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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 March 26, 2009, to April 8, 2009. The last biweekly notice was published on April 7, 2009 (74 FR 15765).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rulemaking and Directives Branch, TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.

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

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

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one Start Printed Page 18252contention will not be permitted to participate as a party.

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

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

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated in August 28, 2007 (72 FR 49139). The E-Filing process requires participants to submit and serve all adjudicatory documents over the Internet or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms 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 electronic filing Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The help electronic filing Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

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

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a Presiding Officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. 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-Start Printed Page 182534209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: August 21, 2008.

Description of amendment request: The proposed amendments would revise the proposed license amendment implements Technical Specification Task Force (TSTF) Changes Travelers TSTF-479, Revision 0, “Changes to Reflect Revision of [Title 10 of the Code of Federal Regulations] 10 CFR 50.55a” and TSTF-497, Revision 0, “Limit Inservice Testing [IST] Program SR 3.0.2 Application to Frequencies of 2 Years or Less”. TSTF-479 and TSTF-497 revise the technical specification Administrative Controls section pertaining to requirements for the IST Program, consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as American Society of Mechanical Engineers (ASME) Code Class 1, Class 2, and Class 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 proposed change revises TS [Technical Specification] 5.5.8, “Inservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed change incorporates revisions to the ASME [American Society of Mechanical Engineers] Code as identified in the TSTFs [Technical Specification Task Force] referenced above.

The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.

Therefore, this 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 revises TS 5.5.8, “Inservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed change incorporates revisions to the ASME Code as identified in the TSTFs referenced above. The proposed change does not involve a modification to the physical configuration of the plant nor does it involve a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.

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

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

Response: No

The proposed change revises TS 5.5.8, “Inservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed change does not involve a modification to the physical configuration of the plant nor does it change the methods governingnormal plant operation. The proposed change incorporates revisions to the ASME Code as identified in the TSTFs referenced above.

The safety function of the affected pumps and valves will be maintained.

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

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

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

NRC Branch Chief: Melanie Wong.

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

Date of amendment request: February 16, 2009.

Description of amendment request: The Arkansas Nuclear One, Unit No. 1 (ANO-1) Technical Specification (TS) 5.5.16, “Reactor Building Leakage Rate Testing Program,” contains reactor building leak rate criteria for overall Type A, B, and C testing. However, TS 5.5.16 does not specify criteria for Type B air lock leakage testing. Entergy Operations, Inc., proposes to modify TS 5.5.16 to add criteria for overall air lock leakage testing and to adopt a low pressure test method relevant to the air lock door seals. This change is consistent with NUREG 1430, Revision 3.1, “Standard Technical Specifications (STS) for Babcock & Wilcox Plants.”

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 reactor building air locks are passive components integral to the reactor building structure and are not associated with accident initiators. Each air lock door is rated for and tested to the maximum calculated post-accident pressure of the reactor building. The air lock door seal pressure test is performed any time the air lock is used for reactor building access during modes of operation when reactor building integrity is required and prior to establishing reactor building integrity. The door seal test is intended to be a gross test to verify that the door seals were not damaged during the opening and closing cycle(s). This test does not replace the required overall barrel leakage test. Based on information provided by the air lock vendor, a test pressure of 10 psig [pounds per square inch gauge] is conservatively sufficient to perform this gross seal verification. This new acceptable leakage rate and test criteria are consistent with NUREG 1430, Rev. 3.1, Standard Technical Specifications for Babcock & Wilcox Plants (STS) and are applicable to ANO-1. While new to the TSs, the ANO-1 program for ensuring compliance with 10 CFR 50, Appendix J has verified leakage within the proposed limiting values.

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 physical changes to the facility are initiated by the proposed change. In addition, the proposed change has no affect on plant configuration, or method of operation of plant structures, systems, or components.

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 does not increase the allowable overall air lock leakage rate, nor Start Printed Page 18254affect the acceptance criteria of the overall integrated containment leakage rate as currently tested to in accordance with the ANO-1 containment leakage rate test program. All of the changes are bounded by existing analyses for all evaluated accidents and do not create any situations that alter the assumptions used in these analyses.

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. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: March 10, 2009.

Description of amendment request: The proposed amendment consists of changes to Technical Specification (TS) 3.4.9, “Pressurizer,” which contains a maximum and minimum level for the pressurizer. The licensee proposes to delete the minimum level requirement. This change is consistent with NUREG 1430, Rev. 3.1, “Standard Technical Specifications [STS] for Babcock and Wilcox Plants.”

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 minimum Pressure level limit currently specified in the TSs does not act to ensure specified fuel design limits are protected. Accident and transient analyses assume lowering or a loss of Pressurizer level. Safety systems are designed and maintained available to mitigate the consequences of an accident or transient that may involve a loss of Reactor Coolant System (RCS) inventory. None of these systems rely upon a predetermined minimum Pressurizer level in order to perform their intended function. Furthermore, the minimum Pressure level limit is unrelated to any anticipated accident initiator.

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 physical changes to the facility are initiated by the proposed change. In addition, the proposed change has no affect on plant configuration, or method of operation of plant structures, systems, or components.

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.

Installed automatic control systems will continue to maintain Pressurizer level at a predetermined setpoint and are independent of a prescribed minimum TS level limit. The deletion of the current TS limit has no impact on guidance or operational response to pressurizer level deviations. Furthermore, the minimum Pressure level limit is not an assumed value for accident prevention or mitigation in the [Arkansas Nuclear One, Unit 1] [Safety Analysis Report].

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.

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

Date of amendment request: February 25, 2009.

Description of amendment request: The proposed change removes the reactor coolant system (RCS) structural integrity requirements contained in Technical Specification (TS) 3/4.4.8, which specifies requirements relating to the structural integrity of American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This specification is redundant to the requirements contained within Title 10 of the Code of Federal Regulations (10 CFR) section 50.55a, “Codes and standards.” With this proposed change, RCS pressure boundary structural integrity will continue to be maintained by compliance with 10 CFR 50.55a, as implemented through the Limerick Generating Station, Units 1 and 2, Inservice Inspection 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, with NRC edits in brackets:

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 remove the RCS structural integrity controls from the TSs does not impact any mitigation equipment or the ability of the RCS pressure boundary to fulfill any required safety function. Since no accident mitigation [equipment] or initiators are impacted by this change, no design basis accidents are affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 will not alter the plant configuration or change the manner in which the plant is operated. No new failure modes are being introduced by the proposed change.

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

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

Response: No.

Removal of TS 3/4.4.8 from the TSs does not reduce the controls that are required to maintain the RCS pressure boundary for ASME Code Class 1, 2, or 3 components.

No equipment or RCS safety margins are impacted due to the proposed change. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Harold K. Chernoff.Start Printed Page 18255

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

Date of amendment request: February 16, 2009.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) by removing the structural integrity requirements contained in TS 3/4.4.10 and the associated TS bases from the TSs. Removal of TS 3/4.4.10 is consistent with NUREG-1431, Revision 3.0, “Standard Technical Specifications Westinghouse Plants,” in that it does not meet the criteria of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36, “Technical Specifications,” for inclusion in the TSs. The proposed amendment would also relocate the reactor coolant pump (RCP) flywheel inspection requirements in Surveillance Requirement (SR) 4.4.10 to SR 4.0.5, and would revise the RCP flywheel inspection interval from 10 years to 20 years. The RCP flywheel inspection interval change is consistent with Nuclear Regulatory Commission approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-421, “Revision to RCP Flywheel Inspection Program (WCAP-15666).”

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 to remove structural integrity controls from the TSs does not impact any mitigation equipment or the ability of the RCS [reactor coolant system] pressure boundary to fulfill any required safety function. The proposed change will continue to ensure the requirements of 10 CFR 50.55a [“Codes and standards”] are maintained as specified in TS 4.0.5. Since no accident mitigation or initiators are impacted by this change, no design basis accidents are affected.

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

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

Response: No.

The proposed change will not alter the plant configuration or change the manner in which the plant is operated. Structural integrity will continue to be maintained as required by 10 CFR 50.55a and specified in TS 4.0.5. No new failure modes are being introduced by the proposed change.

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 the margin of safety?

Response: No.

Removal of TS 3/4.4.10 from the TSs does not reduce the controls that are required to maintain the structural integrity of ASME [American Society of Mechanical Engineers] Code Class 1, 2, or 3 components. No safety margins are impacted due to the proposed change.

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

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Thomas H. Boyce.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station Unit Nos. 1 and 2 (NMP 1 and 2), Oswego County, New York

Date of amendment request: February 11, 2009.

Description of amendment request: The proposed amendment would delete those portions of the Technical Specifications (TSs) superseded by 10 CFR Part 26, Subpart I. The proposed change is consistent with Nuclear Regulatory Commission (NRC)-approved Revision 0 to TS Task Force (TSTF) Change Traveler, TSTF-511-A, “Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.” The availability of the TS improvement was announced in the Federal Register (FR) on December 30, 2008 (73 FR 79923) as part of the consolidated line item improvement process. The licensee concluded that the no significant hazards consideration determination as presented in the FR notice is applicable to NMP 1 and 2.

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

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

The proposed change removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Removal of the Technical Specification requirements will be performed concurrently with the implementation of the 10 CFR Part 26, Subpart I, requirements. The proposed change does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which the SSCs are operated, maintained, modified, tested, and inspected. Worker fatigue is not an initiator of any accident previously evaluated. Worker fatigue is not an assumption in the consequence mitigation of any accident previously evaluated. Therefore, it is concluded that 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 removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Working hours will continue to be controlled in accordance with NRC requirements. The new rule allows for deviations from controls to mitigate or prevent a condition adverse to safety or as necessary to maintain the security of the facility. This ensures that the new rule will not unnecessarily restrict working hours and thereby create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. Therefore, the proposed 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 a Margin of Safety

The proposed change removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Working hours will continue to be controlled in accordance with NRC requirements. The proposed change does not involve any physical changes to the plants or alter the manner in which plant systems are operated, maintained, modified, tested, and inspected. The proposed change does not alter the manner in which safety limits, limiting safety Start Printed Page 18256system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plants and to maintain the plants in a safe shutdown condition. Removal of plant-specific Technical Specification administrative requirements will not reduce a margin of safety because the requirements in 10 CFR Part 26 are adequate to ensure that worker fatigue is managed. Therefore, 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: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: February 20, 2009.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) requirements related to control room envelope habitability in TS 3.7.3, “Plant Systems Control Room Emergency Outside Air Supply (CREOAS) System,” and TS Section 5.5, “Administrative Controls Programs and Manuals.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments to revise the plant specific TS, to strengthen TS requirements regarding control room envelope (CRE) habitability by changing the action and surveillance requirements associated with the limiting condition for operation operability requirements for the CRE emergency ventilation system. A new TS administrative controls program on CRE habitability is being added, including a model safety evaluation and model no significant hazards consideration 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 January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the model NSHC determination in its application dated February 20, 2009.

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

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

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief : Mark Kowal.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: March 23, 2009.

Description of amendment request: The proposed amendment would delete those portions of the Technical Specifications (TSs) superseded by Part 26, Subpart I of Title 10 of the Code of Federal Regulations (10 CFR). This change incorporates NRC approved Revision 0 of Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-511, “Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.” The availability of this TS improvement was announced as part of the consolidated line item improvement process (CLIIP) in the Federal Register on December 30, 2008 (73 FR 79923).

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:

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

The proposed change removes technical specification restrictions on working hours for personnel who perform safety related functions. The technical specification Start Printed Page 18257restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Removal of the technical specification requirements will be performed concurrently with the implementation of 10 CFR Part 26, Subpart I, requirements. The proposed change does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. Worker fatigue is not an assumption in the consequence mitigation of any accident previously evaluated.

Therefore, it is concluded that 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 removes technical specification restrictions on working hours for personnel who perform safety related functions. The technical specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Working hours will continue to be controlled in accordance with NRC requirements. The new rule allows for deviations from controls to mitigate or prevent a condition adverse to safety or as necessary to maintain the security of the facility. This ensures that the new rule will not unnecessarily restrict working hours and thereby create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or effect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected.

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

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

The proposed change removes technical specification restrictions on working hours for personnel who perform safety related functions. The technical specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26.

The proposed change does not involve any physical changes to the plant or alter the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions or operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

Removal of plant-specific technical specification administrative requirements will not reduce a margin of safety because the requirements in 10 CFR Part 26 are adequate to ensure that worker fatigue is managed.

Therefore, 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: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, Baltimore, MD 21202.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment request: December 29, 2008.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” and TS 3.8.5, “DC Sources—Shutdown.” Specifically, this amendment would revise the battery connection resistance limits in Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 from 150 micro-ohms (150E-6 ohm) to 69 micro-ohms (69E-6 ohm). TS 3.8.5 is affected by virtue of SR 3.8.5.1 invoking both SR 3.8.4.2 and SR 3.8.4.5 for DC sources that are required to be operable in Modes 5 and 6.

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 replaces a battery surveillance limit with a value based on voltage drop calculations for each of the four battery subsystems at Callaway under both normal operating and accident load profiles. The new value is more conservative, as well as being more appropriate, as an acceptance criterion for verifying battery operability pursuant to SR 3.8.4.2 and SR 3.8.4.5, thus providing greater assurance that the batteries can perform their specified safety functions with regard to accident mitigation.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. All design, material, and construction standards that were applicable prior to this amendment request will be maintained. There will be no changes to any design or operating limits.

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

The proposed change does not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions.

All accident analysis acceptance criteria will continue to be met with the proposed change. The proposed change will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The applicable radiological dose criteria will continue to be met.

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.

There are no proposed design changes nor are there any changes in the method by which any safety-related plant structure, system, or component (SSC) performs its specified safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. Equipment performance necessary to fulfill safety analysis missions will be unaffected. The proposed change will not alter any assumptions required to meet the safety analysis acceptance criteria.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment.

The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems.

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

3. Does the proposed change does not involve a significant reduction in a margin of safety?

Response: No.

There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (FQ), nuclear enthalpy rise hot channel factor (FΔH), loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met.Start Printed Page 18258

The proposed change does not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications; however, the acceptance criterion for the specified battery resistance surveillances will be more restrictive. None of the acceptance criteria for any accident analysis will be changed.

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: John O'Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 4, 2009.

Description of amendment request: The proposed amendment consists of changes to the approved fire protection program as described in Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR). Specifically, a deviation from certain technical requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix R, Section III.G.2, as documented in Appendix 9.5E of the WCGS USAR, is requested regarding the use of operator manual actions in lieu of meeting circuit separation protection criteria. Table 3-1 of the submittal dated March 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090771269), identifies the proposed feasible and reliable operator manual actions requested for permanent approval and Table 3-2 of the submittal identifies the proposed feasible operator manual actions requested for approval on an interim basis. The interim operator actions will be eliminated with the implementation of associated design change package.

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 design function of structures, systems and components are not impacted by the proposed change. The proposed change involves the performance of operator manual actions to achieve and maintain safe shutdown in the event of a fire outside of the control room and will not initiate an event. The proposed change does not increase the probability of occurrence of a fire or any other accident previously evaluated.

The proposed operator manual actions are feasible and reliable and demonstrate that the plant can be safely shutdown in the event of a fire. No significant consequences result from the performance of the proposed change.

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 design function of structures, systems and components are not impacted by the proposed change. The proposed change involves the performance of operator manual actions to achieve and maintain safe shutdown in response to a fire outside of the control room. The operator manual actions do not involve new failure mechanisms or malfunctions that can initiate a new accident.

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.

For the permanent operator manual actions, adequate time is available to perform the proposed operator manual actions to account for uncertainties in estimates of the time available and in estimates of how long it takes to diagnose and execute the actions. The actions have been verified that they can be performed through demonstration and the actions are proceduralized. The proposed actions are feasible and reliable and demonstrate that the plant can be safely shutdown in the event of a fire.

For the interim operator manual actions adequate time is available to feasibly perform the proposed operator manual actions and a compensatory measure fire watch is provided for the affected area as an added defense in depth fire protection measure.

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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 6, 2009.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 5.2.2.d regarding the requirement to develop and implement administrative procedures to limit the working hours of personnel who perform safety-related functions. The requirements of TS 5.2.2 have been superseded by Title 10 of the Code of Federal Regulations (10 CFR) Part 26, Subpart I. The change is consistent with U.S. Nuclear Regulatory Commission (NRC)-approved Revision 0 to Technical Specification Task Force (TSTF) Improved Technical Specification Change Traveler, TSTF-511, “Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.”

The NRC staff issued a “Notice of Availability of Model Safety Evaluation, Model No Significant Hazards Determination, and Model Application for Licensees That Wish To Adopt TSTF-511, Revision 0, `Eliminate Working Hour Restrictions From TS 5.2.2 To Support Compliance With 10 CFR Part 26,' ” in the Federal Register on December 30, 2008 (73 FR 79923). The notice included a model safety evaluation, a model no significant hazards consideration (NSHC) determination, and a model license amendment request, using the consolidated line item improvement process. In its application dated March 6, 2009, the licensee affirmed the applicability of the model NSHC determination, which is presented below.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC determination 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 removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Removal of the Technical Specification requirements will be performed concurrently with the implementation of the 10 CFR Part Start Printed Page 1825926, Subpart I, requirements. The proposed change does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. Worker fatigue is not an initiator of any accident previously evaluated. Worker fatigue is not an assumption in the consequence mitigation of any accident previously evaluated. Therefore, it is concluded that 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 removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. Working hours will continue to be controlled in accordance with NRC requirements. The new rule allows for deviations from controls to mitigate or prevent a condition adverse to safety or as necessary to maintain the security of the facility. This ensures that the new rule will not unnecessarily restrict working hours and thereby create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or [a]ffect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed change removes Technical Specification restrictions on working hours for personnel who perform safety related functions. The Technical Specification restrictions are superseded by the worker fatigue requirements in 10 CFR Part 26. The proposed change does not involve any physical changes to plant or alter the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. Removal of plant-specific Technical Specification administrative requirements will not reduce a margin of safety because the requirements in 10 CFR Part 26 are adequate to ensure that worker fatigue is managed.

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

The NRC staff has reviewed the analysis adopted by the licensee and, based on this review, it appears that the 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

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

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

Date of application for amendments: June 23, 2008.

Brief description of amendments: This request modifies the subject Technical Specifications (TSs) and Bases by changing the logic configuration of TS Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” Function 5.b.(5), “Turbine Trip and Feedwater Isolation, Feedwater Isolation, Doghouse Water Level—High High.” The existing one-out-of-one (1/1) logic per train per doghouse is being modified to a two-out-of-three (2/3) logic per train per doghouse. The proposed change will improve the overall reliability of this function and will reduce the potential for spurious actuations.

Date of issuance: April 2, 2009.

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

Amendment Nos.: 249/243.

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

Date of initial notice in Federal Register: February 24, 2009 (74 FR 8276).

The Commission's related evaluation, state consultation, and final no significant hazards consideration determination of the amendments are contained in a Safety Evaluation dated April 2, 2009.

No significant hazards consideration comments received: No.

Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: March 20, 2008.

Brief description of amendments: The proposed amendments would revise the McGuire licensing basis by adopting the Alternative Source Term (AST) radiological analysis methodology as Start Printed Page 18260allowed by 10 CFR 50.67, “Accident source term,” for the Loss of Coolant Accident. This amendment request represents full scope implementation of the AST as described in Nuclear Regulatory Commission (NRC) Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0.” Selective implementation of AST for the McGuire Fuel Handling Accidents was approved by the NRC on December 22, 2006. There are no changes proposed to the McGuire Technical Specifications within this amendment request. The application of the AST methodology to the Loss of Coolant Accident (LOCA) radiological analysis will allow McGuire to resolve the Control Room envelope degraded boundary condition as discussed in McGuire's response to NRC Generic Letter 2003-01, “Control Room Habitability,” dated February 19, 2004.

By separate amendment request dated January 22, 2008, Duke proposed to revise the McGuire Technical Specification (TS) requirements related to control room envelope habitability in TS 3.7.9, “Control Room Area Ventilation System.” The proposed changes are consistent with the Industry and NRC-approved Technical Specification Task Force (TSTF) change TSTF-448, Control Room Habitability, Revision 3 and the NRC Consolidated Line Item Improvement Process (CLIIP).

Duke has performed a review of all McGuire License Amendment Requests (LAR) currently under review by the NRC for impacts to this AST LAR. None of these LARs impact any assumptions or results of the LOCA AST radiological analysis.

Date of issuance: March 31, 2009.

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

Amendment Nos.: 251 and 231.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: The amendments revised the license.

Public comments requested as to proposed no significant hazards consideration (NSHC): The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination by March 30, 2009. No comments have been received to date. However, the notice also provided an opportunity to request a hearing by April 28, 2009, but indicated that if the Commission make a final NSHC determination, any such hearing would take place after issuance of the amendment.

Date of initial notice in Federal Register: February 27, 2009 (74 FR 9009).

The supplements dated May 28, 2008, October 6, 2008, December 17, 2008 and February 12, 2009, 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 March 31, 2009.

No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: December 8, 2008.

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

Date of issuance: April 1, 2009.

Effective date: As of the date of issuance and shall be implemented within 15 days from the date of issuance.

Amendment No.: 162.

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

Date of initial notice in Federal Register : January 27, 2008 (74 FR 4770).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 1, 2009.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 21, 2008.

Brief description of amendment: The amendment deleted the exception to Limiting Condition for Operation (LCO) 3.0.4 to the 30-day allowable outage time of the Startup No. 2 Transformer and corrected a spelling error in Technical Specification (TS) 3.8.1. The NRC approved the adoption of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints,” for ANO-1 in TS Amendment 232 dated April 2, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080600006). The intent of TSTF-359 was to eliminate exceptions to LCO 3.0.4 within individual specifications and provide requirements within LCO 3.0.4 to control mode changes when TS-required equipment is inoperable. The licensee omitted deleting this LCO 3.0.4 exception in its October 22, 2007 (ADAMS Accession No. ML073030542), amendment request to adopt TSTF-359.

Date of issuance: March 30, 2009.

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

Amendment No.: 236.

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

Date of initial notice in Federal Register : October 21, 2008 (73 FR 62563).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2009.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: December 16, 2008, as supplemented by letter dated February 19, 2009.

Brief description of amendment: This amendment request would revise the Technical Specifications Section 2.1.2, Safety Limit Minimum Critical Power Ratio (SLMCPR) for two-loop and single-loop operation.

Date of issuance: March 26, 2009.

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

Amendment No.: 232.

Facility Operating License No. DPR-35: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : January 23, 2009 (74 FR 4250).

The supplemental letter dated February 19, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 26, 2009.

No significant hazards consideration comments received: No.Start Printed Page 18261

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of application for amendment: April 9, 2008, as supplemented by letter dated October 1, 2008.

Brief description of amendment: The amendments revise Technical Specifications (TSs) 5.5.6, Pre-Stressed Concrete Containment Tendon Surveillance Program, and 5.6.8, Tendon Surveillance Report, for consistency with the requirements of Title 10 Code of Federal Regulations (10 CFR) Section 50.55a, Codes and standards, paragraph (g)(4) for components classified as American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class CC, by replacing the reference to the specific ASME Code year for the tendon surveillance program with a requirement to use the applicable ASME Code and addenda as required by 10 CFR 50.55a.

Date of issuance: March 26, 2009.

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

Amendment Nos.: Braidwood Unit 1-158; Braidwood Unit 2-158; Byron Unit No. 1-163; and Byron Unit No. 2-163.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revise the TSs and Licenses.

Date of initial notice in Federal Register : July 1, 2008 (73 FR 37504).

The October 1, 2008, supplemental letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 2, 2008.

Brief description of amendment: The amendment requested to amend the CPS Unit No. 1 Technical Specifications (TS) to relocate the TS surveillance requirement (SR) 3.8.3.6 from the TS 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. The request is submitted consistent with the guidance contained in Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force Report 2 (TSTF-2).

Date of issuance: April 2, 2009.

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

Amendment No.: 186.

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

Date of initial notice in Federal Register : November 4, 2008 (73 FR 65687) and January 27, 2009 (74 FR 4771). The notice on January 27, 2009, was inadvertently placed in the Federal Register a second time and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

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

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County, Pennsylvania

Date of application for amendment: November 7, 2008.

Brief description of amendment: The amendment modifies the method used to calculate the available net positive suction head (NPSH) for the BVPS-2 recirculation spray (RS) pumps as described in the BVPS-2 Updated Final Safety Analysis Report (UFSAR). The BVPS-2 UFSAR takes credit for containment overpressure by allowing for the difference between containment total pressure and the vapor pressure of the water in the containment sump in the available NPSH calculation.

Date of issuance: March 26, 2009.

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

Amendment No.: 167.

Facility Operating License No. NPF-73: The amendment revised the License and the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register : December 16, 2008 (73 FR 76411).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 21, 2008.

Brief description of amendment: The amendment modifies Technical Specification 5.6.3, “Radioactive Effluent Release Report,” by changing the required annual submittal date for the report from “within 90 days of January 1 of each year” (i.e., prior to April 1), to “prior to May 1 of each year.”

Date of issuance: March 30, 2009.

Effective date: As of the date of issuance.

Amendment Nos.: 308 (CNP-1), 290 (CNP-2).

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Renewed Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : December 16, 2008 (73 FR 76412).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2009.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 22, 2008, as supplemented by letter dated March 6, 2009.

Brief description of amendment: The amendment modifies the Technical Specification (TS) 2.7, “Electrical Systems,” Limiting Condition for Operation (LCO) 2.7(2)j related to the allowed outage time for the Emergency Diesel Generators (EDGs). The change clarifies LCO 2.7(2)j such that a single period of inoperability for one EDG is limited to 7 consecutive days and that the cumulative total time of inoperability for both EDGs during any calendar month cannot exceed 7 days.

Date of issuance: March 27, 2009.

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

Amendment No.: 258.

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

Date of initial notice in Federal Register : June 17, 2008 (73 FR 34342). The supplemental letter dated March 6, 2009, provided additional information that clarified the application, did not expand the scope of the application as Start Printed Page 18262originally 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 March 27, 2009.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: April 3, 2008, as supplemented by letters dated June 20, October 1, November 6, and December 16, 2008.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System,” to remove Surveillance Requirement (SR) 3.7.5.6, and revised TS 3.7.6, “Condensate Storage Tank (CST) and Fire Water Storage Tank (FWST),” to remove the FWST level requirements, revise the CST level requirements, and revise TS 3.7.6 to be consistent with the NUREG-1431, “Standard Technical Specifications (STS).” Specifically, these changes reflect design changes made to the CSTs and are necessary to support the on-line refurbishment of the FWST and replacement of the recirculation piping for the fire water pumps. The design changes to the CSTs are intended to eliminate the reliance on the FWST for additional seismically-qualified feedwater supply and thus, make the existing TS requirements for the FWST unnecessary.

Date of issuance: March 30, 2009.

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

Amendment Nos.: Unit 1-204; Unit 2-205.

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

Date of initial notice in Federal Register : July 29, 2008 (78 FR 43956). The supplemental letters dated June 20, October 1, November 6, and December 16, 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 March 30, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 18, 2008.

Brief description of amendment: The amendment revised WBN Unit 1 Technical Specification 3.8.7, “Inverters—Operating.” The amendment revised the requirement to two inverters for each of the four channels.

Date of issuance: March 24, 2009.

Effective date: As of the date of issuance and shall be implemented within 240 days of issuance.

Amendment No.: 76.

Facility Operating License No. NPF-90: Amendment revises the Technical Specification 3.8.7 and Updated Final Safety Analysis Report.

Date of initial notice in Federal Register : November 4, 2008 (73 FR 65697).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 24, 2009.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 14, 2007, as supplemented by letters dated April 18, May 9, June 15, August 31, September 12 and 20, October 16, November 16, two letters dated December 14, and December 18, 2007; two letters dated January 18, January 31, February 26 and 28, March 14, April 26, May 14, June 19, and July 31, 2008; and January 16 and 29, and February 17 and 27, 2009.

Brief description of amendment: The amendment revised the licensing basis for the Main Steam and Feedwater Isolation System (MSFIS) controls to incorporate field programmable gate array technology. Other related changes requested in the March 14, 2007, application were previously approved in Amendment No. 174, dated August 28, 2007, Amendment No. 175, dated March 3, 2008, Amendment No. 176, dated March 21, 2008, and Amendment No. 177, dated April 3, 2008.

Date of issuance: March 31, 2009.

Effective date: Effective as of date of issuance and shall be implemented before entry into Mode 3 in the restart from Refueling Outage 17.

Amendment No.: 181.

Renewed Facility Operating License No. NPF-42. The amendment revised the Operating License.

Date of initial notice in Federal Register : June 19, 2007 (72 FR 33785). The supplemental letters dated April 18, May 9, June 15, August 31, September 12 and 20, October 16, November 16, two letters dated December 14, and December 18, 2007; two letters dated January 18, January 31, February 26 and 28, March 14, April 26, May 14, June 19, and July 31, 2008; and January 16 and 29, and February 17 and 27, 2009, 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 March 31, 2009.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 10th day of April 2009.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature End Preamble

[FR Doc. E9-8832 Filed 4-20-09; 8:45 am]

BILLING CODE 7590-01-P