Applications and Amendments to Facility Operating Licenses: Biweekly Notice

As Enacted ( Federal Register)

Cited authorities 8

Cited in

Federal Register: January 27, 2009 (Volume 74, Number 16)

Notices

Page 4764-4781

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

DOCID:fr27ja09-59

NUCLEAR REGULATORY COMMISSION

NRC-2009-0016

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the

Commission or NRC staff) is publishing this regular biweekly notice.

The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from December 31, 2008 to January 13, 2009. The last biweekly notice was published on January 13, 2009 (74 FR 1712).

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- 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.

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Written comments may be submitted by mail to the Chief, Rulemaking,

Directives and Editing Branch, TWB-05-B01M, 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. 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.

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license.

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.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), 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 all adjudicatory 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 Viewer \TM\ to access the Electronic Information Exchange (EIE), a component of the E-

Filing system. The Workplace Forms Viewer\TM\ 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

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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 electronic filing

Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,

Monday through Friday. The help electronic filing Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at

MSHD.Resource@nrc.gov.

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).

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.resource@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50- 529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of amendment request: July 2, 2008.

Description of amendment request: The amendments would revise

Technical Specification (TS) 4.2.2, ``Control Element Assemblies,'' to support replacement of the full strength control element assemblies

(CEAs) with a new design beginning with the 14th refueling outage

(U3R14) for Palo Verde Nuclear Generating Station (PVNGS), Unit 3 in the spring of 2009. Additionally, Arizona Public Service Company (APS) will be updating the TS by removing the registered trademark

``Inconel'' while retaining the generic terminology ``Alloy 625'' and deleting the references to part-length CEAs in TS 4.2.2.

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

Replacement of full-strength compression sleeve control element assemblies with full-strength silver (Ag)-indium (In)-Cadmium (Cd) control element assemblies.

Response: No.

The proposed change involves a new design for the full-strength

Control Element Assemblies (CEA) that replaces a portion of B4C pellets (including the compression sleeve) in the tips of the CEA fingers with hollow silver-indium-cadmium slugs.

The following events are related to inadvertent movement of the

CEAs; however, they are not initiated by the CEAs.

Uncontrolled Control Element Assembly Withdrawal from a

Subcritical or Low (Hot Zero) Power Condition.

Uncontrolled Control Element Assembly Withdrawal at

Power.

Single Full-Strength Control Element Assembly Drop.

Control Element Assembly Ejection.

These previously analyzed accidents are initiated by the failure of plant structures, systems, or components (SSC) other than the CEA itself. The proposed change to the CEA design does not have a detrimental impact on the integrity of any plant SSC that initiates an analyzed event. Additionally, the CEAs mitigate other events. In these events, the chrome plating on the portion of the clad exterior and the added weight has been conservatively accounted for in the

SCRAM [safety control rod axe man] calculation. The change does not adversely affect the protective and mitigative capabilities of the plant, nor does the change affect the initiation or probability of occurrence of any accident. The SSCs will continue to perform their intended safety functions.

The proposed change in CEA design has resulted in a slight (less than 1%) reduction of total reactivity.

Computer modeling events which exhibit sensitivity to time dependent rod worth (sheared shaft/seized rotor, loss of flow from

SAFDL [specified acceptable fuel design limits] and total loss of reactor coolant flow) demonstrate that all acceptance criteria continued to be met.

Therefore this change will not significantly increase the probability or consequences of any accident previously evaluated.

The removal of the registered trademark name ``Inconel''.

Response: No.

This change is considered editorial. Inconel is a registered trademark of Special Metals Corporation, while Alloy 625 is a generic alloy designation from the Unified Numbering System.

Retaining the already referenced term ``Alloy 625'' does not involve a significant increase in the probability or consequences of an accident previously evaluated, as the material properties and application of Alloy 625 have not changed.

Deletion of the references to part-length control element assemblies.

Response: No.

This change is considered editorial. The removal of this information does not involve a significant increase in the probability or consequences of an accident previously evaluated as the part-length CEAs were

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replaced in accordance with License Amendment 152, dated March 23, 2004 (Agency Document Access and Management System (ADAMS) Accession

No. ML040860573) and the information is no longer applicable. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Replacement of full-strength compression sleeve control element assemblies with full-strength silver(Ag)-indium(In)-Cadmium(Cd) control element assemblies.

Response: No.

There are three differences in the replacement CEAs as compared to the current CEAs.

First, there is a very slight change in the outside diameter of a portion of the cladding on the replacement CEAs due to chrome plating on the lower portion of cladding. Analysis demonstrates that this change will not cause interference between the CEA cladding and the guide tube inside diameter in the buffer region. Secondly, there is a slight increase in weight with the Ag-In-Cd CEAs. However, this difference has been analyzed with respect to the performance capability of the CEDMs [Control Element Drive Mechanisms] and found to be within design capabilities and design analyses. Finally, the upper edges of the spider bosses have been chamfered to prevent damage to the self-latching mechanisms that can occur if the CEA hangs up when lifting through the upper guide structure cut outs.

This change is for ease of maintenance and has no impact on operation of the CEAs.

Therefore, the Ag-In-Cd CEAs are identical to the compression sleeve CEAs in terms of form, fit and function and the proposed change will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. The possibility of a new or different malfunction of safety-related equipment is not created. No new accident scenarios, transient precursors, or limiting single failures are introduced as a result of these changes. There will be no adverse effects or challenges imposed on any safety-related system as a result of these changes. Therefore, the possibility of a new or different accident from any accident previously evaluated is not created as a result of any dimensional change.

The removal of the registered trademark name ``Inconel''.

Response: No.

This change is considered editorial. Inconel is a registered trademark of Special Metals Corporation, while Alloy 625 is a generic alloy designation from the Unified Numbering System.

Retaining the already referenced term ``Alloy 625'' does not create the possibility of a new or different kind of accident from any accident previously evaluated, as the material properties and application of Alloy 625 have not changed.

Deletion of the references to part-length control element assemblies.

Response: No.

This change is considered editorial. The removal of this information does not create the possibility of a new or different kind of accident from any accident previously evaluated as the part- length CEAs were replaced in accordance with License Amendment 152, dated March 23, 2004 (Agency Document Access and Management System

(ADAMS) Accession No. ML040860573) and the information is no longer applicable. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Replacement of full-strength compression sleeve control element assemblies with full-strength silver(Ac)-indium(In)-Cadmium(Cd) control element assemblies.

Response: No.

Reactor core safety limits are established in the PVNGS

Technical Specifications to prevent overheating of the fuel and cladding that would result in the release of fission products to the reactor coolant during steady state operation, normal operational transients, and anticipated operational occurrences. The margin to these safety limits is not affected by the CEA design changes under consideration.

Overheating of the fuel is prevented by maintaining steady state, peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. If the local LHR is high enough to cause the fuel centerline temperature to reach the melting point of the fuel, expansion of the pellet caused by centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Compliance with the DNBR [departure from nucleate boiling ratio] and fuel centerline melt specified acceptable fuel design limits

(SAFDLs) is assured through the CEA insertion limits and alignment technical specifications, and through the power distribution limit technical specifications.

There is no change to the operation of the full-strength CEAs due to the change from compression sleeve CEAs to Ag-In-Cd CEAs.

Since the Ag-In-Cd CEAs may be used to control power distribution similar to the compression sleeve CEAs, power distributions will still be controlled and maintained within the limits necessary to assure SAFDLs are met.

The proposed change in CEA design has resulted in a slight (less than 1%) reduction in total reactivity.

Computer modeling results of events which exhibit sensitivity to time dependent rod worth (sheared shaft/seized rotor, loss of flow from SAFDL and total loss of reactor coolant flow) demonstrate that all acceptance criteria continued to be met.

Therefore, since SAFDLs continue to be met, the change from compression sleeve CEAs to Ag-In-Cd CEAs does not involve a significant reduction in a margin of safety.

The removal of the registered trademark name ``Inconel''.

Response: No.

The removal of the registered trademark name ``Inconel'' [ ] is considered editorial. Inconel is a registered trademark of Special

Metals Corporation, while Alloy 625 is a generic alloy designation from the Unified Numbering System. Retaining the already referenced term ``Alloy 625'' does not involve a significant reduction in the margin of safety as the material properties and application of Alloy 625 have not changed.

Deletion of the references to part-length control element assemblies.

Response: No.

This change is considered editorial. The removal of this information does not involve a significant reduction in the margin of safety as the part-length CEAs were replaced in accordance with

Amendment 152, dated March 23, 2004 (Agency Document Access and

Management System (ADAMS) Accession No. ML040860573) and the information is no longer applicable.

The NRC staff has reviewed the licensee's analysis and, based on that review, it appears that the three 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: Michael G. Green, Senior Regulatory Counsel,

Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,

Phoenix, Arizona 85072-2034.

NRC Branch Chief: Michael T. Markley.

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: October 6, 2008.

Description of amendments request: The proposed change would remove work hour controls and/or references to the NRC Generic Letter 82-12 from the administrative control sections of the technical specifications. On April 17, 2007, the NRC approved a final rule that amended 10 CFR Part 26 and, among other changes, established requirements for managing worker fatigue at operating nuclear power plants. Subpart I, ``Managing Fatigue,'' specifically addresses managing worker fatigue by designating individual break requirements, work hour limits, and annual reporting requirements. Subpart I was published in the Federal Register on March 31, 2008 (73 FR 16966), with a required implementation period of 18 months. Compliance is, therefore, required by October 1, 2009. In order to support compliance with 10 CFR Part 26, Subpart I, the licensee is proposing to remove these work hour controls from Technical Specification 5.2.2.e at the

Brunswick Steam Electric Plant, Units 1 and 2.

Basis for proposed no significant hazards consideration determination Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the

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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 remove TS [technical specification] controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26. Removal of the TS requirements will be performed concurrently with the implementation of the 10 CFR Part 26, Subpart I requirements. The proposed changes do not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not impact the initiators or assumptions of analyzed events, nor do they impact the mitigation of accidents or transient events.

Therefore, it is concluded that these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26.

Work hours will continue to be controlled in accordance with NRC requirements. The new rule allows for deviations from controls to mitigate or prevent a condition adverse to safety or as necessary to maintain the security of the facility. This ensures that the new rule will not restrict work hours and thereby create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, add any initiators, or effect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected.

Therefore, it is concluded that this change does 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 remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26. The proposed changes do not involve any physical changes to plant or the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes will not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

Removal of plant-specific TS administrative requirements will not reduce a margin of safety because the requirements in 10 CFR

Part 26 are adequate to ensure that worker fatigue is managed.

Therefore, it is concluded that these changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: David T. Conley, Associate General Counsel

II--Legal Department, Progress Energy Service Company, LLC, Post Office

Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: October 6, 2008.

Robinson Steam Electric Plant, Unit 2.

Basis for proposed no significant hazards consideration determination 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 remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26.

The proposed changes do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected.

Response: No.

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maintained, modified, tested, or inspected. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change.

The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes will not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

Removal of plant-specific TS administrative requirements will not reduce a margin of safety because the requirements in 10 CFR

Part 26 are adequate to ensure that worker fatigue is managed.

Therefore, it is concluded that these changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: David T. Conley, Associate General Counsel

II--Legal Department, Progress Energy Service Company, LLC, Post Office

Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of amendment request: October 6, 2008.

Description of amendment request: The proposed change would remove work hour controls and/or references to the NRC Generic Letter 82-12 from the administrative control sections of the technical specifications. On April 17, 2007, the NRC approved a final rule that amended 10 CFR Part 26 and, among other changes, established requirements for managing worker fatigue at operating nuclear power plants. Subpart I, ``Managing Fatigue,'' specifically addresses managing worker fatigue by designating individual break requirements, work hour limits, and annual reporting requirements. Subpart I was published in the Federal Register on March 31, 2008 (73 FR 16966), with a required implementation period of 18 months. Compliance is, therefore, required by October 1, 2009. In order to support compliance with 10 CFR Part 26, Subpart I, the licensee is proposing to remove these work hour controls from Technical Specification 6.2.2.f at the

Shearon Harris Nuclear Power Plant, Unit 1.

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 remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26.

The proposed changes do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected.

Response: No.

The proposed changes remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26. The proposed changes do not involve any physical changes to the plant or the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes will not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

Removal of plant-specific TS administrative requirements will not reduce a margin of safety because the requirements in 10 CFR

Part 26 are adequate to ensure that worker fatigue is managed.

Therefore, it is concluded that these changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: David T. Conley, Associate General Counsel

II--Legal Department, Progress Energy Service Company, LLC, Post Office

Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

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

No. 2, Pope County, Arkansas

Date of amendment request: November 13, 2008.

Description of amendment request: The proposed change will modify

Technical Specification (TS) 3.3.1.1, ``Reactor Protective

Instrumentation.'' Specifically, Table 4.3-1 and the associated Notes 7 and 8 will be revised to clarify and streamline the reactor coolant system (RCS) flow verification requirements associated with the departure from nucleate boiling ratio (DNBR) reactor trip signal.

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 CPC [Core Protection Calculator] reactor protective function is not considered an accident initiator. The primary function is to initiate an automatic reactor trip signal when specific plant conditions are reached, thereby limiting the consequences of an accident. The proposed change acts to eliminate unnecessary conservatisms and accordingly increase operational margin by eliminating the requirement to use

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calorimetric flow measurement in the CPC flow verification. This method of verification will normally only be used in the future during periods when the COLSS [Core Operating Limits Supervisory

System] RCP [Reactor Coolant Pump] [Delta] p flow measurement is unavailable. Regardless of the method of verification used, the CPC will continue to be verified to have an indicated RCS flow equal to or conservative relative to the measured RCS flow on a once per 12- hour basis. In so doing, the CPC will continue to act to generate a reactor trip on low DNBR as originally designed in order to ensure the DNBR reactor core Safety Limit is not exceeded.

The relocation of measurement uncertainty references to the TS

Bases does not reduce the requirements to account for uncertainties in any Limiting Safety System Setting (LSSS) designed to protect reactor core Safety Limits. The necessary uncertainties will continue to be applied as required and will be controlled in accordance with TS 6.5.14, Technical Specification Bases Control

Program, and station procedures.

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 result in any physical plant modifications or changes in the way the plant is operated. In addition, the CPCs are unrelated to any type of accident initiator previously evaluated.

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 proposed change increases operating margin when the COLSS

RCP [Delta]p flow measurement is available for use while unaffecting the CPC ability to initiate an automatic reactor trip on low DNBR prior to the DNBR reactor core safety limit being exceeded.

Relocating the references to measurement uncertainties to the TS

Bases likewise has no impact on the CPC design function and the uncertainties will continue to be applied as required and controlled in accordance with TS 6.5.14, Technical Specification Bases Control

Program, and station procedures.

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

Attorney for licensee: Terence A. Burke, Associate General

Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,

Mississippi 39213.

NRC Branch Chief: Michael T. Markley.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,

Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,

Louisiana

Date of amendment request: December 8, 2008

Description of amendment request: The proposed amendment adds a license condition to allow a one-time extension of surveillance requirements involving the 18-month channel calibration and logic system functional tests for one channel of the reactor water level instrumentation system. The extension is to account for the effects of rescheduling the next refueling outage from early to late 2009.

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 requested action is a one-time extension to the performance interval of certain TS [Technical Specification] surveillance requirements. The performance of the surveillances, or the failure to perform the surveillances, is not a precursor to an accident.

Performing the surveillances or failing to perform the surveillances does not affect the probability of an accident. Therefore, the proposed delay in performance of the surveillance requirements in this amendment request does not increase the probability of an accident previously evaluated.

A delay in performing the surveillances does not result in a system being unable to perform its required function. Additionally, the defense in depth of the system design provides additional confidence that the safety function is maintained. In the case of this one-time extension request, the relatively short period of additional time that the systems and components will be in service before the next performance of the surveillance will not affect the ability of those systems to operate as designed. Therefore, the systems required to mitigate accidents will remain capable of performing their required function. No new failure modes have been introduced because of this action and the consequences remain consistent with previously evaluated accidents. Therefore, the proposed delay in performance of the surveillance requirement in this amendment request does not involve a significant increase in the consequences of an accident.

Response: No.

The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC), or a change in the way any SSC is operated. The surveillance intervals of the level instrumentation are currently evaluated for 30 months, which bounds the requested interval extension. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the one-time surveillance extension being requested.

Response: No.

The proposed amendment is a one-time extension of the performance-interval of certain TS surveillance requirements.

Extending the surveillance requirements does not involve a modification of any TS Limiting Conditions for Operation. Extending the surveillance frequency does not involve a change to any limit on accident consequences specified in the license or regulations.

Extending the surveillance frequency does not involve a change to how accidents are mitigated or a significant increase in the consequences of an accident. Extending the surveillance frequency does not involve a change in a methodology used to evaluate consequences of an accident. Extending the surveillance frequency does not involve a change in any operating procedure or process. The surveillance intervals of the level instrumentation are currently evaluated for 30 months which bounds the requested interval extension. The components involved in this request have exhibited reliable operation based on the results of the most recent performances of their 18-month surveillance requirements and the associated functional surveillances.

Based on the limited additional period of time that the systems and components will be in service before the surveillance is next performed, as well as the operating experience that these surveillances are typically successful when performed, it is reasonable to conclude that the margin of safety associated with the surveillance requirement will not be affected by the requested extension.

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

Page 4771

Attorney for licensee: Terence A. Burke, Associate General

Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,

Mississippi 39213.

NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Units 1 and 2, Will County, Illinois

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

Date of amendment request: December 4, 2008.

Description of amendment request: The proposed amendments would revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16,

``RCS Specific Activity,'' and Surveillance Requirements 3.4.16.1 and 3.4.16.3. The proposed changes would replace the current TS 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-133 definition that would replace the current E Bar average disintegration energy definition. In addition, the current dose equivalent I-131 definition would be reformatted. The availability of this TS revision was announced in the Federal Register on March 15, 2007 (72 FR 12217) as part of the consolidated line item improvement process. The licensee affirmed the applicability of the model no significant hazards consideration determination in its application.

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

Reactor coolant specific activity is not an initiator for any accident previously evaluated. The Completion Time when primary coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit primary coolant noble gases to concentrations consistent with the accident analyses. The proposed change to the Completion Time has no impact on the consequences of any design basis accident since the consequences of an accident during the extended Completion Time are the same as the consequences of an accident during the Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased.

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 in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter. The change does not create the potential for a new or different kind of accident from any previously calculated.

Criterion 3--The Proposed Change Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change revises the limits on noble gas radioactivity in the primary coolant. The proposed change is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses.

The Nuclear Regulatory Commission (NRC) staff has reviewed the analysis adopted by the licensee 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 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.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power

Station, Unit No. 1, DeWitt County, Illinois

Date of amendment request: September 2, 2008.

Description of amendment request: The proposed amendments would relocate Surveillance Requirements (SR) 3.8.3.6 from the technical specifications (TSs) to a licensee-controlled document. SR 3.8.3.6 requires Emergency Diesel Generator fuel oil storage tanks to be drained, sediment removed, and cleaned on a 10-year interval. The change is consistent with the current revision (i.e., Rev. 3) of the

Improved Standard Technical Specifications (ISTS), NUREG 1434,

``Standard Technical Specifications General Electric Plants, BWR/6.''

The SR was removed from the ISTS under Technical Specification Task

Force (TSTF) Traveler No. 2, ``Relocate the 10-Year Sediment Cleaning of the Fuel Oil Storage Tank to Licensee Control,'' approved by the

Nuclear Regulatory Commission on July 16, 1998.

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 FOSTs [fuel oil storage tanks] provide the storage for the

DG [diesel generator] fuel oil, assuring an adequate volume is available for each DG to operate for seven days in the event of a loss of offsite power concurrent with a loss of coolant accident.

The relocation of the SR to drain and clean the FOSTs to a licensee- controlled document will not impact any of the previously analyzed accidents. Sediment in the tank, or failure to perform this SR, does not necessarily result in an inoperable storage tank. Fuel oil quantity and quality are assured by other TS SRs that remain unchanged. These SRs help ensure tank sediment is minimized and ensure that any degradation of the tank wall surface that results in fuel oil volume reduction is detected and corrected in a timely manner. 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.

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

The proposed change does 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 change does not increase the types and amounts of radiological effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS change does not involve the addition or modification of any plant equipment. Also, the proposed change will

Page 4772

not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs. The proposed TS change does not create any new credible failure mechanisms, malfunctions or accident initiators.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not alter or exceed a design basis or safety limit. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the DGs are able to perform their intended function.

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

Attorney for licensee: Mr. Bradley J. Fewell, Associate General

Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

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

Beaver County, Pennsylvania

Date of amendment request: September 24, 2008.

Description of amendment request: The proposed amendment would modify Technical Specifications (TSs) to allow the BVPS-2 containment spray additive sodium hydroxide (NaOH) to be replaced by sodium tetraborate (NaTB).

Basis for proposed no significant hazards consideration determination: As required by10 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.

Use of NaTB in lieu of NaOH would not involve a significant increase in probability of a previously evaluated accident because the containment spray additive is not an initiator of any analyzed accident. The NaTB would be stored and delivered by a passive method that does not have potential to affect plant operations. Any existing NaOH delivery system equipment which remains in place but is removed from service would meet existing seismic, electrical and containment isolation requirements. Therefore the change in additive, including removal of NaOH equipment from service, would not result in any failure modes that could initiate an accident.

The spray additive is used to mitigate the consequences of a

LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu of NaOH would not involve a significant increase in the consequences of a previously evaluated accident because the amount of NaTB specified in the proposed TS would achieve a pH of 7 or greater, consistent with the current licensing basis. This pH is sufficient to achieve long-term retention of iodine by the containment sump fluid for the purpose of reducing accident related radiation dose following a LOCA.

Response: No.

Regarding the proposed use of NaTB in lieu of NaOH, the NaTB would be stored and delivered by a passive method that does not have potential to affect plant operations. Any existing NaOH delivery system equipment remaining in place but which is removed from service would meet existing seismic, electrical and containment isolation requirements. Hydrogen generation would not be significantly impacted by the change. Therefore, no new failure mechanisms, malfunctions, or accident initiators would be introduced by the proposed change and it would 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.

Since the quantity of NaTB specified in the amended TS would reduce the potential for undesirable chemical effects while achieving radiation dose reductions, corrosion control and hydrogen generation effects that are comparable to NaOH, the proposed change does not involve a significant reduction in a margin of safety. The primary function of an additive is to reduce loss of coolant accident consequences by controlling the amount of iodine fission products released to containment atmosphere from reactor coolant accumulating in the sump during a LOCA. Because the amended technical specifications would achieve a pH of 7 or greater using

NaTB, dose related safety margins would not be significantly reduced. Use of NaTB reduces the potential for undesirable chemical effects that could interfere with recirculation flow through the sump strainers. Any existing NaOH delivery system equipment which remains in place but is removed from service would meet existing seismic, electrical and containment isolation requirements and would not interfere with operation of the existing containment or containment spray system.

Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear

Operating Company, FirstEnergy Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Branch Chief: Mark G. Kowal.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50- 440, Perry Nuclear Power Plant, Unit No. 1 (PNPP), Lake County, Ohio

Date of amendment request: November 18, 2008

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 5.5.6 to incorporate Technical

Specification Task Force (TSTF) Travelers TSTF-479, ``Changes to

Reflect Revision of 10 CFR 50.55a,'' and TSTF 497, ``Limit Inservice

Testing Program SR [Surveillance Requirement] 3.0.2 Application to

Frequencies of 2 Years or Less.''

Response: No.

The proposed amendment revises TS 5.5.6, ``Inservice Testing

Program,'' for consistency with 10 CFR 50.55a(f)(4) requirements regarding inservice testing of pumps and valves. The proposed amendment incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not involve a modification to the physical configuration of the plant. There is no new equipment to be installed or a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new

Page 4773

accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual cumulative occupational exposure.

Therefore, the proposed change does not create the possibility of an accident of a different kind than previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment revises TS 5.5.6, ``Inservice Testing

Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves.

The proposed amendment incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy

Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Russell Gibbs.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: October 6, 2008.

Crystal River Unit 3 Nuclear Generating Plant.

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 remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26.

Response: No.

The proposed changes remove TS controls on working hours for personnel who perform safety related functions. The TS controls are superseded by the worker fatigue requirements in 10 CFR Part 26. The proposed changes do not involve any physical changes to plant or the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes will not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

Removal of plant-specific TS administrative requirements will not reduce a margin of safety because the requirements in 10 CFR

Part 26 are adequate to ensure that worker fatigue is managed.

Therefore, it is concluded that these changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: David T. Conley, Associate General Counsel

II--Legal Department, Progress Energy Service Company, LLC, Post Office

Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: December 17, 2008.

Description of amendments request: The proposed change would revise the Crystal River Unit 3 Improved Technical Specifications

Administrative Controls, Section 5.6, to revise the Inservice Testing

Program to incorporate the Technical Specification Task Force (TSTF)

Standard TS Change Traveler, TSTF-479, Revision 0, ``Changes to Reflect

Revision of 10 CFR 50.55a,'' and TSTF-497, Revision 0, ``Limit

Inservice Testing Program SR 3.0.2 Application to Frequencies of 2

Years or Less.''

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: 4. Does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change revises the CR-3 [Crystal River Unit 3] ITS

Improved Technical Specifications

, Section 5.6.2.9, ``Inservice

Testing Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and

Page 4774

valves which are classified as ASME [American Society of Mechanical

Engineers] Code Class 1, Class 2, and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, this proposed change does not involve an increase in probability or consequences of an accident previously evaluated. 5. Does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change revises the CR-3 ITS, Section 5.6.2.9,

``Inservice Testing Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or involve a change in the methods governing normal plant operation. The proposed change will not introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in types or increases in the amounts of any effluents that may be released offsite and there is no increase in individual or cumulative occupational exposure.

Therefore, the proposed change does not create the possibility of an accident of a different kind than previously evaluated. 6. Does not involve a significant reduction in a margin of safety[.]

The proposed change revises the CR-3 ITS, Section 5.6.2.9,

``Inservice Testing Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change the methods governing normal plant operation.

The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: David T. Conley, Associate General Counsel

II--Legal Department, Progress Energy Service Company, LLC, Post Office

Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50- 306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue

County, Minnesota

Date of amendment request: November 4, 2008.

Description of amendment request: The proposed amendments would make changes to the Technical Specifications to increase the 24 month test load for the Unit 1 Emergency Diesel Generators (EDGs), D1 and D2, reduce the monthly test load for the Unit 2 EDGs, D5 and D6, and reduce the 24 month test loads for the Unit 2 EDGs.

Response: No.

This license amendment request proposes to increase a portion of the Prairie lsland Nuclear Generating Plant Unit 1 emergency diesel generator's 24-month test loading, reduce the Unit 2 emergency diesel generators' monthly test loading which demonstrates Technical

Specification operability and revise the 24-month test to require the Unit 2 emergency diesel generators to operate for at least 2 hours at 100-110% of the continuous rated loading and the remainder of the 24-hour test at or above 4000 kW. The proposed test loads will continue to assure that the emergency diesel generators have the necessary reliability and availability for the design basis accidents and station blackout events.

The emergency diesel generators are required to be operable in the event of a design basis accident coincident with a loss of offsite power to mitigate the consequences of the accident. They are also the alternate AC source for a station blackout on the other

Prairie lsland Nuclear Generating Plant unit. The emergency diesel generators are not accident initiators and therefore these changes do not involve a significant increase in the probability of an accident previously evaluated.

The accident analyses assume that at least one safeguards bus is provided with power either from the offsite sources or the emergency diesel generators. The Technical Specification changes proposed in this license amendment request will continue to assure that the emergency diesel generators have the capacity and capability to assume their maximum auto-connected loads. Thus, the changes proposed in this license amendment request do not involve a significant increase in the consequences of an accident previously evaluated.

The changes proposed in this license amendment 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 to increase a portion of the Prairie Island Nuclear Generating Plant Unit 1 emergency diesel generator's 24-month test loading, reduce the Unit 2 emergency diesel generators' monthly test loading which demonstrates Technical

The proposed Technical Specification changes do not involve a change in the plant design, system operation, or the use of the emergency diesel generators. The proposed changes require the Unit 1 emergency diesel generators to be tested at increased loads and allow the Unit 2 emergency diesel generator to be tested at reduced loads which envelope the required safety function loads. These revised loads continue to demonstrate the capability and capacity of the emergency diesel generators to perform their required functions.

There are no new failure modes or mechanisms created due to testing the emergency diesel generators at the proposed test loading.

Testing of the emergency diesel generators at the proposed test loadings does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the proposed test loadings.

The Technical Specification changes proposed in this license amendment 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.

Page 4775

The proposed Technical Specification changes will continue to demonstrate that the emergency diesel generators meet the Technical

Specification definition of operability, that is, the proposed tests will demonstrate that the emergency diesel generators will perform their safety function and the necessary emergency diesel generator attendant instrumentation, controls, cooling, lubrication and other auxiliary equipment required for the emergency diesel generators to perform their safety function loads are also tested at these proposed loadings. The proposed testing will also continue to demonstrate the capability and capacity of the emergency diesel generators to supply their required loss of offsite power loads coincident with station blackout loads from the opposite unit. Since the proposed surveillance testing will continue to demonstrate operability, and the capability and capacity to supply their required loss of offsite power coincident with opposite unit station blackout loads, the proposed Technical Specification 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.

Attorney for licensee: Peter M. Glass, Assistant General Counsel,

Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: Lois M. James.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

Date of amendment request: March 27, 2008, as supplemented by a letter December 19, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) requirements related to control building envelope habitability in TS Section 3.7.3 Control Room

Emergency Ventilation (CREV) System, and add TS Section 5.5.13, Control

Building Envelope Habitability Program, to the Administrative Section of the TSs. The licensee has included conforming technical changes to the TS Bases. The proposed revision to the Bases also includes editorial and administrative changes to reflect applicable changes to the corresponding TS Bases, which were made to improve clarity, conform to the latest information and references, correct factual errors, and achieve more consistency with the standard TS NUREGs. The proposed revision to the TS and associated Bases is similar to the TSTF-448,

Revision 3. The supplement contains additional information related to smoke and chemical effects and addresses the associated proposed revision to TS Section 3.7.3, TS Section 5.5.13 and TS Bases 3.7.3.

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

No. 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. 2. Does the proposed Technical Specification change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. 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. 3. Does the proposed Technical Specification change involve a significant reduction in a 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.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Thomas H. Boyce.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear

Plant, Unit 2, Limestone County, Alabama

Date of amendment request: December 22, 2008 (TS-463-T).

Description of amendment request: The proposed amendment would, on a one-time basis, extend several Technical Specification (TS) surveillance frequencies approximately 45 days.

Response: No.

The requested action is a one-time extension to the performance interval of a limited number of TS surveillance requirements. The performance of these surveillances, or the failure to perform these surveillances, is not a precursor to an accident. Performing these surveillances or

Page 4776

failing to perform these surveillances does not affect the probability of an accident. Therefore, the proposed delay in performance of the surveillance requirements in this amendment request does not increase the probability of an accident previously evaluated.

A delay in performing these surveillances does not result in a system being unable to perform its required function. In the case of this one-time extension request, the relatively short period of additional time that the systems and components will be in service before the next performance of the surveillance will not affect the ability of those systems to operate as designed. Therefore, the systems required to mitigate accidents will remain capable of performing their required function. No new failure modes have been introduced because of this action and the consequences remain consistent with previously evaluated accidents. Therefore, the proposed delay in performance of the surveillance requirements in this amendment request does not involve a significant increase in the consequences of an accident.

Therefore, operation of the facility in accordance with the proposed license amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC) or a change in the way any

SSC is operated. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the one-time surveillance requirement extensions being requested.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment is a one-time extension of the performance interval of a limited number of TS surveillance requirements. Extending these surveillance requirements does not involve a modification of any TS Limiting Conditions for Operation.

Extending these surveillance requirements does not involve a change to any limit on accident consequences specified in the license or regulations. Extending these surveillance requirements does not involve a change to how accidents are mitigated or a significant increase in the consequences of an accident. Extending these surveillance requirements does not involve a change in a methodology used to evaluate consequences of an accident. Extending these surveillance requirements does not involve a change in any operating procedure or process.

The instrumentation and components involved in this request have exhibited reliable operation based on the results of the most recent performance of their 24-month surveillance requirements.

Based on the limited additional period of time that the systems and components will be in service before the surveillances are next performed, as well as the operating experience that these surveillances are typically successful when performed, it is reasonable to conclude that the margins of safety associated with these surveillance requirements will not be affected by the requested extension.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Thomas Boyce.

Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: August 1, 2008, as supplemented November 25 and December 31, 2008 (2 letters).

Description of amendment request: The proposed amendment would revise the following: (1) Technical Specification (TS) 4.2.1, ``Fuel

Assemblies,'' and TS Surveillance Requirements 3.5.1.4,

``Accumulators,'' and 3.5.4.3, ``RWST [Refueling Water Storage Tank],'' to increase the maximum number of Tritium Producing Burnable Absorber

Rods (TPBARs) that can be irradiated per cycle from 400 to 704.

An application that addressed similar issues was previously submitted on August 1, 2008, and notice of that application was provided in the Federal Register on November 12, 2008 (73 FR 66946).

Due to certain changes in the specifics of the December 31, 2008, revision from those proposed in the August 1, 2008, application, as supplemented on November 25 and December 31, 2008, the application is being renoticed in its entirety. This notice supersedes the notice published in the Federal Register on November 12, 2008.

Response: No.

The proposed change modifies the maximum number of TPBARs in the core. The required boron concentration for the cold leg accumulators

(CLAs) and RWST remains unchanged. The current boron concentration has been demonstrated to maintain the required accident mitigation safety function for the CLAs and RWST with the higher number of

TPBARs and this will be verified for each core that contains TPBARs as part of the normal reload analysis. The CLAs and RWST safety function is to mitigate accidents that require the injection of borated water to cool the core and to control reactivity. These functions are not potential sources for accident generation and the modification of the number of TPBARs will not increase the potential for an accident. Therefore, the possibility of an accident is not increased by the proposed changes. The current boron concentration levels are supported by the proposed number of TPBARs in the core.

Since the current boron concentration levels will continue to maintain the safety function of the CLAs and RWST in the same manner as currently approved, the consequences of an accident are not increased by the proposed changes. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change only modifies the maximum number of TPBARs in the core. The boron concentrations for accident mitigation functions of the CLAs and RWST remain unchanged. These functions do not have a potential to generate accidents as they only serve to perform mitigation functions associated with an accident. The proposed modification will maintain the mitigation function in an identical manner as currently approved. There are no plant equipment or operational changes associated with the proposed revision.

Therefore, since the CLA and RWST functions are not altered and the plant will continue to operate without change, the possibility of a new or different kind of an accident is not created. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

This change proposes a change to the maximum number of TPBARs in the core. The boron concentration requirements that support the accident mitigation functions of the CLAs and RWST remain unchanged.

The proposed change does not alter any plant equipment or components and does not alter any setpoints utilized for the actuation of accident mitigation system or control functions. The proposed number of TPBARs, in conjunction with the current boron concentration values, has been demonstrated to provide an adequate level of reactivity control for accident mitigation and this will be verified for each core that contains TPBARs as part of the normal reload analysis. Therefore, the proposed change will 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

Page 4777

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Acting Branch Chief: P. Milano.

Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

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

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,

Van Buren County, Michigan

Date of amendment request: November 25, 2008.

Brief description of amendment request: The proposed amendment would revise Appendix A, Technical Specifications (TS), as they apply to the spent fuel pool (SFP) storage requirements in TS section 3.7.16 and the criticality requirements for the Region I SFP and north tilt pit fuel storage racks, in TS section 4.3.1.1.

Date of publication of individual notice in Federal Register:

January 2, 2009 (74 FR 123).

Expiration date of individual notice: February 3, 2009.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: December 1, 2008.

Description of amendment request: By letter dated October 31, 2008, the Nuclear Regulatory Commission issued Amendment No. 186, to Callaway

Plant, Unit 1, Facility Operating License No. NPF-30. The amendment allowed a one-time extension of the allowed outage time (completion time) for each of the two essential service water (ESW) trains (ESW

Train A and Train B) from 72 hours to 14 days. The extended completion time was requested to support planned replacement of the underground carbon steel piping with new high density polyethylene (HDPE) piping for ESW Train A and ESW Train B during plant operation. The amendment was issued with a requirement to complete the replacement of carbon steel piping with HDPE piping for both ESW trains by December 31, 2008.

By its application dated December 1, 2008, the licensee informed NRC that it had experienced significant delays in completing the replacement of underground piping/conduit due, in part, to underground obstructions during excavation, a longer refueling outage (Refuel 16) than anticipated, a forced outage at the beginning of Cycle 17, switchyard maintenance, and other equipment and personnel issues.

However, the replacement of ESW Train A carbon steel piping was completed by the required date of December 31, 2008, but the replacement of ESW Train B carbon steel piping was deferred.

Consequently, the licensee proposed to extend the implementation date for completion of replacement of carbon steel piping for ESW Train B from December 31, 2008, to April 30, 2009.

Date of publication of individual notice in Federal Register:

December 23, 2008 (73 FR 78858).

Expiration date of individual notice comment period: January 22, 2009.

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.resource@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,

Unit No. 1, DeWitt County, Illinois

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island

Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of application for amendments: June 20, 2008.

Brief description of amendments: The amendments conform the licenses to reflect the direct transfer of AmerGen Energy Company,

LLC's ownership and operating authority for Clinton Power Station, Unit

No. 1, Oyster Creek Nuclear Generating Station (Oyster Creek), and

Three Mile Island Nuclear Station, Unit 1, to Exelon Generation

Company, LLC, (ECG) as approved by Commission Order dated December 23, 2008. Transfer of the license for Oyster Creek will also authorize EGC to store spent fuel in the Oyster Creek independent spent fuel storage installation.

Date of issuance: January 8, 2009.

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

Page 4778

Amendment Nos.: CPS-183, Oyster Creek-271, and TMI-1-267.

Facility Operating License Nos. NPF-62, DPR-16, and DPR-50: The amendments revised the Technical Specifications and Licenses.

Date of initial notice in Federal Register: August 26, 2008 (73 FR 50368). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 23, 2008.

No significant hazards consideration comments received: The NRC received three comments on August 27, 2008, one for each plant's initial notice. The comments did not provide any information additional to that in the application, nor did they provide any information contradictory to that provided in the application.

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

Station, Kewaunee County, Wisconsin

Date of application for amendment: April 4, 2008.

Brief description of amendment: The amendment revised the Technical

Specifications by removing the operability and surveillance requirements for the shield building ventilation (SBV) and auxiliary building special ventilation filter train heaters, and reducing the operating time required to verify the SBV system operability from 10 hours to 15 minutes.

Date of issuance: December 30, 2008.

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

Amendment No.: 201.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 3, 2008 (73 FR 31720) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 30, 2008.

No significant hazards consideration comments received: No.

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

Station, Kewaunee County, Wisconsin

Date of application for amendment: April 14, 2008, as supplemented by letter dated October 17, 2008.

Brief description of amendment: The amendment adds a new footnote to Kewaunee Technical Specifications Table 3.5-4, ``Instrument

Operating Conditions for Isolation Functions.'' The new footnote allows the main steam line isolation circuitry to be inoperable when both main steam isolation valves are closed and deactivated.

Date of issuance: January 12, 2009.

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

Amendment No.: 202.

Facility Operating License No. DPR-43: Amendment revised the operating license and Technical Specifications.

Date of initial notice in Federal Register: June 17, 2008 (73 FR 34340) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 12, 2009.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, et. al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: December 11, 2007, as supplemented December 18, 2008.

Brief description of amendments: The amendments revised the

Technical Specifications sections to allow the bypass test times and

Completion Times (CTs) for Limiting Condition for Operation (LCOs) 3.3.1, ``Reactor Trip System (RTS) Instrumentation;'' 3.3.2,

``Engineered Safety Feature Actuation System (ESFAS) Instrumentation;'' 3.3.6, ``Containment Air Release and Addition Isolation

Instrumentation,'' and 3.3.9, ``Boron Dilution Mitigation System

(BDMS).''

The proposed license amendment request (LAR) adopts changes as described in Westinghouse Commercial Atomic Power (WCAP) topical report

WCAP-14333-P-A, Revision 1, ``Probabilistic Risk Analysis of the

Reactor Protection System and Engineered Safety Features Actuation

System Test Times and Completion Times,'' issued October 1998 and approved by U.S. Nuclear Regulatory Commission (NRC) letter dated July 15, 1998. Implementation of the proposed changes is consistent with

Technical Specification Task Force (TSTF) Traveler TSTF-418, Revision 2, ``RPS [Reactor Protection System] and ESFAS Test Times and

Completion Times (WCAP-14333).'' The NRC approved TSTF-418, Revision 2, by letter dated April 2, 2003.

In addition, the proposed LAR adopts changes as described in WCAP- 15376-P-A, Revision 1, ``Risk-Informed Assessment of the RTS and ESFAS

Surveillance Test Intervals and Reactor Trip Breaker Test and

Completion Times,'' issued March 2003, as approved by NRC letter dated

December 20, 2002. Implementation of the proposed changes is consistent with TSTF Traveler TSTF-411, Revision 1, ``Surveillance Test

Interval Extension for Components of the Reactor Protection System

(WCAP-15376).'' The NRC approved TSTF-411, Revision 1, by letter dated

August 30, 2002. The licensee also requested additional changes not specifically included in the above topical reports. These changes will be evaluated in a future amendment.

Date of issuance: December 22, 2008.

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

Amendment Nos.: 247 and 240.

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

Date of initial notice in Federal Register: March 25, 2008 (73 FR 15783). The supplement dated December 18, 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 December 22, 2008.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, et. al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: December 11, 2007, as supplemented by letter dated December 18, 2008.

Brief description of amendments: The amendments revised the

Technical Specification sections to allow the bypass test times and

Completion Times for Limiting Condition for Operation 3.3.1, ``Reactor

Trip System (RTS) Instrumentation'' and 3.3.2, ``Engineered Safety

Feature Actuation System (ESFAS) Instrumentation.''

By letter dated December 30, 2008 (Agencywide Documents Access and

Management System Accession No. ML0083460216), the NRC issued Amendment

No. 247 and Amendment No. 240 for Catawba Units 1 and 2, respectively, for all the proposed changes approved by the NRC in TSTFs 411 and 418.

The December 30, 2008, amendment stated that the following changes would be evaluated in a future amendment:

Surveillance requirement (SR) 3.3.1.5, Safety injection input from

ESFAS, Condition J, Feedwater isolation with low average core temperature coincident with reactor trip P-4, SR 3.3.2.2, turbine

Page 4779

trip and feedwater isolation for steam generator water level high high.

(P-14), SR 3.3.2.4 turbine trip and feedwater isolation for steam generator water level high high (P-14), and SR 3.3.2.5 turbine trip and feedwater isolation for low average core temperature trip coincident with reactor trip P-4.

This amendment approves the above changes.

Date of issuance: January 9, 2009.

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

Amendment Nos.: 248 and 241.

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

Date of initial notice in Federal Register: March 25, 2008 (73 FR 15783). The supplement dated December 18, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 9, 2009.

No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,

Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,

Vermont

Date of application for amendment: February 6, 2008, as supplemented by letter dated July 29, 2008.

Brief description of amendment: The amendment revised the

Surveillance Requirements (SRs) for control rod exercising from weekly to monthly in Technical Specification (TS) 4.3.A.2, revise verification of control rod coupling integrity as described in TS 4.3.B.1, revise the scram insertion time Limiting Conditions for Operation (LCOs) and

SRs as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, the LCO and SR for Control Rod Accumulators.

Date of issuance: January 7, 2009.

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

Amendment No.: 233.

Facility Operating License No. DPR-28: Amendment revised the

License and Technical Specifications.

Date of initial notice in Federal Register: March 11, 2008 (73 FR 13024). The supplemental letter dated July 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 this amendment is contained in a Safety Evaluation dated January 7, 2009.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: July 30, 2008, as supplemented by letter dated October 2, 2008.

Brief description of amendment: The amendment revises the current

TS 3.6.6.3 surveillance requirements for sodium hydroxide (NaOH) concentration. Specifically, the amendment changes the surveillance requirements of the NaOH tank solution concentration from between 5.0 weight (wt.) percent and 16.5 wt. percent to between 6.0 wt. percent and 8.5 wt. percent.

Date of issuance: January 13, 2009.

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

Amendment No.: Unit 1--234.

Renewed Facility Operating License No. DPR-51: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: November 4, 2008, (73

FR 65694). The supplement dated October 2, 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 amendment is contained in a Safety Evaluation dated January 13, 2009.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: June 30, 2008.

Brief description of amendment: The amendment (1) deleted Technical

Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR 3.1.3.3; (2) removed the reference to SR 3.1.3.2 from Required Action

A.2 of TS 3.1.3, ``Control Rod OPERABILITY''; (3) clarified the requirement to fully insert all insertable rods for the limiting condition for operation in TS 3.3.1.2 Required Action E.2, ``Source

Range Monitoring Instrumentation''; and (4) revised Example 1.4-3 in

Section 1.4, ``Frequency,'' to clarify the applicability of the 1.25 surveillance test interval extension.

Date of issuance: December 31, 2008.

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

Amendment No: 180.

Facility Operating License No. NPF-29: The amendment revised the

Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 26, 2008 (73 FR 50359).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 31, 2008.

No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3, Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: January 17, 2008.

Brief description of amendment: The amendment revises the Crystal

River, Unit 3 Improved Technical Specification Surveillance Requirement 3.7.5.2, ``Emergency Feedwater System,'' to align the text for the emergency feedwater system surveillance frequency with the text in the

Technical Specifications Task Force Standard Technical Specification

Change Traveler-101, Revision 0 and the NRC technical report, NUREG- 1430, Volume 1, Revision 3, ``Standard Technical Specifications Babcock and Wilcox Plants--Specification.''

Date of issuance: January 9, 2009.

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

Amendment No.: 231.

Facility Operating License No. DPR-72: Amendment revises the technical specifications.

Date of initial notice in Federal Register: May 20, 2008 (73 FR 29163).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 9, 2009.

No significant hazards consideration comments received: No.

Page 4780

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of application for amendments: December 17, 2007, as supplemented by letters dated October 2, and November 18, 2008.

Brief description of amendments: The amendments increase the completion times (CTs) for required actions related to Technical

Specifications (TS) 3.5.2, regarding the Emergency Core Cooling System, and 3.6.6, regarding the Containment Spray and Cooling Systems from 72 hours to 14 days. In addition, invalid notes were deleted from TSs 3.5.2 and 3.6.6 and new notes were added to specify the limitations on the use of the 14-day extended CT.

Date of issuance: December 31, 2008.

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

Amendment Nos.: Unit 1--202; Unit 2--203.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: January 29, 2008 (73 FR 5227). The supplement(s) dated October 2 and November 18, 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 December 31, 2008.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam

Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendments: July 7, 2008.

Brief description of amendments: The amendments revised the

Technical Specification (TS) testing frequency for the Surveillance

Requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The change revised the frequency of SR 3.1.4.2, control rod scram time testing, from ``120 days cumulative operation in Mode 1'' to ``200 days cumulative operation in Mode 1.'' These changes are based on TS Task

Force (TSTF) change traveler TSTF-460 (Revision 0) that has been approved generically for the Boiling-Water Reactor (BWR) Standard TS,

NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency of

SR 3.1.4.2, control rod scram time testing, from ``120 days cumulative operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''

Date of issuance: January 2, 2009.

Effective date: January 2, 2009.

Amendment Nos.: 249 for Unit 1 and 228 for Unit 2

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

Date of initial notice in Federal Register: October 7, 2008 (73 FR 58675).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 2, 2009.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam

Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendments: July 7, 2008.

Brief description of amendments: The amendment adopted the Nuclear

Regulatory Commission (NRC) approved Technical Specification Task Force

(TSTF) change traveler TSTF-475, (Revision 1), ``Control Rod Notch

Testing Frequency and SRM [Source Range Monitor] Insert Control Rod

Action,'' to change the Standard Technical Specifications (STS) for

General Electric (GE) Plants (NUREG-1433, BWR/4 to the plant-specific

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 the LPSP of RWM'' to ``31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP [Low Power Set Point] of the

RWM [Rod With Minimizer]'', 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.

Date of issuance: January 2, 2009.

Effective date: January 2, 2009.

Amendment Nos.: 250 for Unit 1 and 229 for Unit 2.

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

Date of initial notice in Federal Register: October 7, 2008 (73 FR 58675).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 2, 2009.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of application for amendment: September 19, 2008.

Brief description of amendment: The amendment modifies the Final

Safety Analysis Report by requiring an inspection of the ice condenser within 24 hours of experiencing a seismic event greater than or equal to an operating basis earthquake within the 5-week period after ice basket replenishment has been completed to confirm that adverse ice fallout has not occurred that could impede the ability of the ice condenser lower inlet doors to open.

Date of issuance: January 6, 2009.

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

Amendment No.: 73.

Facility Operating License No. NPF-90: Amendment authorizes revision to the FSAR.

Date of initial notice in Federal Register: November 4, 2008 (73 FR 65698).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 6, 2009.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of application for amendment: March 27, 2008, as supplemented

September 26, 2008.

Brief description of amendment: The amendment revises the allowable value listed for Function 3, ``Containment Purge Exhaust Radiation

Monitors,'' in Table 3.3.6-1, ``Containment Vent Isolation

Instrumentation,'' of the limited condition for operation 3.3.6.

Date of issuance: January 8, 2009.

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

Amendment No.: 74.

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

Technical Specifications and License.

Date of initial notice in Federal Register: May 6, 2008 (73 FR 25047). The supplement dated September 26,

Page 4781

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 amendment is contained in a Safety Evaluation dated January 8, 2009.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Date of application for amendment: December 17, 2007, as supplemented on July 22, 2008, September 26, 2008, and November 25, 2008.

Brief description of amendment: These amendments revised Technical

Specification (TS) 3.8.3 to allow a one-time extended 14-day completion time (CT) for each of the two underground diesel fuel oil storage tanks

(FOST) to permit removal of the current coating and to recoat the tanks in preparation for use of ultra-low sulfur diesel fuel oil. The change revised the TS to extend the CT associated with an inoperable emergency diesel generator FOST from 7 days to 14 days, applicable once for each of the two tanks.

Date of issuance: December 31, 2008.

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

Amendment Nos.: 254 and 235.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the licenses and the technical specifications.

Date of initial notice in Federal Register: January 15, 2008 (73 FR 2552). The supplements dated July 22, 2008, September 26, 2008, and

November 25, 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 December 31, 2008.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 15th day of January 2009.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

Director, Division of Operating Reactor Licensing, Office of Nuclear

Reactor Regulation.

FR Doc. E9-1568 Filed 1-26-09; 8:45 am

BILLING CODE 7590-01-P

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Applications and Amendments to Facility Operating Licenses: Biweekly Notice

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