance with the proposed amendment create the Start Printed Page 10297possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Implementation of this amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed request does not affect the basic operation of the NSWS or any of the systems that it supports. These include the Emergency Core Cooling System, the Containment Spray System, the Containment Valve Injection Water System, the Auxiliary Feedwater System, the Component Cooling Water System, the Control Room Area Ventilation System, the Control Room Area Chilled Water System, the Auxiliary Building Filtered Ventilation Exhaust System, or the Diesel Generators. During proposed single supply header operation, the NSWS will remain capable of fulfilling all of its design basis requirements, even when assuming the required single failure.

No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant which will introduce any new type of accident outside those assumed in the UFSAR.

[Third Standard]

Does operation of the facility in accordance with the proposed amendment involve a significant reduction in the margin of safety?

Response: No.

Implementation of this amendment will not involve a significant reduction in any margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed TS amendment. During single supply header operation, the NSWS and its supported systems will remain capable of performing their required functions even assuming the postulated single failure. No safety margins will be impacted.

The PRA conducted for this proposed amendment demonstrated that the impact on overall plant risk remains acceptable during single supply header operation. Therefore, there is not 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 amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Melanie C. Wong, Acting.

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

Date of amendment request: September 27, 2007.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.7.2 (Main Steam Isolation Valves) and TS 3.7.3 (Main Feedwater Isolation Valves, Main Feedwater Control Valves, Associated Bypass Valves and Tempering Valves) by removing the specific isolation time for the isolation valves from the associated Surveillance Requirements.

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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below.

Criterion 1: The Proposed Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed changes allow relocating main steam and main feedwater valve isolation times to the licensee-controlled document that is referenced in the Bases. The proposed changes are described in Technical Specification Task Force (TSTF) Standard TS Change Traveler TSTF-491 related to relocating the main steam and main feedwater valves isolation times to the licensee-controlled document that is referenced in the Bases and replacing the isolation time with the phrase, “within limits.” The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed changes relocate the main steam and main feedwater isolation valve times to the licensee-controlled document that is referenced in the Bases. The requirements to perform the testing of these isolation valves are retained in the TSs. Future changes to the Bases or licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, test and experiments,” to ensure that such changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupational/public radiation exposures. Therefore, the changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2: The Proposed Changes Do Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed changes relocate the main steam and main feedwater valve isolation times to the licensee-controlled document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TS with the phrase “within limits”. The changes do not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal plant operation. The requirements in the TSs continue to require testing of the main steam and main feedwater isolation valves to ensure the proper functioning of these isolation valves. Therefore, the changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3: The Proposed Changes Do Not Involve a Significant Reduction in the Margin of Safety.

The proposed changes relocate the main steam and main feedwater valve isolation times to the licensee-controlled document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TSs with the phrase “within limits.” Instituting the proposed changes will continue to ensure the testing of main steam and main feedwater isolation valves. Changes to the Bases or license-controlled document are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that main steam and feedwater isolation valve testing is conducted such that there is no significant reduction in the margin of safety. The margin of safety provided by the isolation valves is unaffected by the proposed changes since there continue to be TS requirements to ensure the testing of main steam and main Start Printed Page 10298feedwater isolation valves. The proposed changes maintain sufficient controls to preserve the current margins of safety.

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Melanie C. Wong, Acting.

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: December 21, 2007.

Description of amendment request: The proposed amendment revises Technical Specification (TS) Surveillance Requirements (SR) 3.8.4.2 and 3.8.4.5 to add an additional acceptance criterion to verify that total battery connector resistance is within pre-established limits that ensure the batteries can perform their design functions. The proposed amendment is in response to a non-cited violation that was documented in NRC Component Design Bases Inspection Report 05000254/2006003(DRS), 05000265/2006003(DRS).

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 revisions of SR 3.8.4.2 and SR 3.8.4.5 to add a battery connector resistance acceptance criterion will not challenge the ability of the safety-related batteries to perform their safety function. Appropriate monitoring and maintenance will continue to be performed on the safety-related batteries. In addition, the safety-related batteries are within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with this equipment.

Current TS requirements will not be altered and will continue to require that the equipment be regularly monitored and tested. Since the proposed change does not alter the manner in which the batteries are operated, there is no significant impact on reactor operation.

The proposed change does not involve a physical change to the batteries, nor does it change the safety function of the batteries. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components.

Therefore, these changes will not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes revising SR 3.8.4.2 and SR 3.8.4.5 to add an additional acceptance criterion for battery connector resistance is an increase in conservatism, without a change in system testing methods, operation, or control. Safety-related batteries installed in the plant will be required to meet criteria more restrictive and conservative than current acceptance criteria and standards. The proposed change does not affect the manner in which the batteries are tested and maintained; therefore, there are no new failure mechanisms for the system.

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

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

Response: No.

The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated. The change is conservative and further ensures safety-related battery operability and availability.

As such, sufficient DC capacity to support operation of mitigation equipment is enhanced, which results in an increase in the margin of safety.

Therefore, the proposed 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 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.

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

Date of amendment request: December 20, 2007.

Description of amendment request: Duane Arnold Energy Center (DAEC) requests a change, consistent with the adoption of TSTF-475, Revision 1, an approved change to the Standard Technical Specifications (STS) for General Electric (GE) Plants (NUREG-1433, BWR/4) and plant specific technical specifications (TS), that allows: (1) Revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch testing of fully withdrawn control rod, from “7 days after the control rod is withdrawn and THERMAL POWER is greater than 20% [Rated Thermal Power] RTP” to “31 days after the control rod is withdrawn and THERMAL POWER is greater than 20% RTP” and (2) revising Example 1.4-3 in Section 1.4 “Frequency” to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the “SURVEILLANCE” column in addition to the time periods in the “FREQUENCY” column.

The NRC staff acknowledges that, in item (1) above, the wording that is to be adopted by the Duane Arnold TS in SR 3.1.3.2 (“31 days after the control rod is withdrawn and THERMAL POWER is greater than 20% RTP”) is a deviation from the language in the Improved STS (“31 days after the control rod is withdrawn and THERMAL POWER is greater than the [Low Power Setpoint] LPSP of the [Rod Worth Minimizer] RWM.”) This deviation from NUREG-1433 was incorporated into the DAEC TS by Amendment 223 dated May 22, 1998, in the conversion of the DAEC TS to the Improved STS.

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 (NSHC) through incorporation by reference of the NSHC determination (NSHCD) published in the Federal Register Notice dated November 13, 2007, that announced the availability of TS improvement through the consolidated line item improvement process (CLIIP). The NSHCD, with references to BWR/6 information deleted and with clarifying comments inserted within brackets [ ], 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 generically implements TSTF-475, Revision 1, “Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.” TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) Start Printed Page 10299STS. The changes: (1) Revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, “Control Rod OPERABILITY” and (2) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension.

The consequences of an accident after adopting TSTF-475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously analyzed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

TSTF-475, Revision 1 [, as adopted by DAEC TS,] will: (1) Revise the TS SR 3.1.3.2 frequency in TS 3.1.3, “Control Rod OPERABILITY” and (2) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension.

The GE Nuclear Energy Report, “CRD Notching Surveillance Testing for Limerick Generating Station,” dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF-475, Revision 1 [, as adopted by DAEC TS,] do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Marjan Mashhadi, Florida Power & Light Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.

NRC Acting Branch Chief: Patrick Milano.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

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

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

Date of application for amendment: November 15, 2007, as supplemented by letter dated December 21, 2007.

Brief description of amendment: The amendment is a one-time change that revised Technical Specification (TS) Section 3.1.7, “Rod Position Indication.” The requirements related to one inoperable bank demand position indicator (DPI) are modified by a footnote to allow two DPIs to be inoperable per bank for one or more banks on a temporary basis during the current operating cycle (Cycle 25). This provision allows for corrective maintenance on three inoperable DPIs in the rod position indication system that necessitates removing both DPIs for the affected rod banks from service during the repair. This amendment expires at the end of operating Cycle 25.

Date of issuance: January 29, 2008.

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

Amendment No. 217.

Renewed Facility Operating License No. DPR-23: The amendment revises the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register: November 28, 2007 (72 FR 67321).

Public comments requested as to proposed no significant hazards consideration (NSHC): No.

The Commission's related evaluation of the amendment and final NSHC determination are contained in a safety evaluation dated January 29, 2008.

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602-1551.

NRC Branch Chief: Thomas H. Boyce.

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

Date of application for amendment: October 2, 2007.

Brief description of amendment: The amendment revises Technical Specification Sections 3.7, “Auxiliary Electrical Systems,” and 4.6, “Periodic Testing of Emergency Power System,” to change the testing requirements for ensuring operability of the remaining operable emergency diesel generator (EDG) when the other EDG is inoperable. In addition, the amendment adds a new specification when two EDGs are inoperable and revises the surveillance requirements for the EDGs.

Date of issuance: February 7, 2008.

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

Amendment No.: 194.

Facility Operating License No. DPR-43: Amendment revised the License and Technical Specifications. Start Printed Page 10300

Date of initial notice in Federal Register: November 20, 2007 (72 FR 65363)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 7, 2008.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: February 16, 2007.

Brief description of amendment: The proposed amendment would revise Technical Specification 3/4.4.3, “Reactor Coolant System, Relief Valves” to modify the method of testing the pressurizer Power Operated Relief Valves (PORVs). Specifically, the requirement for bench testing the valves is changed to accommodate testing of the PORVs while installed in the plant. The change is requested due to the installation of new PORVs that are welded to the piping rather than bolted into the system.

Date of issuance: February 12, 2008.

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

Amendment No.: 302.

Facility Operating License No. DPR-65: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: November 19, 2007 (72 FR 65084).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 12, 2008.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas Nuclear One, Units 1 and 2, Pope County, Arkansas

Date of amendment request: April 24, 2007, as supplemented by letter dated August 2, 2007, and electronic mail dated January 8, 2008.

Brief description of amendments: The amendments relocate the Fuel Handling Area Ventilation System and associated Ventilation Filter Testing Program requirements that are included in the Unit 1 Technical Specifications (TS) 3.7.12 and 5.5.11 and the Unit 2 TS 3.9.11 and 6.5.11 to the unit-specific Technical Requirements Manuals (TRMs). The TRMs are licensee-controlled documents which are controlled under 10 CFR 50.59, “Changes, tests, and experiments.”

Date of issuance: February 4, 2008.

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

Amendment Nos.: Unit 1-231; Unit 2-274.

Renewed Facility Operating License Nos. DPR-51 and NPF-6: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 5, 2007 (72 FR 31098). The supplemental letter dated August 2, 2007, and electronic mail dated January 8, 2008, 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 February 4, 2008.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: October 24, 2007.

Brief description of amendment: The amendment revises the containment buffering agent used for pH control under post loss-of-coolant accident (LOCA) conditions, from trisodium phosphate to sodium tetraborate.

Date of issuance: February 7, 2008.

Effective date: As of the date of issuance, and shall be implemented prior to entry into Mode 4 following completion of the spring 2008 refueling outage.

Amendment No.: 253.

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

Date of initial notice in Federal Register: December 4, 2007 (72 FR 68211).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 7, 2008.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: July 25, 2007, as supplemented November 1, 2007.

Brief description of amendment: The proposed amendment would modify the Technical Specifications by adding an Action Statement to the Limiting Conditions for Operation (LCOs) for TS 3.7.4, “Control Room Air Conditioning (AC) System.” Specifically, the new Action statement allows 72 hours to restore one control room air conditioning subsystem to operable status and requires verification that the control room temperature remains below 90 °F every 4 hours during the period of inoperability. The change is consistent with NRC-approved Revision 3 to Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-477, “Add Action Statement for Two Inoperable Control Room Air Conditioning Subsystems.”

Date of issuance: January 23, 2008.

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

Amendment No.: 290.

Facility Operating License No. DPR-59: The amendment revises the License and the Technical Specifications.

Date of initial notice in Federal Register: September 11, 2007 (72 FR 51855).

The November 1, 2007, supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 23, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 6, 2006, supplemented by letters dated August 10, 2007, and December 20, 2007.

Brief description of amendment: The amendment would revise Appendix A, technical specification (TS), Core Operating Limits Report analytical methods referenced in TS 5.6.5.b to add EMF-2103 (P)(A), “Realistic Large Break LOCA Methodology for Pressurized Water Reactors.”

Date of issuance: January 31, 2008.

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

Amendment No.: 229.

Facility Operating License No. DPR-20: Amendment revised the technical specifications. Start Printed Page 10301

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75995)

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 31, 2008.

No significant hazards consideration comments received: No.

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

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

Date of application for amendment: January 8, 2007 as supplemented by letter dated October 12, 2007.

Brief description of amendment: The amendments extended the reactor trip system and engineered safety features actuation system completion times, bypass test times, and surveillance test intervals for technical specifications (TS) 3.3.1, “RTS Instrumentation,” TS 3.3.2, “ESFAS Instrumentation,” and TS 3.3.6, “Containment Ventilation Isolation Instrumentation.”

Date of issuance: January 29, 2008.

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

Amendment Nos.: 153, 153, 148, and 148.

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

Date of initial notice in Federal Register: March 27, 2007 (72 FR 14305).

The October 12, 2007, supplement, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 29, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 12, 2007.

Brief description of amendments: The amendments revise TS 3.1.3.2, “Position Indication Systems—Operating,” to allow for the use of an alternate method, other than the movable incore detectors, to monitor the position of a control rod or shutdown rod in the event of a problem with the analog rod position indication system. The use of this alternate method will reduce the required frequency of flux mapping using the movable incore detectors to determine the position of the non-indicating rod, thus reducing the wear on the movable incore detector system that is also used to complete other required TS surveillances.

Date of issuance: January 28, 2008.

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

Amendment Nos: 237 and 232.

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

Date of initial notice in Federal Register: November 28, 2007 (72 FR 67323).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 28, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 19, 2007.

Brief description of amendment: The amendment revises Limiting Condition for Operation 3.10.1 to expand its scope to include provisions for temperature excursions greater than 200 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4, using the Consolidated Line Item Improvement Process.

Date of issuance: February 7, 2008.

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

Amendment No.: 121.

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

Date of initial notice in Federal Register: November 20, 2007 (72 FR 65368).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 7, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: February 15, 2007, as supplemented on November 30, 2007.

Brief description of amendment: The amendment revised the Technical Specifications Surveillance Requirement (SR) 3.8.4.2, “DC [Direct Current] Sources—Operating,” to specify that the Division 1 battery chargers are verified to supply ≥150 amps and the Division 2 battery chargers are verified to supply ≥110 amps.

Date of issuance: January 30, 2008.

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

Amendment No.: 153.

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

Date of initial notice in Federal Register: April 24, 2007 (72 FR 20384).

The supplemental letter contained clarifying information, did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 30, 2008.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 21, 2007.

Brief description of amendment: The amendment revises Technical Specifications (TS) safety limit (SL) requirements related to the use of a non-cycle specific peak linear heat rate (PLHR) SL of 22 kW/ft to fuel centerline melt (FCM). The TS change is consistent with the Technical Specification Task Force (TSTF) 445-A, Revision 1. Because these Limiting Safety Systems Setting (LSSS) values appear in the FCS TS Bases Sections of TS 1.3, TS 1.0, Safety Limits and Limiting Safety System Settings, was also revised to more clearly align with the Combustion Engineering (CE) Standard Technical Specifications (STS) 2.0 in content. Therefore, TS Section 1.1, Safety Limits—Reactor Core, is revised to incorporate the TSTF-445-A, Revision 1, peak fuel centerline temperature criteria and TS 1.2, Safety Limits—Reactor Coolant System Pressure, is revised to incorporate the SL violation Start Printed Page 10302action which is currently delineated in administrative control TS 5.7.1. TS Section 1.3, Limiting Safety System Settings, was relocated to the currently unused TS Section 2.13 to be more consistent with the content of the CE STS (i.e., the LSSS will be located in the Limiting Conditions for Operation (LCO) section of the FCS TS which is similar to the LCO/Surveillance Requirements Section 3.0 of the STS). As noted above, the administrative control in TS 5.7.1, Safety Limit Violation, is relocated. Also, administrative control TS 5.9.5, Core Operating Limits Report (COLR), item a., is revised to add TS 2.13, RPS Limiting Safety System Settings, Table 2-11, Items 6, 8, and 9, to the list of items that shall be documented in the COLR. The TS Table of Contents (TOC) is also updated to reflect the deletion and subsequent renumbering of Section 1.3 and Table 1-1 to TS 2.13 and Table 2-11, respectively. The TOC is also updated to delineate the new TS subsections 1.1.1 and 1.1.2, provide the revised titles for TS 1.0, 1.1, 1.2, and 2.13, and to reflect TS 5.7.1 as “Not used.”

Date of issuance: February 4, 2008.

Effective date: As of its date of issuance and prior to startup from the 2008 refueling outage.

Amendment No.: 252.

Renewed Facility Operating License No. DPR-40: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 6, 2007 (72 FR 62690). The Commission's related evaluation of the amendment is contained in a safety evaluation dated February 4, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 11, 2007, as supplemented on October 25, December 4 and 26, 2006, February 13, March 14 and 22, April 13, 17, 23, 26, and 27, May 3, 9, 14, and 21, June 1, 4, 8, 14, 20, and 27, July 6, 12, 13, 30, and 31, August 3, 13, 15, and 28, September 19, October 5, November 30, December 10, 2007, and January 9, 24, and 29, 2008.

Brief description of amendments: The amendments increase the SSES 1 and 2 licensed thermal power to 3952 Mega-watts thermal (MWt), which is 20% above the original rated thermal power (RTP) of 3293 MWt, and approximately 13% above the current RTP of 3489 MWt. The amendments revise the SSES 1 and 2 Operating License and Technical Specifications necessary to implement the increased power level.

Date of issuance: January 30, 2008.

Effective date: As of the date of issuance and to be implemented in accordance with the issued License Conditions.

Amendment Nos.: 246 and 224.

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

Date of initial notice in Federal Register: March 13, 2007 (72 FR 11392). The supplements dated October 25, December 4 and 26, 2006, February 13, March 14 and 22, April 13, 17, 23, 26, and 27, May 3, 9, 14, and 21, June 1, 4, 8, 14, 20, and 27, July 6, 12, 13, 30, and 31, August 3, 13, 15, and 28, September 19, October 5, November 30, December 10, 2007, and January 9, 24, and 29, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 30, 2008.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 15th day of February 2008.

For The Nuclear Regulatory Commission.

Catherine Haney,

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

End Signature End Preamble

[FR Doc. E8-3481 Filed 2-25-08; 8:45 am]

BILLING CODE 7590-01-P