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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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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) 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 November 17 to November 30, 2011. The last biweekly notice was published on November 29, 2011 (76 FR 73727).

ADDRESSES:

Please include Docket ID NRC-2011-0285 in the subject line of your comments. Comments submitted in writing or in electronic form will be posted on the NRC Web site and on the Federal rulemaking Web site http://www.regulations.gov. Because your comments will not be edited to remove any identifying or contact information, the NRC cautions you against including any information in your submission that you do not want to be publicly disclosed.

The NRC requests that any party soliciting or aggregating comments received from other persons for submission to the NRC inform those persons that the NRC will not edit their comments to remove any identifying or contact information, and therefore, they should not include any information in their comments that they do not want publicly disclosed.

You may submit comments by any one of the following methods.

Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-2011-0285. Address questions about NRC dockets to Carol Gallagher (301) 492-3668; email Carol.Gallagher@nrc.gov.

Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Fax comments to: RADB at (301) 492-3446.

You can access publicly available documents related to this notice using the following methods:

NRC's Public Document Room (PDR): The public may examine and have copied for a fee publicly available documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

NRC's Agencywide Documents Access and Management System (ADAMS): Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. From this page, the public can gain entry into ADAMS, which provides text and image files of NRC's public documents. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC's PDR reference staff at 1-(800) 397-4209, (301) 415-4737, or by email to pdr.resource@nrc.gov.

Federal Rulemaking Web Site: Public comments and supporting materials related to this notice can be found at http://www.regulations.gov by searching on Docket ID: NRC-2011-0285.

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 Title 10 of the Code of Federal Regulations (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; (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 Start Printed Page 77566will 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.

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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20874. NRC regulations are accessible electronically from the NRC Library on the NRC Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. 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 requestor/petitioner 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 requestor/petitioner 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 requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/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. 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 requestor/petitioner to relief. A requestor/petitioner 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, 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, then 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 (72 FR 49139, August 28, 2007). 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 an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request (1) A digital identification (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 (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions Start Printed Page 77567should 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 documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing 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 email notice confirming receipt of the document. The E-Filing system also distributes an email 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 using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by email at MSHD.Resource@nrc.gov, or by a toll-free call at 1-(866) 672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, 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. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the 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.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20874. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-(800) 397-4209, (301) 415-4737, or by email to pdr.resource@nrc.gov.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: October 18, 2011.

Description of amendment request: The proposed amendment involves administrative changes. The proposed changes include correcting typographical errors, removing unwarranted formatting, clarifying symbols and pages, reformatting of previously deleted pages, incorporating a consistent abbreviation of average reactor coolant temperature, deleting notes that are no longer applicable, and replacing certain drawing figures with versions that are more clear.

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

Response: No.

No physical changes to the facility will occur as a result of this proposed amendment. The proposed changes will not alter the physical design or operational procedures associated with any plant structure, system, or component. The proposed changes are administrative in nature and have no affect on plant operation.

Therefore, the proposed changes do 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 changes are administrative in nature. The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function.

Therefore, the proposed changes do 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 changes [maintain compliance with the requirements contained in 10 CFR 50.36, “Technical specifications.”] The proposed changes are administrative in nature. The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant.

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, with the NRC edits above, it Start Printed Page 77568appears 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.

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

Date of amendment request: August 11, 2011.

Description of amendment request: The proposed amendments would make changes to the diesel fuel oil license bases and amend technical specifications (TS) 3.7.8, “Cooling Water (CL) System” and 3.8.3, “Diesel Fuel Oil.” The proposed TS changes would revise current requirements to reflect the addition of the license bases, resolve non-conservative emergency diesel generator fuel oil supply volumes, incorporate portions of Technical Specification Task Force Traveler 501, “Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control,” and provide administrative changes to the 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

This license amendment request proposes addition of a diesel fuel oil supply license basis and revision of the associated Technical Specifications to require an adequate emergency diesel generator and diesel driven cooling water pump fuel oil supply for mitigation of a design basis accident with a loss of offsite power. This license amendment request also proposes to: adopt provisions of Technical Specifications Task Force (TSTF) industry traveler 501 (TSTF-501) to specify diesel fuel oil supply requirements as required days for the supply and relocate the corresponding volume to the Technical Specification Bases; and, make minor wording changes to improve conformance to the content guidance of NUREG-1431, “Standard Technical Specifications, Westinghouse Plants.”

The emergency diesel generators, diesel driven cooling water pumps and their supporting diesel fuel oil storage systems are not accident initiators and therefore the proposed diesel fuel oil supply license basis addition and proposed Technical Specification changes do not involve an increase in the probability of an accident.

The proposed change to the emergency diesel generator fuel oil supply license basis and the associated Technical Specification changes will assure that the emergency diesel generator's diesel driven cooling water pumps perform their required design basis accident mitigation safety function with a loss of offsite power. Since the emergency diesel generators will provide required electrical power as assumed in the accident analyses and the cooling water diesel will provide cooling water as assumed in the accident analyses, the results of the previous accident analyses are not changed and the license basis changes proposed in this license amendment request do not involve a significant increase in the consequences of an accident.

Specification of the diesel fuel oil supply requirements as required days supply in accordance with TSTF-501 continues to assure an adequate quantity of diesel fuel oil is required to be stored; the emergency diesel generators and diesel driven cooling water pumps will have sufficient diesel fuel oil to mitigate a design basis accident with a loss of offsite power, as assumed in the accident analyses, until the fuel supply can be replenished; and therefore, this change does not involve a significant increase in the consequences of an accident.

The proposed minor Technical Specification wording changes to improve alignment with the content guidance of NUREG-1431 are administrative and thus do not involve an increase in the consequences of an accident.

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

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

Response: No.

This license amendment request proposes addition of a diesel fuel oil supply license basis and revision of the associated Technical Specifications to require an adequate emergency diesel generator and diesel driven cooling water pump fuel oil supply for mitigation of a design basis accident with a loss of offsite power. This license amendment request also proposes to: adopt provisions of Technical Specifications Task Force (TSTF) industry traveler 501 (TSTF-501) to specify diesel fuel oil supply requirements as required days for the supply and relocate the corresponding volume to the Technical Specification Bases; and, make minor wording changes to improve conformance to the content guidance of NUREG-1431, “Standard Technical Specifications, Westinghouse Plants.”

The proposed diesel fuel oil supply license basis change and the associated Technical Specification changes assure that each emergency diesel generator and diesel driven cooling water pump has an adequate supply of diesel fuel oil, assuming an active single failure, to mitigate a design basis accident with a loss of offsite power until the fuel oil supply can be replenished. The proposed license basis change and associated Technical Specification changes do not create new failure modes or mechanisms and no new accident precursors are generated. The proposed specification of the diesel fuel oil supply requirements as required days supply in accordance with TSTF-501 does not create new failure modes or mechanisms and does not generate new accident[s]. These proposed changes do not challenge the performance or integrity of any safety-related system. Surveillance requirements for the emergency diesel generator and diesel driven cooling water pump fuel oil supplies will continue to demonstrate that the Limiting Conditions for Operation are met and the emergency diesel generators and diesel driven cooling water pumps have adequate supplies of diesel fuel oil to perform their safety functions.

The proposed minor Technical Specification wording changes to improve alignment with the content guidance of NUREG-1431 are administrative and thus do not create the possibility of a new or different kind of accident.

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

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

Response: No.

This license amendment request proposes addition of a diesel fuel oil supply license basis and revision of the associated Technical Specifications to require an adequate emergency diesel generator and diesel driven cooling water pump fuel oil supply for mitigation of a design basis accident with a loss of offsite power. This license amendment request also proposes to: adopt provisions of Technical Specifications Task Force (TSTF) industry traveler 501 (TSTF-501) to specify diesel fuel oil supply requirements as required days for the supply and relocate the corresponding volume to the Technical Specification Bases; and, make minor wording changes to improve conformance to the content guidance of NUREG-1431, “Standard Technical Specifications, Westinghouse Plants.”

The proposed diesel fuel oil supply licensing basis addition and the associated Technical Specification changes involve the addition of a new requirement to assure that each emergency diesel generator and diesel driven cooling water pump has an adequate supply of diesel fuel oil, assuming an active single failure, to mitigate a design basis accident with a loss of offsite power until the fuel oil supply can be replenished. The current license basis for mitigation of an external flood without a single failure will be maintained. Therefore, margins of safety are increased and thus no margin of safety is reduced due to these changes.

Specification of the diesel fuel oil supply requirements as required days supply in accordance with TSTF-501 continues to assure an adequate quantity of diesel fuel oil is required to be stored and thus does not reduce a margin of safety.Start Printed Page 77569

The proposed minor Technical Specification wording changes to improve alignment with the content guidance of NUREG-1431 are administrative and thus do not involve a significant reduction in a margin of safety.

The proposed Technical Specification changes do not adversely affect the availability, operability, or performance of safety-related systems and components: the emergency diesel generators [and] diesel driven cooling water pumps will continue to perform their safety functions. The ability of operable structures, systems, and components to perform their designated safety functions are unaffected by these proposed changes. The operability requirements of the proposed Technical Specifications are consistent with the initial condition assumptions of the safety analyses, and the Surveillance requirements for the emergency diesel generator and diesel driven cooling water pump fuel oil supplies will assure that the Limiting Conditions for Operation are met and the emergency diesel generator's diesel driven cooling water pumps have adequate supplies of diesel fuel oil to perform their safety functions.

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 requests involve no significant hazards consideration.

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Acting Branch Chief: Terry A. Beltz.

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

Date of amendment request: June 1, 2011.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System,” TS 3.6.6, “Containment Spray and Cooling Systems,” TS 3.8.1, “AC [Alternating Current] Sources—Operating,” TS 3.8.9, “Distribution Systems—Operating,” and TS 1.3, “Completion Times,” Example 1.3-3. These changes are consistent with Technical Specification Task Force (TSTF) Change Travelers TSTF-245, Revision 1, “AFW Train Operable when in Service,” TSTF-340, Revision 3, “Allow 7 day Completion Time for a Turbine-driven AFW Pump Inoperable,” TSTF-412, Revision 3, “Provide Actions for One Steam Supply to Turbine Driven AFW/EFW [Emergency Feedwater] Pump Inoperable,” and TSTF-439, Revision 2, “Eliminate Second Completion Times Limiting Time From Discovery of Failure to Meet an LCO [Limiting Condition for Operation].”

Specifically, the changes consistent with TSTF-245, Revision 1, and TSTF-340, Revision 3, would revise TS 3.7.5 to clarify the operability of an AFW train during alternate alignments and provide added flexibility in Mode 3 to repair and test the turbine-driven AFW (TDAFW) pump following a refueling outage. The changes consistent with TSTF-412, Revision 3, would revise TS 3.7.5 to establish conditions, required actions, and completion times for the condition where one steam supply to the TDAFW is inoperable concurrent with an inoperable motor-driven AFW (MDAFW) train. The TSTF-412, Revision 3, Notice of Availability was published in the Federal Register on July 17, 2007 (72 FR 39089), using the consolidated line item improvement process (CLIIP). The changes consistent with TSTF-439, Revision 2, would remove second completion times from TS Example 1.3-3; TS 3.6.6 Required Actions A.1, A.2, and C.1; TS 3.7.5 Required Actions A.1 and B.1; TS 3.8.1 Required Actions A.2 and B.4; and TS 3.8.9 Required Actions A.1, B.1, and C.1. In addition, the amendment would add a new Condition B, required actions, and completion times to TS 3.7.5 to provide specific actions to be taken when automatic control of the MDAFW level control valves is not functional.

Basis for proposed no significant hazards consideration determination: For the proposed changes related to TSTF-245, Revision 1, TSTF-340, Revision 3, and new TS 3.7.5 Condition B, 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 the requirements in Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System,” to clarify the OPERABILITY of an AFW train during alternate alignments, to provide added flexibility in MODE 3 to repair and test the turbine driven AFW pump following a refueling outage, and to clarify the OPERABILITY of the turbine driven AFW train with one steam supply inoperable. The AFW System is not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The AFW System is used to respond to accidents previously evaluated. The proposed change affects only the actions taken when portions of the AFW System are unavailable and does not affect the design of the AFW System. The change to TS 3.7.5 adding actions for inoperable automatic control of level control valves does not change any of the assumptions in accidents previously evaluated and would not have an impact on accident consequences. No physical changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated.

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

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

Response: No.

The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes. The changes to the Conditions and Required Actions do not change any existing accident scenarios, nor create any new or different accident scenarios.

The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Manual control of AFW level control valves is not an accident initiator.

Therefore, it is concluded that 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 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 impacted by this change. The proposed change will not result in plant operation in a configuration outside the design basis.

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

For the proposed changes related to TSTF-412, Revision 3, in its application dated June 1, 2011, the licensee has affirmed the applicability of the model no significant hazards consideration published in the Federal Register as part of the CLIIP (72 FR 39093; July 17, 2007). As required by 10 CFR 50.91(a), an analysis of the issue of no significant Start Printed Page 77570hazards consideration, from the model application, is presented below:

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

Response: No.

The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an initiator of any design basis accident or event, and therefore the proposed changes do not increase the probability of any accident previously evaluated. The proposed changes to address the condition of one or two motor driven AFW/EFW trains inoperable and the turbine driven AFW/EFW train inoperable due to one steam supply inoperable do not change the response of the plant to any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, the changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

Response: No.

The proposed changes do not result in a change in the manner in which the AFW/EFW System provides plant protection. The AFW/EFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes. The changes to the Conditions and Required Actions do not change any existing accident scenarios, nor create any new or different accident scenarios.

The 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. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements.

The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

Therefore, the changes do 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 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 impacted by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis.

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

For the proposed changes related to TSTF-439, Revision 2, 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 eliminate certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Time are no different than the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. 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 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 accident from any accident previously evaluated?

Response: No.

The 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 changes do not alter any assumptions made in the safety analysis.

Therefore, the proposed change does 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 change to delete the second Completion Time 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 changes will not result in plant operation in a configuration outside of the design basis.

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

Based on the above, the NRC staff has reviewed the licensee's analyses 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: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: October 12, 2011.

Description of amendment request: The amendment requests authorization to update the facility's Final Safety Analysis Report to exempt five Unit 1 high-head safety injection system (HHSI) containment isolation valves (CIVs) from the VCSNS, Unit No. 1 Local Leak Rate Testing (LLRT) Program requirements.

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

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

Response: No.

The amendment request is to remove five Containment Isolation Valves (XVG08801A, XVG08801B, XVG08884, XVG08885, and XVG08886) from the Local Leak Rate Test (LLRT) program. These valves were originally included in the LLRT under 10 CFR [part] 50, Appendix J, in what is now Option A. VCSNS has been approved for 10 CFR [Part] 50, Appendix J, Option B under License Amendment No. 135. Under Option B, valves Start Printed Page 77571may be excluded from LLRT Type C testing if they are not a potential containment atmosphere leakage path. Based on the design and operation of the Safety Injection System, the valves do not constitute a containment atmospheric leakage path as covered in the Safety Evaluation. Since the valves are not a leakage path, there is no impact on the consequence of an accident. Moreover, the valves are not a part of the Reactor Coolant Pressure Boundary and are normally closed during plant operation, thus they do not affect the probability of an accident in any way. [The change does not affect plant equipment or operating practices and therefore does not significantly increase the probability or consequences of an accident previously evaluated.]

2. Does the proposed change create the possibility of a new or different kind of accident of malfunction that has not previously been evaluated?

Response: No.

The system design and operation are not changing. This test [* * *] [change] does not change the way the valves are used as a part of the Safety Injection System. A detailed Failure Modes and Effects Analysis were completed to confirm the system operation would meet the containment isolation design function. [The change does not add new or change existing plant equipment or affect the operating practices of the facility. Therefore, the 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 test [* * *] [change] is within existing regulatory requirements. The application of a closed loop outside of containment is appropriate and consistent with regulatory positions. The closed loop is applied to cold leg recirculation alignment of less than 8 hours when a run failure of a charging pump or RHR [residual heat removal] pump occurs. The probability of an HHSI\Charging Pump failure to run is 7.025E-06 per hour and for a LHSI [low-head safety injection]\RHR Pump is 7.689E-06 per hour. With containment integrity maintained within the allowable regulatory framework, there is no reduction in the margin of safety. [The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in margin of safety.]

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

Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Branch Chief: Gloria Kulesa.

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

Date of amendment request: October 12, 2011.

Description of amendment request: The amendment request proposes changes to allow for a one time extension to the 10-year frequency of the VCSNS containment leakage rate test (e.g., integrated leak rate test (ILRT) or “Type A test”) required by Technical Specification (TS) 6.8.4(g). The proposed change would permit the existing ILRT frequency to be extended from 10 years to approximately 10.9 years.

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 changes in brackets.

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

Response: No.

The proposed [* * *] [change] 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 130 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.

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 the [American Society of Mechanical Engineers (ASME), Section Xl, Boiler and Pressure Vessel Code,] 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 change 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 [Nuclear Energy Institute] 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 Xl 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 Start Printed Page 77572acceptance 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: J. Hagood Hamilton, Jr., South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Branch Chief: Gloria Kulesa.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of amendment request: September 2, 2011.

Description of amendment request: The amendments would revise a number of Technical Specification (TS) requirements, to impose similar restrictions on the movement of non-irradiated fuel assemblies to those currently in place for movement of irradiated fuel assemblies. The additional restrictions will limit the movement of all fuel assemblies over irradiated fuel assemblies in containment or in the fuel storage pool.

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 Technical Specifications applicability wording regarding the movement of fuel assemblies in containment and the fuel storage pool at the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 to include the movement of both irradiated and non-irradiated fuel assemblies. The proposed applicability is more comprehensive than the current applicability.

Expanding the applicability of the relevant Technical Specifications is necessary to account for updated fuel drop analyses which demonstrate that impacted spent fuel assemblies may be damaged. Consequently, movement of nonirradiated fuel assemblies could result in a Fuel Handling Accident that has radiological consequences. Changing the applicability of the relevant Technical Specifications does not affect the probability of a Fuel Handling Accident. The expanded applicability provides assurance that equipment designed to mitigate a Fuel Handling Accident is capable of performing its specified safety function.

The dose consequences due to failure of two assemblies remain within the Regulatory Guide 1.183 and 10 CFR 50.67 acceptance criteria limits. The Exclusion Area Boundary (EAB), Low Population Zone (LPZ) and Control Room dose results and associated limits are presented below:

FHA inside fuel handing buildingNew analysis FHA-FHB (rem TEDE)Regulatory guide 1.183 limit (rem TEDE)10 CFR 50.67 limit (rem TEDE)
EAB1.7≤6.325
LPZ<0.16.3≤25
Control Room0.6≤5≤5
FHA inside containmentNew analysis FHA-IC (rem TEDE)Regulatory guide 1.183 limit (rem TEDE)10 CFR 50.67 Limit (rem TEDE)
EAB1.7≤6.3≤25
LPZ<0.1≤6.3≤25
Control Room0.6≤5≤5

Consequently, 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 updated fuel assembly drop analysis demonstrates that impacted fuel assemblies may be damaged as the result of a dropped fuel assembly. The existing SONGS Technical Specifications regarding movement of fuel assemblies are not applicable for movement of non-irradiated fuel assemblies. A drop of a non-irradiated fuel assembly that has radiological consequences could occur during periods when equipment that would be required to mitigate those consequences is not required to be OPERABLE in accordance with the existing Technical Specifications.

The proposed change to the Technical Specifications applicability language regarding the movement of fuel assemblies in containment and the fuel storage pool at SONGS Units 2 and 3 ensure that Limiting Conditions for Operation and appropriate Required Actions for required equipment are in effect during fuel movement. This provides assurance that any Fuel Handling Accident that may occur will remain within the initial assumptions of accident analyses.

Consequently, there is no possibility of a new or different kind of accident due to the proposed change.

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

Response: No.

The proposed change will not affect protection criterion for plant equipment and will not reduce the margin of safety. By extending the Technical Specification applicability to the movement of non-irradiated fuel assemblies, the current margin of safety is maintained.

Consequently, there is no significant reduction in a margin of safety due to the proposed change.

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 requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston County, Alabama

Date of amendment request: September 9, 2011.Start Printed Page 77573

Description of amendment request: The proposed change would add Surveillance Requirement (SR) 3.3.1.14 to FNP TS Table 3.3.1-1, “Reactor Trip System [RTS] Instrumentation,” Function 3, “Power Range Neutron Flux High Positive Rate” to the Technical Specifications. SR 3.3.1.14 requires verification that the RTS Response Time is within limits every 18 months on a Staggered Test Basis. Function 3 is the Power Range Neutron Flux High Positive Rate Trip (PFRT) function.

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 Farley Nuclear Plant (FNP) Technical Specification (TS) 3.3.1, “Reactor Trip System (RTS) Instrumentation,” Table 3.3.1-1, “Reactor Trip System Instrumentation,” does not significantly increase the probability or consequences of an accident previously evaluated in the Update[d] Final Safety Analysis Report (UFSAR). The overall protection system performance will remain within the bounds of the accident analysis since there are no hardware changes. The design of the Reactor Trip System (RTS) instrumentation, specifically the power range neutron flux high positive rate trip (PFRT) function, will be unaffected. The reactor protection system will continue to function in a manner consistent with the plant design basis. All design, material, and construction standards, that were applicable prior to the request, are maintained.

The proposed change imposes additional surveillance requirements to assure safety related structures, systems, and components (SSCs) are verified to be consistent with the safety analysis and licensing basis. In this specific case, a response time verification requirement will be added to the PFRT function.

The proposed changes will not modify any system interface. The proposed changes will not affect the probability of any event initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance. The proposed change will not alter any assumptions nor change any mitigation actions in the radiological consequences evaluations in the UFSAR.

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter nor prevent the ability of SSCs from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change is consistent with the safety analyses assumptions and resultant consequences.

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

Response: No.

There are no hardware changes nor are there any changes in the method by which any safety related plant system performs its safety function. This change will not affect the normal method of plant operation nor change any operating parameters. No performance requirements will be affected; however, the proposed change does impose additional surveillance requirements. The additional surveillance requirements are consistent with assumptions made in the safety analyses and licensing basis.

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

Therefore, the proposed change does not create the possibility of a new or different 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 does not affect the acceptance criteria for any analyzed event nor is there a change to any Safety Limits. There will be no effect on the manner in which Safety Limits or Limiting Conditions of Operations are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.

The safety analyses limits assumed in the accident analysis are unchanged. The imposition of additional surveillance requirements increases the margin of safety by assuring that the affected safety analyses assumptions on equipment response time are verified on a periodic frequency.

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: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Branch Chief: Gloria J. Kulesa.

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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20874. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at 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 NRC's PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by email to pdr.resource@nrc.gov.

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: April 11, 2011.

Brief description of amendment: The amendment modified Technical Specification (TS) 3.4.7, “RCS [Reactor Coolant System] Leakage Detection Start Printed Page 77574Instrumentation,” to define a new time limit for restoring inoperable reactor coolant system (RCS) leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable; and make TS Bases changes which reflect the proposed changes and more accurately reflect the contents of the facility design basis related to operability of the RCS leakage detection instrumentation. These changes are consistent with NRC-approved Revision 3 to Technical Specification Task Force (TSTF) Change Traveler TSTF-514, “Revise BWR [Boiling-Water Reactor] Operability Requirements and Actions for RCS Leakage Instrumentation,” as part of the consolidated line item improvement process.

Date of issuance: November 21, 2011.

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

Amendment No.: 172.

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

Date of initial notice in Federal Register: June 28, 2011 (76 FR 37847).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 21, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 10, 2010, as supplemented by letter dated August 26, 2011.

Brief description of amendments: The change revised the PPL Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specifications (TSs) Surveillance Requirements (SRs) 3.4.3.1 “Safety/Relief Valves (S/RVs)” to the lower tolerances from −3% to −5%. These changes would be limited to the lower tolerances and does not affect the upper tolerances. These changes only apply to the lower as-found tolerances and not to the as-left tolerances, which will remain unchanged at ±1% of the safety lift setpoint. The as-found tolerances are used for determining past operability and to increase sample sizes for S/RV testing should the upper tolerances be exceeded. There will be no revision to the actual setpoints of the valves installed in the plant due to this change.

Date of issuance: November 17, 2011.

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

Amendment Nos.: 257 for Unit 1 and 237 for Unit 2.

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

Date of initial notice in Federal Register: February 22, 2011 (76 FR 9828).

The supplement dated August 26, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 17, 2011.

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: December 21, 2010.

Brief description of amendments: The amendments revised Technical Specification (TS) 5.3.1, “FUEL ASSEMBLIES,” by adding Optimized ZIRLOTM fuel rods to the fuel matrix in addition to Zircaloy or ZIRLOTM fuel rods that are currently in use. The amendments also added a reference to an NRC-approved Westinghouse Electric Company, LLC topical report regarding Optimized ZIRLOTM to Section 6.9.1.6, “Core Operating Limits Report (COLR).”

Date of issuance: November 17, 2011.

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

Amendment Nos.: Unit 1—198; Unit 2—186.

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: April 5, 2011 (76 FR 18804).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 17, 2011.

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: August 10, 2011.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.8.1 “AC [Alternating Current] Sources—Operating.” The change modified Surveillance Requirement (SR) Notes associated with SR 3.8.1, SR 3.8.1.9, SR 3.8.1.10, SR 3.8.1.11, SR 3.8.1.13, SR 3.8.1.16, SR 3.8.1.18, and SR 3.8.1.19. The amendment changed the WBN Unit 1 TS 3.8.1 to permit performance of the WBN Unit 2 integrated safeguards test without requiring WBN Unit 1 be shut down.

Date of issuance: November 22, 2011.

Effective date: As of the date of issuance and shall be implemented no later than 30 days from date of issuance.

Amendment No.: 89.

Facility Operating License No. NPF-90: Amendment revised the License and TSs.

Date of initial notice in Federal Register: September 20, 2011 (76 FR 58306).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 22, 2011.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 2nd day of December 2011.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

End Signature End Preamble

[FR Doc. 2011-31901 Filed 12-12-11; 8:45 am]

BILLING CODE 7590-01-P