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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from October 9, 2008 to October 22, 2008. The last biweekly notice was published on October 21, 2008 (73 FR 370501).

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 Start Printed Page 65686considered 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 6D44, 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). Start Printed Page 65687

Each petitioner/requestor will need to download the Workplace Forms ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM is free and is available at http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html. Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html.

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

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

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

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

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

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

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

Date of amendment request: September 2, 2008.

Description of amendment request: The proposed amendment would relocated surveillance requirement (SR) 3.8.3.6 from the technical specifications (TSs) to a licensee-controlled document. SR 3.8.3.6 requires the emergency diesel generator fuel oil storage tanks to be drained, sediment removed, and cleaned on a 10-year interval.

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] DG 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 a 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 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 an accident previously evaluated. Further, the proposed change does not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. Start Printed Page 65688

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 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.

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 change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: August 29, 2008.

Description of amendment request: The amendments would modify Technical Specification (TS) 5.6.5, Core Operating Limits Report (COLR), by updating TS 5.6.5b to reflect the current analytical methods used to determine the core operating limits in Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3. The proposed amendment is an administrative change and all of the analytical methods have been previously reviewed and approved by the Nuclear Regulatory Commission (NRC).

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 the list of methodologies used at PVNGS [PVNGS, Units 1, 2, and 3] to determine the various COLR limits is an administrative change which updates the list in the TS to include NRC reviewed and approved COLR methodologies for PVNGS. It does not add or modify any previously used methodologies; it updates the list to include those already approved for use. This change does not make any physical changes to any structure, system or component, and it does not affect any design basis accident evaluation.

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

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

Response: No.

The proposed change to the list of methodologies used at PVNGS to determine the various COLR limits is an administrative change which updates the list in the TS to include all of the NRC reviewed and approved COLR methodologies for PVNGS. This change does not create any new failure modes or affect the interaction between any structure, system or component.

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

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

Response: No.

The proposed change to the list of methodologies used at PVNGS to determine the various COLR limits is an administrative change which updates the list in the TS to include all of the NRC reviewed and approved COLR methodologies for PVNGS. This change does not modify any margin of safety.

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

The NRC staff has reviewed the licensee's analysis and, based on 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.

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

Date of amendments request: August 29, 2008.

Description of amendments request: The amendment would revise Calvert Cliffs Nuclear Power Plant (CCNPP) Operating License Nos. DPR-53 and DPR-69 and Technical Specifications (TSs) by increasing the licensed core power of CCNPP, Unit Nos. 1 and 2 by 1.38 percent to 2737 MWt. The power uprate amendment request is based on the use of the Caldon Leading Edge Flow Measurement (LEFM) CheckPlus system for more accurate determination of main feedwater flow and the associated determination of reactor power through the performance of the power calorimetric calculation currently required by CCNPP TSs.

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. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

In support of this measurement uncertainty recapture (MUR) power uprate, a comprehensive evaluation was performed for Nuclear Steam Supply System (NSSS), balance of plant systems and components, and analyses that could be affected by this change. A power calorimetric uncertainty calculation was performed, and the impact of increasing plant power by 1.38 percent on the plant's design and licensing basis was evaluated. The result of these evaluations is that structures, systems, and components required to mitigate transients will continue to be capable of performing their design function at an uprated core power of 2737 MWt. In addition, an evaluation of the accident analyses demonstrates that applicable analysis acceptance criteria continue to be met. No accident initiators are affected by this uprate and no challenges to any plant safety barriers are created by this change. Therefore, operation of the facility in accordance with the proposed change will not involve a significant increase in the probability of an accident previously evaluated.

The proposed change does not affect the radiological release paths, the frequency of release, or the source-term for release for any accidents previously evaluated in the Updated Final Safety Analysis Report. Structures, systems, and components required to mitigate transients remain capable of performing their design functions, and thus were found acceptable. The reduced uncertainty in the feedwater flow input to the power calorimetric measurement ensures that Start Printed Page 65689applicable accident analyses acceptance criteria continue to be met in support of operation at a core power of 2737 MWt. Analyses performed to assess the effects of mass and energy remain valid. The source-terms used to assess radiological consequences have been reviewed and determined to bound operation at the uprated condition. Therefore, operation of the facility in accordance with the proposed change will not involve a significant increase in the consequences of an accident previously evaluated.

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

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

No new accident scenarios, failure mechanisms, or single-failures are introduced as a result of the proposed changes. The installation of the Caldon LEFM CheckPlus feedwater flow instrumentation system has been analyzed, and failures of this system will have no adverse effect on any safety-related system or any structures, systems, and components required for transient mitigation. All structures, systems and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety-related system.

This change does not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than was previously evaluated. Operating at a core power level of 2737 MWt does not create any new accident initiators or precursors. The reduced uncertainty in the feedwater flow input to the power calorimetric measurement ensures that applicable accident analyses acceptance criteria continue to be met to support operation at a core power of 2737 MWt. Credible malfunctions continue to be bounded by the current accident analysis of record or evaluations that demonstrate that applicable acceptance criteria continue to be met.

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

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The margins of safety associated with the MUR power uprate are those pertaining to core power. This includes those associated with the fuel cladding, Reactor Coolant System pressure boundary, and containment barriers. A comprehensive engineering review was performed to evaluate the 1.38 percent increase in the licensed core power from 2700 MWt to 2737 MWt. The 1.38 percent increase required that revised NSSS design thermal and hydraulic parameters be established, which then served as the basis for all of the NSSS analyses and evaluations. This engineering review concluded that no design modifications are required to accommodate the revised NSSS design conditions. The NSSS components were evaluated and it was concluded that the NSSS components have sufficient margin to accommodate the 1.38 percent power uprate. The NSSS accident analyses were evaluated for the 1.38 percent power uprate. In all cases, the evaluations demonstrate that the applicable analyses acceptance criteria continue to be met. As a result, the margins of safety continue to be bounded by the current analyses of record for this change.

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

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

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

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: September 11, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specifications, extending the 15-year interval between containment Type A tests specified by Specification 4.4.a, “Integrated Leak Rate Test,” by 6 months. The current Type A test interval expires at the end of April 2009. The proposed amendment would extend this interval, on a one-time basis, to October 2009 to coincide with completion of the next scheduled refueling outage.

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 probability or consequences of accidents previously evaluated in the Updated Safety Analysis Report are unaffected by this proposed change. There is no change to any equipment response or accident mitigation scenario, and this change results in no additional challenges to fission product barrier integrity. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. As a result, the probabilities of previously evaluated accidents are unaffected. The proposed extension to the Type A test interval does not involve a significant increase in consequences because, as discussed in NUREG-1493, Performance Based Containment Leak Rate Test Program, Type B and C tests identify the vast majority (approximately 97 percent) of all potential leakage paths. Further, Type A tests identify only a few potential leakage paths that cannot be identified through Type B and C testing, and leaks found by Type A testing have been only marginally greater than existing requirements. The frequency and methods of performance of Type B and Type C testing are unaffected by this proposed change. In addition, periodic inspections of containment required by the ASME [American Society of Mechanical Engineers] code and the maintenance rule, which are capable of detecting any significant degradation, are unaffected by the proposed change.

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

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

Response: No.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. The proposed change does not install or remove any plant equipment. The proposed change does not alter the design, physical configuration, or mode of operation of any plant structure, system, or component. No physical changes are being made to the plant, so no new accident causal mechanisms are being introduced. The proposed change only changes the frequency of performing the next Type A test; the Type A test implementation and acceptance criteria are unchanged. Type B and Type C testing frequency and method of performance are not affected by this proposed change.

Therefore, the proposed amendment 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 margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. The ability of operable structures, systems, and components to perform their designated safety function is unaffected by this proposed change. NUREG-Start Printed Page 656901493 concluded that reducing the frequency of Type A tests to one-in-20 years resulted in an imperceptible increase in risk. Type B and Type C testing frequency and method of performance are unaffected by this proposed change. Also, [other] inspections of containment required by the ASME code and the maintenance rule [will] provide reasonable assurance that containment will not degrade in a manner that is only detectable by Type A testing. In addition, the inherent risk of an additional plant shutdown would be eliminated by the proposed amendment, further ensuring no significant reduction in safety margin.

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

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

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

NRC Branch Chief: Lois M. James.

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: July 28, 2008.

Description of amendment request: The proposed amendment would: (1) Delete Technical Specification (TS) surveillance requirement (SR) 3.1.3.2 and revise SR 3.1.3.3, (2) remove reference to SR 3.1.3.2 from Required Action A.2 of TS 3.1.3, “Control Rod OPERABILITY,” (3) clarify the requirement to fully insert all insertable rods for the limiting condition for operation (LCO) in TS 3.3.1.2, required Action E.2, “Source Range Monitoring Instrumentation,” and (4) revise Example 1.4-3 in Section 1.4, “Frequency,” to clarify the applicability of the 1.25 surveillance test interval extension.

The NRC staff issued a notice of opportunity to comment in the Federal Register on August 16, 2007 (72 FR 46103), on possible amendments to revise the plant-specific TSs, modify TS control rod SR testing frequency, clarify TS control insertion requirements, and clarify SR frequency discussions, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on November 13, 2007 (72 FR 63935). The licensee affirmed the applicability of the model NSHC determination in its application dated July 28, 2008.

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

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

The proposed change generically implements TSTF-475, Revision 1, “Control Rod Notch Testing Frequency and SRM [Source Range Monitor] Insert Control Rod Action.” TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) STS. The changes: (1) revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, “Control Rod OPERABILITY,” (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, “Source Range Monitoring Instrumentation” (NUREG-1434 only), and (3) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The consequences of an accident after adopting TSTF-475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

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

TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 frequency in TS 3.1.3, “Control Rod OPERABILITY,” (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, “Source Range Monitoring Instrumentation,” and (3) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The GE [General Electric] Nuclear Energy Report, “CRD [Control Rod Drive] Notching Surveillance Testing for Limerick Generating Station,” dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF-475, Revision 1 are acceptable and do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: July 21, 2008.

Description of amendment request: The proposed amendment would support a proposed change to the in-service inspection program that is based on topical report WCAP-16168-NP-A, Revision 2, “Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.” In the referenced safety evaluation of the topical report, the NRC required licensees to amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) be submitted for NRC staff review and approval within one year of completing the required reactor vessel weld inspection. Entergy Nuclear Operations, Inc., proposes to add a new license condition to provide this information.

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

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

Response: No.

The proposed amendment changes the renewed facility operating license by adding a license condition to require that the Start Printed Page 65691information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) will be submitted for NRC staff review and approval within one year of completing the required reactor vessel weld inspection. The proposed amendment does not involve operation of the required structures, systems or components (SSCs) in a manner or configuration different from those previously recognized or evaluated.

The proposed changes are administrative and have no impact on plant operation or equipment.

Therefore, operation of the facility in accordance with the proposed 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 license amendment does not involve a physical alteration of any SSC or change the way any SSC is operated. The proposed license amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated.

The proposed changes are administrative and have no impact on plant operation or equipment.

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

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

Response: No.

The proposed changes are administrative and have no impact on plant operation or equipment or on any margin of safety.

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

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

Attorney for licensee: Mr. William 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 No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of amendment request: August 28, 2008.

Description of amendment request: The proposed amendment would change Technical Specifications (TS) Administrative Controls section 5 to incorporate NRC-approved Technical Specification Task Force (TSTF) Improved Technical Specification (ITS) TSTF-363, “Revise Topical Report references in ITS 5.6.5, [Core Operating Limits Report] COLR,” revision 0. ENO also proposes to make an administrative change to the plant staff qualifications section.

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

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

Response: No.

The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative or provide clarification only.

The proposed changes do not have any impact on the integrity of any plant system, structure, or component (SSC) that initiates an analyzed event. The proposed changes will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Thus, the probability of any accident previously evaluated is not significantly increased.

The proposed changes do not affect the ability to mitigate previously evaluated accidents, and do not affect radiological assumptions used in the evaluations. The proposed changes do not change or alter the design criteria for the systems or components used to mitigate the consequences of any design-basis accident. The proposed amendment does not involve operation of the required SSCs in a manner or configuration different from those previously recognized or evaluated. Thus, the radiological consequences of any accident previously evaluated are not increased.

Therefore, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not involve a physical alteration of any SSC or a change in the way any SSC is operated. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested.

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

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

Response: No.

The amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not affect any margin of safety. The proposed amendment does not involve any physical changes to the plant or manner in which the plant is operated.

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

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

Attorney for licensee: Mr. William 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 No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: September 30, 2008.

Description of amendment request: The proposed amendment would revise the Facility Operating License and Technical Specification Section 4.0 by changing the names of the licensees to Enexus Nuclear Pilgrim LLC and EquaGen Nuclear LLC.

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:

The proposed amendment would only change the names of the licensees and reflect the referenced NRC Order requirements. Principal management and operational staffing for the restructured organization remain largely unchanged. The proposed changes do not: (a) Involve a significant increase in the probability or consequences of an accident previously evaluated; (b) create the possibility of a new or different kind of accident from any accident previously evaluated; or (c) involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Start Printed Page 65692Hamilton 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 4, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) Section 5.1, “Site,” to remove the restriction on the sale and lease of site property and replace the restriction with a requirement to retain complete authority to determine and maintain sufficient control of all activities, including the authority to exclude or remove personnel and property, within the minimum exclusion area as described in 10 CFR 100.3.

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 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 that affects the probability of an accident or that supports mitigation or consequences of an accident previously evaluated. The proposed change establishes requirements for sale or lease of property within the exclusion area. Additionally, ENO [Entergy Nuclear Operations, Inc.] will retain authority to determine all activities within the exclusion area and to remove personnel and property from the area as necessary to ensure the regulatory exposure limits are met.

The proposed change does not affect reactor operations or accident analysis and there is no change to the radiological consequences of a previously analyzed accident. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. The proposed change establishes requirements for sale or lease of property within the exclusion area. Any additional activities performed within the exclusion area will be reviewed by ENO and verified to not represent a new hazard or that they have been accommodated in the plant licensing and design basis. As such, no new or different types of equipment will be installed or operated without additional review and approval by ENO. Operation of existing installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

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

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

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: September 22, 2008.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) to remove the requirement to perform quarterly closure time testing of the Main Steam Isolation Valves (MSIVs) by deleting TS Surveillance Requirement 4.7.D.1.c. Operability testing of the MSIVs will continue to be required by the Vermont Yankee Inservice Test Program.

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. 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 deletes the specific surveillance requirement to exercise the MSIVs once per quarter from the TS. Following implementation of the proposed change, the VY TS still will require operability testing of the MSIVs by reference to the VY IST program. The quarterly exercise involves a timed full stroke closure of each individual MSIV and subsequent reopening to the full open position. Details of MSIV testing requirements will continue to be contained in the VY IST program. The MSIV closure time setpoint values related to the safety functions of the MSIVs will continue to be contained in the VY UFSAR [Updated Final Safety Analysis Report] and the VY TRM [Technical Requirements Manual]. Changes to the VY UFSAR and TRM are evaluated per the requirements of 10 CFR 50.59. These controls are adequate to ensure the required inservice testing is performed to verify the MSIVs are operable and capable of performing their safety functions. The proposed amendment introduces no new equipment or changes to how equipment is operated. Therefore, the proposed amendment will 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.

The proposed amendment deletes the specific surveillance requirement to exercise the MSIVs once per quarter from the TS. Following implementation of the proposed change, the VY TS still will require operability testing of the MSIVs by reference to the VY IST program. The quarterly exercise involves a timed full stroke closure of each individual MSIV and subsequent reopening to the full open position. 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.

The proposed amendment deletes the specific surveillance requirement to exercise the MSIVs once per quarter from the TS. Following implementation of the proposed change, the VY TS still will require operability testing of the MSIVs by reference to the VY IST program. The quarterly exercise involves a timed full stroke closure of each individual MSIV and subsequent reopening to the full open position. The proposed amendment does not change the design or function of any component or system. The proposed amendment does not involve any safety limits or safety settings. The ability of the MSIVs to perform their safety function will continue to be tested in Start Printed Page 65693accordance with the IST Program, through TS SR 4.7.D.1.b.

Since the proposed controls are adequate to ensure the required inservice testing is performed, there will still be high assurance that the components are operable and capable of performing their respective safety functions, and that the systems will respond as designed to mitigate the subject events. Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in [a] margin to safety.

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

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

NRC Branch Chief: Mark G. Kowal.

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 30, 2008.

Description of amendment request: The proposed amendment would revise the Facility Operating License and Technical Specification Section 5.0 by changing the names of the licensees to EquaGen Nuclear LLC and Enexus Nuclear Vermont Yankee LLC, respectively.

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:

The proposed amendment would only change the names of the licensees and reflect the referenced NRC Order requirements; principal management and operational staffing for the restructured organization remain largely unchanged. The proposed changes do not: (a) Involve a significant increase in the probability or consequences of an accident previously evaluated; (b) create the possibility of a new or different kind of accident from any accident previously evaluated; or (c) involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Mark G. Kowal.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas

Date of amendment requests: July 21, 2008.

Description of amendment request: The proposed change allows a delay time for entering a supported system Technical Specification (TS) when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.8 is added to the TS to provide this allowance and define the requirements and limitations for its use.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff issued a notice of opportunity for comment in the Federal Register on November 24, 2004 (69 FR 68412), on possible amendments concerning TSTF-372, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252).

Basis for proposed no significant hazards consideration determination: Entergy Operations, Inc. (Entergy) has reviewed the proposed NSHC determination published in the Federal Register as part of the CLIIP. Entergy has concluded that the proposed NSHC determination presented in the Federal Register notice is applicable to Arkansas Nuclear One, Unit 2 and is presented below:

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in Regulatory Guide 1.177. A bounding risk assessment was performed to justify the proposed TS changes. The proposed LCO 3.0.8 defines limitations on the use of the provision and includes a requirement for the licensee to assess and manage the risk associated with operation with an inoperable snubber. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley. Start Printed Page 65694

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 on October 2, 2008.

Description of amendment request: Entergy Operations Inc. (the licensee) proposes to modify the technical specifications (TS) 3.6.6, “Spray Additive System.” Specifically, this amendment proposes to revise the Sodium Hydroxide (NaOH) tank concentration stated in TS 3.6.6.3 from between 5.0 percent and 16.5 percent to between 6.0 percent and 8.5 percent.

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.

There are no changes to the design or operation of the plant that could affect system, component, or accident functions as a result of changing the sodium hydroxide (NaOH) tank solution concentration limits. In addition, the dose reduction provided by maintaining the sump pH above 7.0 is retained, and therefore, dose consequences resulting from iodine dissolution remain unchanged. The proposed change simply imposes more restrictive operating conditions than are within the current TS limits. 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.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of the proposed change. Structures, systems, and components previously required for mitigation of an accident remain capable of fulfilling their intended design function with this change to the TS. The proposed change has no new adverse effects on safety-related systems or components and does not challenge the performance or integrity of safety-related systems. The proposed change simply imposes more restrictive operating conditions that are within the current TS limits. 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 imposes more restrictive operating conditions that are within the current TS limits. Revising the NaOH tank solution concentration limits reduces the amount of chemical precipitates formed under post-loss-of-coolant accident conditions. The margin of safety related to ensuring that the sump pH remains above 7.0 is not reduced. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas

Date of amendment request: August 21, 2008.

Description of amendment request: Entergy Operations Inc. (the licensee) proposes a one-time amendment for next containment integrated leakage rate test (ILRT) or Type A test at the Arkansas Nuclear One, Unit No. 2 (ANO-2). The ILRT is required by Technical Specification (TS) 6.5.16, “Containment Leakage Rate Testing Program,” to be performed every ten-years. The amendment would permit the existing ILRT frequency to be extended from 120 months (10 years) to approximately 135 months.

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

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

Response: No.

The proposed exemption involves a one-time extension to the current interval for Type A containment testing. The current test interval of 120 months (10 years) would be extended on a one-time basis to no longer than approximately 135 months from the last Type A test. The proposed extension does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the reactor containment itself and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated nor does it create the possibility of a new or different kind of accident.

This proposed extension is for the Type A containment leak rate tests only. The Type B and C containment leak rate tests will continue to be performed at the frequency currently required by the ANO-2 TS. As documented in NUREG 1493, Type B and C tests have identified a very large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is very small. ANO-2's Type A test history supports this conclusion.

The integrity of the reactor containment is subject to two types of failure mechanisms which can be categorized as (1) activity based and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment itself combined with the containment inspections performed in accordance with ASME, Section XI, the Maintenance Rule, and Licensing commitments serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extension does not involve a significant increase in the consequences of an accident previously evaluated.

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

Response: No.

The proposed revision to the TS involves a one-time extension to the current interval for Type A containment testing. The reactor containment and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve the prevention or identification of any precursors of an accident. The proposed TS change does not involve a physical change to the plant or the manner in which the plant is operated or controlled. 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 change involve a significant reduction in a margin of safety?

Response: No.

The proposed change to the TS involves a one-time extension to the current interval for Type A containment testing. The proposed TS change does not involve a physical Start Printed Page 65695change to the plant or a change in the manner in which the plant is operated or controlled. The specific requirements and conditions of the Primary Containment leak Rate Testing Program, as defined in the TS, exist to ensure that the degree of reactor containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests. The proposed surveillance interval extension is bounded by the 15 month extension currently authorized within NEI 94-01, Revision 0. Type B and C containment leak rate tests will continue to be performed at the frequency currently required by TS. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME, Section XI and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety that is in plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards will continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A test interval. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: July 21, 2008.

Description of amendment requests: The proposed amendments would modify the Technical Specification (TS) by adding Limiting Condition for Operation (LCO) 3.0.8 on the inoperability of snubbers using the Consolidated Line Item Improvement Process (CLIIP). The proposed amendments would also make conforming changes to TS LCO 3.0.1. This request is consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Traveler No. 372, Revision 4, “Addition of LCO 3.0.8, Inoperability of Snubbers.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on November 24, 2004 (69 FR 68412), on possible amendments concerning TSTF-372, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252). Basis for proposed no significant hazards consideration determination: Entergy Operations, Inc. (Entergy) has reviewed the proposed NSHC determination published in the Federal Register as part of the CLIIP. Entergy has affirmed the applicability of the following NSHC for Arkansas Nuclear One, Unit 1 in its application and as published in the Federal Register.

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

The proposed changes allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering a supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, these changes do not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed changes allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in NRC Regulatory Guide 1.177. A bounding risk assessment was performed to justify the proposed TS changes. The application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and management of plant risk [which is required by the proposed TS 3.0.8]. The net change to the margin of safety is insignificant. Therefore, these changes do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael Markley.

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

Date of amendment request: September 17, 2008.

Description of amendment request: The proposed change will revise the Operating License to modify Note 2 of Waterford 3 Technical Specification Table 4.3-1. The licensee stated that the proposed change will result in the addition of conservatism to Core Protection Calculator (CPC) power indications when calibrations are required in certain conditions.

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? Start Printed Page 65696

Response: No.

The proposed change will redefine the tolerance band allowed for the Reactor Protection System (RPS) linear power, Core Protection Calculator (CPC) ΔT [Delta Temperature] power, and CPC neutron flux power signals, and clarify the intent of the calibration requirements for CPC power indications when at less than 15% [percent] power, and specify that adjustment limits are percentages of RATED THERMAL POWER instead of percentages of current power. Redefining the tolerance band is in conformance with the safety analysis. The consequences of an accident will be in conformance with the safety analysis.

Clarifying the intent of there being no calibration requirements for CPC power indications when at less than 15% power is essentially editorial. At this low power level, CPC calculations compensate for any potential de-calibration. Specifying that adjustment limits are percentages of RATED THERMAL POWER instead of percentages of current power is essentially editorial. This change is made to avoid confusion in interpreting the requirements. This amendment request does not change the design, analysis or operation of any plant systems or components.

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 to Technical Specification power calibration tolerance limits is in conformance with the safety analysis. This amendment request does not change the design, analysis or operation of any plant systems or components. CPC's cannot cause an accident, and this change will not create the possibility of a new or different type of accident.

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

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

Response: No.

The proposed change to Technical Specification power calibration tolerance limits is in conformance with the safety analysis. This proposed change maintains the margin of safety for design basis events. Therefore, this change does not involve a significant reduction in the margin of safety.

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

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

Date of amendment request: September 17, 2008.

Description of amendment request: Entergy has proposed to add a license condition on the extension of the reactor vessel inservice inspection interval. This proposed license condition is the result of a condition in the Nuclear Regulatory Commission (NRC) safety evaluation, issued by letter dated May 8, 2008, on topical report WCAP-16168-NP-A, Revision 2, “Risk-Informed Extension of the Reactor Vessel In-Service Inspection [ISI] Interval,” dated June 8, 2008. The ISI interval extension part of a relief request is being separately evaluated by NRC and independent of this amendment request.

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

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

Response: No.

The proposed change will revise the license to require the submission of information and analyses to the NRC following completion of each ASME Code, Section XI, Category B-A and B-D reactor vessel weld inspection. The extension of the ISI interval from 10 to 20 years is being evaluated as part of the relief request independent from this license change. Submission of the information and analyses are administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

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 a physical alteration of any SSC or change the way any SSC is operated. The proposed addition of the license condition has no impact on any plant configurations or on system performance that is relied upon to mitigate the consequences of an accident. The license condition is administrative in nature and does not result in a change to the physical plant or to the modes of operation defined in the facility license. Entergy has demonstrated that the Limitations and Conditions associated with the NRC SE will be met.

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 addition of the license condition is administrative in nature and has no impact on plant operation or equipment or on any margin of safety. The license condition to submit information and analyses is an administrative tool to assure the NRC has the ability to independently review information developed by the Licensee.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: June 27, 2007, as supplemented on September 4, 2008.

Description of amendment request: The proposed amendment request dated June 27, 2008, would revise Technical Specifications (TS) Surveillance Requirements 3.8.1.2, 8, 12, 13, 16, and 19, changing the steady state frequency and voltage of all diesel generators (DGs) from the currently allowed frequency range of 59.4-61.2 Hz to 59.4-60.5 Hz (i.e., a decrease of the upper limit, resulting in narrowing of the current range). The licensee stated that the current frequency range is nonconservative and could result in undesirable effects such as centrifugal charging pump motor brake horsepower exceeding its nameplate maximum horsepower, and overloading the DGs. The Commission previously noticed this proposed amendment request on August 14, 2007 (72 FR 45458).

The scope of the June 27, 2008, proposed amendment request was expanded as described in a supplemental letter dated September 4, 2008. The expanded scope would revise (1) TS Surveillance Requirements 3.8.1.8, 13, 16, and 22, changing the minimum voltage and frequency that Start Printed Page 65697the DGs must achieve within 10 seconds after starting from ≥ 3740 Volts (V) to ≥ 3910 V and ≥ 58.8 Hz to ≥ 59.4 Hz, respectively, and (2) TS Surveillance Requirement 3.8.1.10, changing the maximum DG frequency allowed to occur within 2 seconds following a load rejection of the single largest post-accident load from ≤ 61.2 Hz to ≤ 60.5 Hz. The changes proposed by the supplement indirectly affect TS 3.8.2.1 which requires that TS Surveillance Requirements 3.8.1.8, 10, and 16 be met.

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

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

Response: No.

The more restrictive transient voltage and frequency limits ensures that the equipment powered from the DGs will function as designed to mitigate an accident as described in the Update Final Safety Analysis Report (UFSAR). The DGs and the equipment they power are part of the systems required to mitigate accidents; no accident analyzed in the UFSAR is initiated by mitigation equipment. Therefore, the proposed change to the allowed frequency range of the DGs will not have any impact on the probability of an accident previously evaluated. Furthermore, other than requiring more restrictive transient voltage and frequency limits of DGs, there is no other design or operational change. Therefore, the proposed change does not increase the probability of malfunction of the DGs or the equipment they power.

The more restrictive DG transient voltage and frequency limits will ensure that the equipment powered by the DGs will perform as originally designed and analyzed to mitigate the consequences of any accident described in the UFSAR. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated in the UFSAR.

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

Response: No.

There is no design change associated with the proposed amendment. Making an existing DG requirement more restrictive alone will not alter plant configuration because no new or different type of equipment will be installed, and because no methods governing plant operation will be changed. The proposed change to transient voltage and frequency limits will not have any effect on the assumptions of accident scenarios previously made in the UFSAR. 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.

Despite the proposed change to the DG transient voltage and frequency limits, the DGs and equipment powered by the DGs will continue to perform as originally designed, and originally analyzed in the UFSAR. There is no associated change to the methods and assumptions used to analyze DG performance. The proposed change will maintain the required function of the DGs and the equipment powered by the DGs to ensure that operation of structures, systems, or components is as currently set forth in the UFSAR. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: James M. Petro, Jr., One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: Lois M. James.

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

Date of amendment request: July 2, 2008.

Description of amendment request: The proposed amendment would correct several typographical errors and make administrative clarifications to the Technical Specifications (TS).

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

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

Response: No.

The proposed changes correct typographical and administrative errors, or make clarifications that more accurately reflect TS requirements. Administrative and editorial changes such as these are not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident with the incorporation of these administrative and editorial changes are no different than the consequences of the same accident without these changes. As a result, the consequences of an accident previously evaluated are not affected by these changes.

The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated.

Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes are consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed changes do not involve an increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed changes do not alter any assumptions made in the safety analysis. Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated.

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

Response: No.

The proposed changes consist of administrative and editorial changes to correct typographical or administrative errors and oversights or clarify the meaning of the TS. The 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 these changes. The proposed changes will not result in plant operation in a configuration outside of the design basis. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee

Date of amendment request: September 18, 2008.

Description of amendment request: The proposed amendment would revise technical specification (TS) 3.8.7, Start Printed Page 65698“Inverters—Operating.” The current TS requires one inverter for each of the four channels. The proposed amendment would revise TS 3.8.7 to require two inverters for each of the four channels.

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

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

Response: No.

The proposed revisions to WBN's Vital AC [alternating current] Power System do not alter the safety functions of the Vital Inverters or the Unit 1 and Unit 2 120V [volt] AC Vital Instrument Power Boards. The initial conditions for the DBAs [design-basis accidents] defined in the WBN UFSAR [Updated Final Safety Analysis Report] assume the ESF [engineered safety feature] systems are operable. The vital inverters are designed to provide the required capacity, capability, redundancy, and reliability to ensure the availability of necessary power to vital instrumentation so that the fuel, reactor coolant system, and containment design limits are not exceeded. Separating the Unit 2 loads from the Unit 1 inverters does not alter the accident analyses. Design calculations document that the inverters have adequate capacity to support the loads required for Unit 1 operation and no changes are proposed that will impact the separation of the Vital AC Power System.

The inverters and the associated 120V AC Vital Instrument Power Boards are utilized to support instrumentation that monitor critical plant parameters to aid in the detection of accidents and to support the mitigation of accidents, but are not considered to be an initiator of design basis accidents. Based on this and the preceding information, the proposed amendment 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.

When implemented, the proposed TS amendment will allow the Unit 2 Vital Instrument Power Boards to receive their UPS [uninterruptible power supply] power from new Unit 2 inverters. Calculations have verified that the loads will not affect the ability of the inverters to perform their intended safety functions. In addition, the inverters and the 120V AC Vital Instrument Power Boards are not considered to be an initiator of a DBA. These components provide power to instrumentation that supports the identification and mitigation of accidents as well as system control functions during normal plant operations. The functions of the inverters are not altered by the proposed TS change and will not create the possibility of a new or different accident. Further, the separation of the Unit 2 loads from the Unit 1 inverters is the principal change to the inverter system, and this change is bounded by previously evaluated accident analyses. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The plant setpoints and limits that are utilized to ensure safe operation and detect accident conditions are not impacted by the proposed TS amendment. The inverters and the 120V Vital Instrument Power Boards will continue to provide reliable power to safety-related instrumentation for the identification and mitigation of accidents and to support plant operation. Therefore, the margin of safety is not reduced.

Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of “no significant hazards consideration” is justified.

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

NRC Branch Chief: L. Raghavan.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: September 18, 2008.

Description of amendment request: The proposed amendment would revise technical specification (TS) Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” to modify Mode 1 and 2 Applicability for Function 6.e, and would revise limiting condition for operation (LCO) 3.3.2, Condition J.

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 design basis events which impose [auxiliary feedwater] AFW safety function requirements are loss of normal main feedwater, main feed line or main steam line break, loss of offsite power (LOOP), and small break loss of coolant accident. These design bases event evaluations assume actuation of the AFW due to LOOP signal, low-low steam generator level or a safety injection signal. The anticipatory AFW auto-start signals from the turbine driven main feedwater (TDMFW) pumps are not credited in any design basis accidents and are, therefore, not part of the primary success path for postulated accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion 3. Modifying Mode 1 and 2 Applicability for this function will not impact any previously evaluated design basis accidents. 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.

This TS change allows for an operational allowance during Mode 1 and 2 for placing TDMFW pumps in service or securing TDMFW pumps. This change involves an anticipatory AFW auto-start function that is not credited in the accident analysis. Since this change only affects the conditions at which this auto-start function needs to be operable and does not affect the function that actuates AFW due to loss of offsite power, low-low steam generator level or a safety injection signal, it will not be an initiator to 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 TS change involves the automatic start of the AFW pumps due to trip of both TDMFW pumps, which is not an assumed start signal for design basis events. This change does not modify any values or limits involved in a safety related function or accident analysis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of “no significant hazards consideration” is justified.

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

NRC Branch Chief: L. Raghavan.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee

Date of amendment request: September 19, 2008. Start Printed Page 65699

Description of amendment request: The proposed amendment would modify the WBN Final Safety Analysis Report (FSAR) 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 (OBE) within the five week period after ice basket replenishment has been completed.

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 analyzed accidents of consideration in regard to changes potentially affecting the ice condenser are a loss of coolant accident and a steam or feedwater line break inside Containment. The ice condenser is an accident mitigator and is not postulated as being the initiator of a LOCA [loss-of-coolant accident] or HELB [high energy line break]. The ice condenser is structurally designed to withstand a Safe Shutdown Earthquake plus a Design Basis Accident and does not interconnect or interact with any systems that interconnect or interact with the Reactor Coolant, Main Steam, or Feedwater systems. Because the proposed changes do not result in, or require any physical change to the ice condenser that could introduce an interaction with the Reactor Coolant, Main Steam, or Feedwater systems, there can be no change in the probability of an accident previously evaluated.

Under the proposed change, there is some finite probability that, within 24 hours following a seismic disturbance, a LOCA or HELB in Containment could occur within five weeks of the completion of ice basket replenishment. However, several factors provide defense-in-depth and tend to mitigate the potential consequences of the proposed change.

Design basis accidents are not assumed to occur simultaneously with a seismic event. Therefore, the coincident occurrence of a LOCA or HELB with a seismic event is strictly a function of the combined probability of the occurrence of independent events, which in this case is very low. Based on the Probabilistic Risk Assessment model and seismic hazard analysis, the combined probability of occurrence of a seismic disturbance greater than or equal to an OBE during the 5 week period following ice replenishment coincident with or subsequently followed by a LOCA or HELB during the time required to perform the proposed inspection (24 hours) and if required by Technical Specifications, complete Unit shutdown (37 hours), is less than 3.7E-09 for WBN. This probability is well below the threshold that is typically considered credible.

Even if ice were to fall from ice baskets during a seismic event occurring coincident with or subsequently followed by an accident, the ice condenser would be expected to perform its intended safety function. Due to the ice servicing methodology utilized by WBN, the relatively small amount of ice that may potentially fallout from the ice baskets to the floor behind the lower inlet doors during the seismic event is such that complete blockage of flow into the ice condenser is not credible during a LOCA or HELB.

Based on the above, the proposed changes do not involve a significant increase in the probability or consequences. The ice condenser is expected to perform its intended safety function under all circumstances following a LOCA or HELB in Containment.

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 provides an alternate methodology to confirm the ice condenser lower inlet doors are capable of opening if a seismic event occurs within five weeks of ice basket replenishment. As previously discussed, the ice condenser is not postulated as an initiator of any design basis accident. The proposed change does not impact any plant system, structure, or component that is an accident initiator. The proposed change does not involve any hardware changes to the ice condenser or other changes that could create new accident mechanisms. Therefore, there can be no new or different accidents created from those previously identified and 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.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the Reactor Coolant system, and the Containment system. The performance of the fuel cladding and the Reactor Coolant system will not be impacted by the proposed change.

The requirement to inspect the ice condensers within 24 hours of experiencing seismic activity greater than or equal to an OBE during the five (5) week period following the completion of ice basket replenishment will confirm whether the ice condenser lower inlet doors are capable of opening. This inspection will either confirm that the ice condenser doors remain fully capable of performing their intended safety function under credible circumstances or that a Unit shutdown is required.

The ice condenser has reasonable assurance of performing its intended function during the highly unlikely scenario in which a postulated accident (LOCA or HELB) occurs coincident with or subsequently following a seismic event.

The proposed change affects the assumed timing of a postulated seismic and design basis accident applied to the ice condenser and provides an alternate methodology in confirming the ice condenser lower inlet doors are capable of opening. As previously discussed, the combined probability of occurrence of a LOCA or HELB and a seismic disturbance greater than or equal to an OBE during the “period of potential exposure” is less than 3.7E-09 for WBN. This probability is well below the threshold that is considered credible.

Therefore, the proposed change does not involve a significant reduction in the margin of safety. The WBN ice condenser will perform its intended safety function under credible circumstances.

The changes proposed in this LAR [license amendment request] do not make any physical alteration to the ice condensers, nor does it affect the required functional capability of the ice condenser in any way. The intent of the proposed change to the FSAR is to eliminate an overly restrictive waiting period prior to Unit ascent to power operations following the completion of ice basket replenishment. The required inspection of the ice condenser following a seismic event greater than or equal to an OBE will confirm whether the ice condenser lower inlet doors will continue to fully perform their safety function as assumed in the WBN safety analyses.

Thus, it can be concluded that the proposed change does not involve a significant reduction in the margin of safety.

Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of “no significant hazards consideration” is justified.

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

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: April 2, 2008.

Description of amendment request: The proposed change revises Technical Specification (TS) Section 5.0, “Design Features,” to delete certain design details and descriptions included in TS 5.0 that are already contained in the Updated Final Safety Analysis Report Start Printed Page 65700(UFSAR), or are redundant to existing TS requirements, and are not required to be included in the TSs pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36(d)(4). The proposed change also revises the format of, and incorporates design descriptions into, TS 5.0 consistent with Nuclear Regulatory Commission (NRC) policy and NUREG-1431, “Standard Technical Specifications Westinghouse Plants, Revision 3.0,” to the extent practical. An editorial change is also proposed to address a minor TS discrepancy introduced by a previous license amendment. More specifically, the proposed change includes removing Section 5.2, “Containment,” from the TSs in its entirety. This section contains the minimum spray flows for the Containment Spray (CS) and Recirculation Spray (RS) Subsystems. The proposed change also removes the statement describing how draining of the spent fuel pool is prevented, and includes a statement in the TS that would limit draining the spent fuel pool below the elevation of 41 feet, 2 inches mean sea level. Additionally, the licensee proposes to incorporate the spent fuel pool storage capacity of 1044 assemblies into the TSs. This limit was previously established by Amendment Nos. 37 and 36 to Surry Power Station, Unit Nos. 1 and 2, respectively.

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

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

Response: No.

The proposed changes to Section 5.0, “Design Features,” removes certain details from the TSs that are not required to be maintained in the TSs by 10 CFR 50.36(d)(4), or are adequately controlled by other existing TSs, incorporates previously approved TS limits that meet the 10 CFR 50.36(d)(4) inclusion criteria, and revises the TSs for consistency with NUREG-1431. An additional change addresses a minor editorial discrepancy introduced by a previous amendment. The minimum spray flow values for the CS and RS Subsystems are removed, but operability and performance of both subsystems are adequately controlled by existing TSs ensuring they will continue to perform their design functions. The proposed changes remove the statement describing how draining of the spent fuel pool is prevented (does not meet the criteria of 10 CFR 50.36(d)(4)for inclusion in the TSs) and includes a statement in the TS that would limit draining the spent fuel pool below the elevation of 41 feet, 2 inches mean sea level (as analyzed in the UFSAR and consistent with the content and format of NUREG-1431). The proposed change incorporates the spent fuel pool storage capacity of 1044 assemblies into the TSs. This limit was evaluated in previously approved Amendment Nos. 37 and 36 to Surry Power Station, Unit Nos. 1 and 2, respectively. The proposed changes are considered administrative in nature and do not affect initiators of previously analyzed events or assumed mitigation of accident or transient events. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

There is no physical alteration of the plant (no new or different type of equipment will be installed) associated with the proposed amendment. The proposed changes will not have any effect on the assumptions of accident scenarios previously made in the UFSAR. The proposed changes do not alter or prevent the ability of structures, systems, and components to perform their intended function to mitigate the consequences of an initiating event. The proposed changes are considered administrative in nature. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The spent fuel pool and the CS and RS Subsystems will continue to perform as designed and analyzed in the UFSAR. There is no associated change to the methods and assumptions used to analyze their performance. Their required function will be maintained as currently set forth in the UFSAR and existing TSs. The proposed changes do not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The dose analysis is also not affected. The proposed changes are considered administrative in nature and do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Melanie C. Wong.

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.

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

Date of amendment request: July 30, 2007, as supplemented on April 7 and September 8, 2008.

Description of amendment request: This amendment would modify Technical Specification 3.7.3, “Control Room Envelope Air Conditioning (AC) System,” by adding an Action Statement to the Limiting Conditions for Operation. The new Action Statement allows a finite time to restore one control room envelope AC subsystem to operable status and requires verification that the control room temperature remains <90 °F every 4 hours.

Date of publication of individual notice in Federal Register: (73 FR 55166) September 24, 2008.

Expiration date of individual notice: November 23, 2008.

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 Start Printed Page 65701License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing in connection with these actions was published in the Federal Register as indicated.

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

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

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: October 18, 2007, as supplemented by letter dated July 3, 2008.

Description of amendment request: The amendment changed the Oyster Creek Technical Specifications Section 4.5.M.1.e.1 regarding the mechanical snubber functional test acceptance test acceptance criteria. Specifically, the change replaced the snubber breakaway test with the drag force test.

Date of issuance: October 10, 2008.

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

Amendment No.: 270.

Facility Operating License No. DPR-16: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: June 17, 2008 (73 FR 34339). The supplement dated July 3, 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 amendment is contained in a Safety Evaluation dated October 10, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 29, 2007.

Brief description of amendment: The amendment consists of changes to Technical Specification Section 3.6.8, “Isolation Valve Seal Water (IVSW) System.” The amendment revises Surveillance Requirements (SR) 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change clarifies the wording of SR 3.6.8.2, and revises SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.

Date of issuance: October 3, 2008.

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

Amendment No. 220.

Renewed Facility Operating License No. DPR-23: The amendment revises the technical specifications and facility operating license.

Date of initial notice in Federal Register: January 15, 2008 (73 FR 2548). The Commission's related evaluation of the amendment is contained in a safety evaluation dated October 3, 2008.

No significant hazards consideration comments received: No.

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

NRC Branch Chief: Thomas H. Boyce.

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: July 17, 2007, as supplemented by letters dated August 7, 2007, and September 2, 2008.

Brief description of amendment: The amendment added a new license condition (43) on the control room envelope habitability program, revised Technical Specification (TS) requirements related to the control room envelope habitability in TS 3.7.3, “Control Room Fresh Air (CRFA) System,” and added the new TS 5.5.13, “Control Room Envelope Habitability Program.”

Date of issuance: October 14, 2008.

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

Amendment No: 178.

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

Date of initial notice in Federal Register: September 25, 2007 (72 FR 54473). The supplemental letters dated August 7, 2007, and September 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 October 14, 2008.

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: December 5, 2007, as supplemented by letters dated July 21 and August 28, 2008.

Brief description of amendment: The amendment changed Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR),” to add a reference to an analytical method that will be used to determine core operating limits. The new reference, NEDC-33383P, “GEXL97 Correlation Applicable to ATRIUM-10 Fuel,” will allow the licensee to use a Global Nuclear Fuel method to determine fuel assembly critical power of AREVA ATRIUM-10 fuel. Additionally, the amendment made an administrative change to an existing reference in TS 5.6.5.

Date of issuance: October 16, 2008.

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

Amendment No.: 179. Start Printed Page 65702

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

Date of initial notice in Federal Register: December 31, 2007 (72 FR 74358). The supplements dated July 21 and August 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 amendment is contained in a Safety Evaluation dated October 16, 2008.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: December 27, 2007, as supplemented by letter dated July 14, 2008.

Brief description of amendment: The amendment revised Technical Specifications (TS) Section 3.4.1, “RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to increase the minimum RCS flow rate from 341,100 to 354,000 gallons per minute. The increased flow rate supports a new analysis of a large break loss-of-coolant accident (LOCA). The new analysis is performed using an NRC-approved methodology set forth in Westinghouse Topical Report WCAP-16009-P-A, “Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM).” This methodology will be endorsed and reflected by a revision to TS Section 5.6.5, “Core Operating Limits Report (COLR).”

Date of issuance: October 17, 2008.

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

Amendment No.: 306.

Facility Operating License No. DPR-58: Amendment revised the Renewed Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 29, 2008 (73 FR 5223). The supplement dated July 14, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination published in the Federal Register on January 29, 2008.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 17, 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: January 28, 2008, as supplemented by letters dated July 28 and September 25, 2008.

Brief description of amendments: The amendments revised (1) Action 5 in Technical Specification (TS) 3.3.1, “Reactor Trip Instrumentation,” for one inoperable channel of extended range neutron flux instrumentation and (2) Action c in TS 3.4.1.4.2, “Reactor Coolant System, Cold Shutdown—Loops Not Filled.” The amendments do not complete the Nuclear Regulatory Commission staff's review of the licensee's proposed TS changes in the application. The remaining proposed TS changes to Action 5 will be addressed in a future letter to the licensee.

Date of issuance: October 16, 2008.

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

Amendment Nos.: Unit 1-187; Unit 2-174.

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: March 25, 2008 (73 FR 15788). The supplemental letters dated July 28 and September 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 as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 16, 2008.

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: October 26, 2007.

Brief description of amendment: The amendment revises the Technical Specifications (TS) to adopt TS Task Force (TSTF) Change Traveler TSTF-448, Revision 3, “Control Room Envelope Habitability.”

Date of issuance: October 8, 2008.

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

Amendment No.: 70.

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

Date of initial notice in Federal Register: August 29, 2008 (73 FR 51014). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 8, 2008.

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: October 24, 2007, as supplemented by letter dated August 7, 2008.

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

Date of issuance: October 9, 2008.

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

Amendment Nos.: 253, 234.

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: December 18, 2007 (72 FR 71717). The supplement dated August 7, 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 9, 2008.

No significant hazards consideration comments received: No. Start Printed Page 65703

Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement Or Emergency Circumstances)

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 for the amendment 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.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

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

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, 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 System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, 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 identify 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 Start Printed Page 65704for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner 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.[1] 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 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.

Each contention shall be given a separate numeric or alpha designation within one of the following groups:

1. Technical—primarily concerns/issues relating to technical and/or health and safety matters discussed or referenced in the applications.

2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.

3. Miscellaneous—does not fall into one of the categories outlined above.

As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.

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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

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

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

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

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

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

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later than Start Printed Page 6570511:59 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.

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

Date of amendment request: October 13, 2008.

Description of amendment request: The amendment revised the surveillance frequency for Technical Specification Surveillance Requirement 3.8.1.10 for the endurance test conducted every 2 years on the diesel generators.

Date of issuance: October 20, 2008.

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

Amendment No.: 255.

Facility Operating License No. DPR-26: Amendment revises the Technical Specifications and License.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Public notice of the proposed amendment was published in The Journal News newspaper, located in Westchester County, New York on October 17 and October 18, 2008. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated October 20, 2008.

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.

Start Signature

Dated at Rockville, Maryland, this 24th day October 2008.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature End Preamble

Footnotes

1.  To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel and discuss the need for a protective order.

Back to Citation

[FR Doc. E8-25882 Filed 11-3-08; 8:45 am]

BILLING CODE 7590-01-P