73 FR 223 pgs. 68451-68460 - Biweekly Notice; Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

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Type: NOTICEVolume: 73Number: 223Pages: 68451 - 68460

FR document: [FR Doc. E8-27110 Filed 11-17-08; 8:45 am]

Agency: Nuclear Regulatory Commission

Official PDF Version: PDF Version

NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from October 23, 2008, to November 5, 2008. The last biweekly notice was published on November 4, 2008 (73 FR 65685).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant HazardsConsideration 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.

Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, person(s) may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request via electronic submission through the NRC E-Filing system for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's "Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms 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 confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

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

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

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

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

For further details with respect to this amendment action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr.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: October 1, 2008.

Description of amendment request: The amendment would modify Technical Specification (TS) 5.5.16, Containment Leakage Rate Testing Program, by adding exceptions to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," that would allow the next integrated leak rate test (ILRT) (Type A test) to be performed at a 15-year interval at Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3. The proposed amendment is risk-informed and follows the guidance in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."

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

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

Response: No.

The proposed change to extend the next ILRT interval from 10 to 15 years one time does not involve a physical change to PVNGS[,] Units 1, 2, and 3, or a change in the manner in which the plant is operated or controlled. The containment vessel is designed to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment for any postulated accidents. As such, the reactor containment itself and the testing guidelines invoked to periodically demonstrate the integrity of the containment exist to ensure the containment can mitigate the consequences of any accident and do not involve the prevention or identification of any precursors of any accidents. There is no design basis accident that is initiated by a failure of the containment leakage mitigation function. The extension of the ILRT will not create any adverse interactions with other systems that could result in initiation of a design basis accident. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

Based on a completed probability risk assessment of the affects of this change to the ILRT interval there is a slight increase in risk dose. This increase in risk in terms of person-rem year within 50 miles of the plant resulting from design basis accidents is significantly less than one percent and of a magnitude that NUREG-1493 indicates is imperceptible. The risk assessment also analyzed the increase in risk in terms of the frequency of large early releases from accidents. The increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in Regulatory Guide 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. The increase in the conditional containment failure probability from reducing the ILRT frequency from one test per 10 years to one test per 15 years is less than one percent and considered insignificant. Continued containment integrity is assured by the history of successful ILRTs, and the established programs for local leakage rate testing and in-service inspections which are not affected by the proposed change. Therefore, the consequences of an accident previously analyzed are not significantly increased.

In summary, the probability of occurrence and the consequences of an accident previously evaluated are not significantly increased.

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

Response: No.

The proposed change to extend the ILRT interval from 10 to 15 years does not create any new or different accident initiators or precursors. The length of the ILRT interval does not affect the manner in which any accident begins. The proposed change does not physically change the plant, does not create any new failure modes for the containment and does not affect the interaction between the containment and any other system. Thus, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The risk-based margins of safety associated with the containment ILRT are those associated with the estimated person-rem per year, the large early release frequency, and the conditional containment failure probability. The potential effect of the proposed change on the parameters have been quantified and it has been determined that the effect is considered insignificant. The non-risk-based margins of safety associated with the containment ILRT are those involved with its structural integrity and leak tightness. The proposed change to extend the ILRT interval from 10 to 15 years does not adversely affect either of these attributes. The proposed change only affects the frequency at which these attributes are verified. Therefore, the proposed change does not involve a significant reduction in margin of safety.

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.

Entergy Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point Plant, Charleviox County, Michigan

Date of amendment request: September 22, 2008.

Description of amendment request: The proposed amendment would amend the facility operating license by changing the names of the licensees from Entergy Nuclear Operations, Inc., and Entergy Nuclear Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades, LLC, respectively.

The proposed amendment would only change the names of the licensees and reflect associated order requirements. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes do not create the possibility of a new or different kind of accident from an accident previously evaluated. The proposed changes do not involve a significant reduction in margin of safety.

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

Attorney for licensee: Mr. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: Lois M. James.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-286, Indian Point Nuclear Generating Unit Nos. 1, 2 and 3, Westchester County, New York

Date of amendment request: September 30, 2008 (2 letters).

Description of amendment request: This is an administrative change which would reflect the creation of new companies as approved by the NRC Order dated July 28, 2008. The amendments would not be implemented until the restructuring transactions have been completed. The amendments would revise the names on the plant licenses to match the names of the new companies. Entergy Nuclear Indian Point 2, LLC would be changed to Enexus Nuclear Indian Point 2, LLC. Entergy Nuclear Indian Point 3, LLC would be changed to Enexus Nuclear Indian Point 3, LLC. Entergy Nuclear Operations, Inc. would be changed to EquaGen Nuclear LLC.

The proposed amendment would only change the names of the licensees and reflect associated order requirements. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: September 22, 2008.

Description of amendment request: The proposed amendment would amend the renewed facility operating license and Technical Specifications Design Features, Section 4, by changing the names of the licensees from Entergy Nuclear Operations, Inc. and Entergy Nuclear Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades, LLC, respectively.

Attorney for licensee: Mr. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: Lois M. James.

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

Date of amendment request: September 22, 2008.

Description of amendment request: The proposed amendment would relocate the contents of the Vermont Yankee (VY) Technical Specification (TS) relating to the Reactor Building crane to the VY Technical Requirements Manual (TRM).

1. The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

This proposed change relocates the VY TS and associated Bases related to the Reactor Building crane to the VY TRM. The proposed amendment does not impact the operability of any structure, system or component that affects the probability of an accident or that supports mitigation of an accident previously evaluated. The proposed amendment does not affect reactor operations or accident analysis and has no radiological consequences. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

This proposed change relocates the VY TS and associated Bases related to the Reactor Building crane to the VY TRM. The proposed amendment does not change the design or function of any component or system. No new modes of failure or initiating events are being introduced. Therefore, operation of VY in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

This proposed change relocates the VY TS and associated Bases related to the Reactor Building crane to the VY TRM. The proposed amendment does not change the design or function of any component or system. The proposed amendment does not involve any safety limits, safety settings or safety margins. The ability of the Reactor Building crane to perform its intended functions will continue to be required in accordance with the VY TRM.

Since the proposed controls are adequate to ensure the operability of the Reactor Building crane, there will still be high assurance that the components are operable and capable of performing their respective functions. Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in [a] margin to safety.

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: September 22, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) to change requirements related to Battery Systems specified in TS Section 3.10 resulting in removing the Limiting Condition for Operation pertaining to 345 kV switchyard batteries, chargers and associated direct current distribution panel.

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

Response: No.

The change does not impact the function of any Structure, System or Component (SSC) that affects the probability of an accident or that supports mitigation or consequences of an accident previously evaluated. The proposed change removes unnecessary information from the Technical Specifications that is not required in accordance with 10 CFR 50.36. The proposed change does not affect any plant equipment operation or accident analysis and has no radiological consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety related system performs their function. The proposed change removes unnecessary information from the Technical Specifications that is not required in accordance with 10 CFR 50.36. As such, no new or different types of equipment will be installed or removed from the facility. Operation of existing installed equipment is unchanged. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

This change does not change any existing design or operational requirements and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant operation as a result of the proposed change. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: June 3, 2008.

Description of amendment request: The proposed amendment would revise the analysis methodology in the Final Safety Analysis Report, Section 5.4.3, "Structural Design Criteria," and Section 5.4.5.3, "Missile Analysis." The amendment would allow the licensee to use the yield line theory methodology to qualify the east wall of the Auxiliary Building for tornado wind and missile loading.

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

The proposed LAR [license amendment request] will revise the methodology used to qualify the east wall of the CR-3 [Crystal River Unit 3 Nuclear Generating Plant] Auxiliary Building for all expected and postulated loads including tornado wind and missile loading. The Yield Line Theory methodology is an industry standard that is used for the design and analysis of concrete slabs. The Yield Line Theory methodology is used for investigating the failure mechanisms of flat reinforced concrete slabs at the ultimate limit (failure point). In other words, this methodology determines either the moments in a slab at the point of failure or the load at which the slab will fail. A change in the methodology of an analysis used to verify qualification of an existing structure will not have any impact on the probability of accidents previously evaluated.

The analysis performed demonstrates that the CR-3 Auxiliary Building east wall will remain structurally intact following the worst case loadings assumed in the calculation. Therefore, this proposed change does not involve a significant increase in the probability or consequences previously evaluated.

2. Does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The function of the CR-3 Auxiliary Building wall is to house and protect the equipment that is important to safety from damage during normal operation, transients and design basis accidents. The use of the Yield Line Theory methodology for qualifying the east wall of the CR-3 Auxiliary Building has no impact on the capability of the structure. A calculation that uses the Yield Line Theory methodology demonstrated that the structure meets required design criteria. This ensures that the wall is capable of performing its design function without alteration or compensatory actions of any kind. No changes to any plant system, structure, or component (SSC) are proposed. No changes to any plant operating practices, procedures, computer firmware/ software will occur.

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

3. Does not involve a significant reduction in a margin on safety.

The design basis of the plant requires structures to be capable of withstanding normal and accident loads including those from a design basis tornado. The Yield Line Theory methodology, as applied in an approved plant calculation, has demonstrated that the east wall of the CR-3 Auxiliary Building is capable of performing its design function. There is a slight reduction in conservatism between the method used for the remaining Class 1 structures, American Concrete Institute (ACI) standard 318-63 and the Yield Line Theory methodology, but the calculation performed with the Yield Line Theory methodology validates the requirement that the east wall of the CR-3 Auxiliary Building will protect the important to safety SSCs located in proximity to the wall from damage.

ACI 318-63 utilizes conservative methods, and due to the assumptions and technique, results in a Code defined value for strength that is not the maximum limit. The Yield Line Theory methodology uses assumptions and techniques that will define the failure point. However, the calculation performed for the east wall of the CR-3 Auxiliary Building demonstrates that there is margin to this "failure point," and the strength of the wall exceeds the applied loads, including the tornado wind and pressure drop loads, and will not fail due to tornado missile impact.

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

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

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

NRC Branch Chief: Thomas H. Boyce.

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

Date of amendment request: July 31, 2008.

Description of amendment request: The proposed amendments would change the PPL Susquehanna, LLC (PPL) Units 1 and 2 Technical Specification (TSs) 3.6.1.3 "Primary Containment Isolation Valves (PCIVs)." It proposes to revise the Secondary Containment Bypass Leakage (SCBL) limit in Surveillance Requirement 3.6.1.3.11 from "less than or equal to 9 standard cubic foot/feet per hour (scfh)" to "less than or equal to 15 scfh when pressurized to greater than or equal to P a ."

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

Response: No.

The structures, systems and components affected by the proposed change act as mitigators to the consequences of accidents. These components are not initiators of any accident analyzed in the Final Safety Analysis Report (FSAR). As such, the proposed change does not increase the probability of an accident previously evaluated. Based on the revised analysis, the proposed change does revise the performance requirement; however, the proposed change does not involve a revision to the parameters or conditions that could contribute to the initiation of a DBA [design-basis accident] discussed in Chapter 15 of the FSAR.

Plant-specific radiological analysis has been performed using the increased Secondary Containment Bypass Leakage (SCBL) limit. This analysis demonstrates that the CRHE [control room habitability envelope] dose consequences meet the regulatory guidance provided for use with the Alternative Source Term (AST), and the offsite doses are well within acceptable limits (10 CFR 50.67, Regulatory Guide (RG) 1.183, and Standard Review Plan Section (SRP) 15.0.1).

Therefore, the proposed amendment does not result in a significant increase in the consequences of any previously evaluated accident.

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

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

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

Response: No.

The results of the revised accident analysis are subject to the acceptance criteria in 10 CFR 50.67. The revised Secondary Containment Bypass Leakage rate limit is used in the LOCA [loss-of-coolant accident] radiological analysis. The analysis has been performed using conservative methodologies. Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed LOCA event has been carefully selected and margin has been retained to ensure that the analysis adequately bounds postulated event scenarios. The dose consequences of the limiting event is within the acceptance criteria presented in 10 CFR 50.67, RG 1.183, and SRP 15.0.1. The effect of the revision to the Technical Specification requirements has been analyzed and doses resulting from the pertinent design basis accident have been found to remain within regulatory limits. The change continues to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief : Mark Kowal.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of amendment request: July 18, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) 2.1.1.2 to decrease the safety limit minimum critical power ratio (SLMCPR) from 1.11 to 1.09 for single recirculation loop operation and from 1.09 to 1.07 for two recirculation loop operation.

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

No. The proposed amendment establishes a revised SLMCPR value for single and two recirculation loop operation. The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed SLMCPR values preserve the existing margin to transition boiling and the probability of fuel damage is not increased. Since the change does not require any physical plant modifications or physically affect any plant components, no individual precursors of an accident are affected and the probability of an evaluated accident is not increased by revising the SLMCPR values.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. The revised SLMCPR values have been determined using NRC-approved methods and procedures. The basis of the MCPR Safety Limit is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. These calculations do not change the method of operating the plant and have no effect on the consequences of an evaluated accident. Therefore, the proposed TS change does not involve an 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 license amendment involves a revision of the SLMCPR value for single and two recirculation loop operation based on the results of an analysis of the Unit 1 Cycle 8 core. Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in the allowable methods of operating the facility. This proposed license amendment does not involve any modifications of the plant configuration or changes in the allowable methods of operation. Therefore, the proposed TS 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?

No. The margin of safety as defined in the TS bases will remain the same. The new SLMCPR values were calculated using referenced fuel vendor methods and procedures, which are in accordance with the fuel design and licensing criteria. The SLMCPR remains high enough to ensure that greater than 99.9 percent of all fuel rods in the core are expected to avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. Therefore, the proposed TS change does not involve a reduction in the 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.

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

Date of application for amendment: September 27, 2007, as supplemented by letter dated September 5, 2008.

Brief description of amendment: The amendment modified the technical specifications (TS) by relocating references to specific American Society for Testing and Materials standards for fuel oil testing to licensee-controlled documents as part of the implementation of Technical Specification Task Force (TSTF) Traveler No. 374. This proposed change to the standard technical specifications was submitted by the TSTF in TSTF-374, "Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil," and is applicable to all nuclear power reactors.

Date of issuance: October 30, 2008.

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

Amendment No.: 182.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register : December 18, 2007 (72 FR 71705). The September 5, 2008 supplement, contained clarifying information, 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 amendment is contained in a Safety Evaluation dated October 30, 2008.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut

Date of application for amendment: March 25, 2008, as supplemented by letter dated September 30, 2008.

Brief description of amendment: The amendment revises the reactor coolant system (RCS) specific activity to utilize a new indicator, Dose Equivalent Xenon-133 and only take into account the noble gas activity in the primary coolant, instead of using the average disintegration energy (E Bar). Specifically, the Technical Specification 3.4.8, "Specific Activity," limit on RCS gross specific activity has a new limit on RCS noble gas specific activity. The changes are based on Technical Specification Task Force (TSTF) change traveler TSTF-490, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech. Spec. [Technical Specification]."

Date of issuance: October 27, 2008.

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

Amendment Nos.: 307 and 246.

Renewed Facility Operating License Nos. DPR-65 and NPF-49: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : July 29, 2008 (73 FR 43955-43956). The supplement dated September 30, 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 amendment is contained in a Safety Evaluation dated October 27, 2008.

No significant hazards consideration comments received: No.

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

Date of application of amendments: October 16, 2007, as supplemented May 7, September 2 and October 23, 2008.

Brief description of amendments: The amendments revised the Technical Specifications to accommodate plant modifications that address water hammer concerns described in Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Conditions," dated September 30, 1996.

Date of Issuance: October 29, 2008.

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

Amendment Nos.: 363, 365, 364.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : November 20, 2007 (72 FR 65364). The supplements dated May 7, September 2 and October 23, 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 October 29, 2008.

No significant hazards consideration comments received: No.

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

Date of application of amendments: October 22, 2007, supplemented July 14, September 17, and October 27, 2008.

Brief description of amendments: The amendments revised the Technical Specifications related to accommodate the use of AREVA NP Mark-B-HTP fuel.

Date of Issuance: October 29, 2008.

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

Amendment Nos.: 362, 364, 363.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : November 20, 2007 (72 FR 65365). The supplements dated July 14, September 17, and October 27, 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 October 29, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 13, 2007, as supplemented by letter dated July 10, 2008.

Brief description of amendment: The amendment revises the Technical Specifications (TSs) by adding three Emergency Core Cooling System (ECCS) valves and removing four ECCS valves from a TS surveillance requirement for checking valve position every 7 days.

Date of issuance: October 29, 2008.

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

Amendment No.: 256.

Facility Operating License No. DPR-26: The amendment revised the License and the TSs.

Date of initial notice in Federal Register : March 25, 2008 (73 FR 15784). The July 10, 2008, supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register .

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendment: October 19, 2007, supplemented by letters dated March 14, 2008, March 26, 2008, and July 18, 2008.

Brief description of amendment: The amendments consist of changes to the technical specifications of each unit, increasing the allowed surveillance interval for local power range monitor calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH.

Date of issuance: October 28, 2008.

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

Amendment Nos.: 195 and 156.

Facility Operating License Nos. NPF-39 and NPF-85. These amendments revised the license and the technical specifications.

Date of initial notice in Federal Register : July 8, 2008 (73 FR 39055). The supplements dated March 14, 2008, March 26, 2008 and July 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 determination.

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

No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: April 4, 2008, as supplemented by letter dated August 6, 2008.

Brief description of amendment: The amendment revised the Technical Specifications by adding a new Limiting Condition for Operation (LCO), LCO 3.0.9. This LCO establishes conditions under which systems would remain operable when required physical barriers are not capable of providing their related support function. This amendment is consistent with approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-427, Revision 2.

Effective date: As of the date of issuance and shall be implemented within 90 days following startup from the 2009 Refueling Outage.

Amendment No.: 157.

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

Date of initial notice in Federal Register : September 9, 2008 (73 FR 52418).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 22, 2008.

No significant hazards consideration comments received: None.

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

Date of application for amendments: October 29, 2007, as supplemented by letters dated April 24 and June 13, 2008.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plants, Units 1 and 2. The amendments revise TS 3.8.1 "AC Sources-Operating" by revising Surveillance Requirement 3.8.1.9 to require the emergency diesel generator 24-hour load test be performed at or below a power factor of 0.85.

Date of issuance: October 21, 2008.

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

Amendment Nos.: 189, 178.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register : December 18, 2007 (72 FR 71713). The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in Safety Evaluation dated October 21, 2008.

No significant hazards consideration comments received: No.

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: October 15, 2007, as supplemented by letter dated July 8, 2008.

Brief description of amendments: The amendments relocate surveillance frequencies of most surveillance tests from the Technical Specifications (TS) to a licensee-controlled document, the Surveillance Frequency Control Program (SFCP). Once relocated, changes to the surveillance frequencies may be made using a risk-informed methodology, Nuclear Energy Institute (NEI) document NEI 04-10 Rev. 1, as specified in the Administrative Controls of the TS. The NRC staff has previously approved NEI 04-10 Rev. 1, as acceptable for referencing in licensing applications.

Date of issuance: October 30, 2008.

Effective date: As of its date of issuance and shall be implemented within 360 days from the date of issuance.

Amendment Nos.: Unit 1-200; Unit 2-201.

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: November 20, 2007 (72 FR 65370). The supplement dated July 8, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register .

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

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: January 17, 2008, as supplemented August 15, 2008.

Brief description of amendment: The amendment will strengthen the control room envelope habitability requirements, adds a new administrative controls program, and adds an additional condition as described in Technical Specification Task Force traveler 448, Revision 3, "Control Room Habitability."

Date of issuance: October 27, 2008.

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

Amendment No.: 180.

Renewed Facility Operating License No. NPF-12: Amendment revises the Appendix A Technical Specifications and the Appendix C Additional Conditions.

Date of initial notice in Federal Register: February 12, 2008 (73 FR 8071). The supplement dated August 18, 2008, provided clarifying information that did not change the scope of the January 17, 2008, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 27, 2008.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 23, 2007, as supplemented by letter dated May 20, 2008.

Brief description of amendments: The amendments revised the Technical Specifications (TS) to relocate surveillance frequencies of most surveillance tests from the TS to a licensee-controlled surveillance frequency control program (SFCP). Once relocated, the surveillance frequency changes are permitted based on the risk-informed methodology as specified in the Administrative Controls section of the TS.

Date of issuance: October 31, 2008.

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

Amendment Nos.: Unit 1-188; Unit 2-175.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: December 18, 2007 (72 FR 71716). The supplemental letter dated May 20, 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 October 31, 2008.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee

Date of amendment request: October 26, 2007.

Description of amendment request: The amendments modify the Technical Specifications (TSs) to establish more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope in accordance with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification change traveler TSTF-448, Revision 3, "Control Room Habitability." Specifically, the amendments modify TS 3.7.7, "Control Room Emergency Ventilation System" and TS Section 6, "Administrative Controls." The amendments also add a new license condition regarding initial performance of the new surveillance and assessment requirements of the revised TSs.

Date of issuance: October 28, 2008.

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

Amendment Nos: 321 and 313.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the license and the TSs.

Date of initial notice in Federal Register: December 4, 2007 (72 FR 68219).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 31, 2007, as supplemented by letters dated February 21, March 7, April 17, May 6, July 10, and August 13, 2008.

Brief description of amendment: The amendment revises Technical Specifications to extend for one time the Completion Times for both essential service water trains and the emergency diesel generators from 72 hours to 14 days. The revision to TS would apply when each train of ESW system is inoperable during respective ESW system piping replacements.

Date of issuance: October 31, 2008.

Effective date: As of its date of issuance and shall be implemented by December 31, 2008.

Amendment No.: 186.

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

Date of initial notice in Federal Register: December 31, 2008 (72 FR 74362). The supplements dated February 21, March 7, April 17, May 6, July 10, and August 13, 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 October 31, 2008.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 6th day of November 2008.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

[FR Doc. E8-27110 Filed 11-17-08; 8:45 am]

BILLING CODE 7590-01-P