2012-16656. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

Background

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (Commission or the 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 or combined license, as applicable, 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 June 14 to June 27, 2012. The last biweekly notice was published on June 26, 2012 (77 FR 38094-38099).

ADDRESSES:

You may access information and comment submissions related to this document, which the NRC possesses and are publically available, by searching on http://www.regulations.gov under Docket ID NRC-2012-0161. You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0161. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail].
• 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.

For additional direction on accessing information and submitting comments, see “Accessing Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2012-0161 when contacting the NRC about the availability of information regarding this document. You may access information related to this document by any of the following methods:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0161.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. Documents may be viewed in ADAMS by performing a search on the document date and docket number.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2012-0161 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS, and the NRC does not edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined 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; 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.

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 or combined 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 20852. The NRC regulations are accessible electronically from the NRC Library on the NRC's 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 [e-mail], 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's 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 should be in Portable Document Format (PDF) in accordance with the NRC guidance available on the NRC's 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's 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's Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email at [e-mail], 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 20852. 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 [e-mail].

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: February 22, 2012.

Description of amendment request: The proposed amendments would allow the use of the nuclear service water system (NSWS) pump discharge crossover valves and associated piping to cross tie McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2) NSWS trains to mitigate a Loss of Service Water (LOSW) event at McGuire 1 or 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:

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

Response: No.

MNS' [McGuire Nuclear Station's] Final Safety Analysis Report (FSAR) conforms to the standard format and content of Revision 1 to Regulatory Guide (RG) 1.70 with exceptions described in the applicable sections of the FSAR. With regard to Chapter 15 “Accident Analysis,” MNS committed to analyzing the anticipated operational occurrences and postulated design basis accidents listed in Chapter 15 on pages 15T-1, 15T-2, and 15T-3 of RG 1.70 Revision 1. MNS' FSAR Chapter 15 described an exception to a Loss of Service Water event (RG 1.70, Rev. 1, page 15T-3, item 30) and stated, in part, “Loss of the Nuclear Service Water System is not considered a credible accident because of the redundancy provided in the system.” The FSAR was later updated (UFSAR) to conform to Chapter 15 accidents listed on pages 15-10, 15-11, and 15-12 of RG 1.70 Revision 3. The initial FSAR Chapter 15 exception to RG 1.70 Rev. 1 LOSW event was no longer required since LOSW events were no longer included in Chapter 15 of subsequent RG 1.70 revisions (revision 2 or 3). Based on the licensing history, the LOSW event is not an anticipated operational occurrence or postulated design basis accident and was not previously analyzed in Chapter 15 of the UFSAR. A failure of the NSWS does not initiate any of the accidents previously evaluated in Chapter 15 of the UFSAR; therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2:

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

Response: No.

McGuire [Nuclear Station] is a multi-unit site comprised of two nuclear stations, Unit 1 and Unit 2. Each unit has two NSWS trains and each train is designed to remove core decay heat following a design basis LOCA. Each train has a service water pump discharge crossover valve installed which allows the trains to be cross-connected in any combination. The NSWS pump discharge crossover valves are described in the UFSAR as providing operational flexibility. Although designed to cross-connect unit NSWS trains, MNS has never licensed their use. The proposed change, consistent with the UFSAR description and [Generic Letter] GL 91-13, will provide the operational flexibility to allow one unit's NSWS to be aligned to another unit that has lost all service water.

During normal operation, only one pump, per unit, is in operation to supply NSWS flow to the essential and non-essential headers for each unit. Cross-connecting NSWS between units will require a unit's standby NSWS pump to be placed in service (operating), opening its respective discharge crossover valve, and opening a LOSW unit's NSWS pump discharge crossover valve to establish service water flow to a LOSW unit's NSWS train. With exception to the flow path, the shared train is operated as designed. If the proposed [license amendment request] LAR is approved, the necessary site procedures will be revised to govern system operation and use of the crossover design feature to mitigate a LOSW event.

The use of the NSWS pump discharge crossover valves within their design limitations and maintaining compliance to [technical specification] TS 3.7.7 [limiting condition for operation] LCO does not create any credible new failure mechanisms, malfunctions, or accident initiators that will prevent the ability of the NSWS to perform its design function. Operating the NSWS within the allowances of TS 3.7.7, which allow a train to be removed from service for up to 72 hours, does not impact the redundant capabilities afforded by the other train or the “low probability of a design basis accident (DBA) occurring during this time period” as stated in TS 3.7.7 Bases. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3:

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

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these barriers will not be impacted by the proposed change. The use of a NSWS pump discharge cross-over to cross-tie units is not a credited flow path in design basis and is not needed to perform the specified safety function. Cross-connecting the units is an additional strategy made available if a total LOSW should occur.

The proposed change will allow a unit to share a portion of an available service water train's capacity with a unit that has lost all service water. The shared alignment requires the use of service water pump discharge crossover valves which are not designated as shared components. Their use will improve the availability of service water and decreases the probability of core damage. Therefore the change will improve the margin of safety for each unit with respect to mitigating LOSW events.

Placing a NSWS train in a shared alignment prevents the train from automatically performing its safety function and the train does not comply with GDC-5 [10 CFR Part 50, Appendix A, “General Design Criteria for Nuclear Power Plants,” Criterion 5, “Sharing of structures, systems, and components”] and is declared inoperable. Limiting the time a train is inoperable to 72 hours manages the vulnerability to single failure consistent with current TS required actions and completion times. In accordance with TS LCO 3.0.2 allowances, TS 3.7.7 allows one train to be removed from service for up to 72 hours to perform surveillance testing, preventive maintenance, corrective maintenance, modifications, or investigation of operational problems. Although a NSWS train is declared inoperable for these activities, several can be accomplished while maintaining the train available while others, such as corrective maintenance, may also render the NSWS train unavailable. The 72 hour [completion time] CT is bounded by the worst case allowed by TS LCO 3.0.2 which assumes a train is both inoperable and unavailable.

Sharing a unit's redundant [nuclear service water] NSW pump requires the shared unit's service water pump to be taken out of standby and placed in service (operating). Therefore, the shared train remains available to the shared unit in event it must be restored. The shared train will be supplying the service water necessary to support operation of the shared unit's diesel generator (emergency power) and to assure long term operation of the shared pump. Although redundancy is lost in terms of performing its specified safety function on the designated unit, availability and functionality is maintained by the proposed amendment.

The reason a redundant NSWS pump is inoperable and/or unavailable does not change the probability its redundant train will fail during the 72 hour CT or change the probability of a [loss-of-coolant-accident] LOCA occurring during that time. In the event a train fails while its redundant train is shared, immediate action can be taken to restore the shared train from the shared alignment or the unit can be shutdown.

Since a unit's redundant service water train is placed in a shared configuration to mitigate a LOSW event, margin of safety is considered on each unit. Technical Specifications allows a nuclear service water train to be removed from service for up to 72 hours. The shared unit's margin of safety is maintained by limiting the shared alignment to <72 hour completion time consistent with current TS allowances. Implementation of this amendment will improve the margin of safety on a unit experiencing a LOSW event consistent with the intent of NRC Generic Letter 91-13.

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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Nancy L. Salgado.

Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, Unit 1 (ONS 1), Oconee County, South Carolina

Date of amendment request: April 3, 2012.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to authorize a one-time, 15 month extension to the integrated leak rate test (ILRT) of the reactor containment building (also known as the containment), which would align the test schedule with the refueling outage schedule. The ILRT is normally performed every 10 years. The upcoming ILRT is currently due by December 8, 2013.

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

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

Response: No.

The proposed exemption involves a one-time extension to the current interval for ONS 1 Type A containment testing. The current test interval of 120 months (10 years) would be extended on a one-time basis to no longer than approximately 135 months from the last Type A test. The proposed extension does not involve either 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 containment and the testing requirements invoked to periodically demonstrate the integrity of the 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.

This proposed extension is for next ONS 1 Type A containment leak rate test only. The Type B and C containment leak rate tests would continue to be performed at the frequency currently required by the ONS 1 TS. As documented in NUREG 1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The ONS 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that 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 combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would 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.

Therefore, it is concluded that the proposed amendment does not significantly increase 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 amendment to the TS involves a one-time extension to the current interval for the ONS 1 Type A containment test. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

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 amendment to the TS involves a one-time extension to the current interval for the ONS 1 Type A containment test. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of 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 for ONS 1. The proposed surveillance interval extension is bounded by the 15-month extension currently authorized within NEI 94-01, Revision 0. Type B and C containment leak rate tests would continue to be performed at the frequency currently required by TS. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would 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 change does not involve a significant reduction in a margin of safety.

Based on the NRC staff's 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Nancy L. Salad.

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

Date of amendment request: December 16, 2011, as supplemented by letters dated January 20, March 1, March 16, and April 18, 2012.

Description of amendment request: The proposed amendments would revise the Technical Specifications and the Updated Final Safety Analysis Report to add the new Protected Service Water (PSW) System to the plant's licensing basis as an additional method of achieving and maintaining safe shutdown of the reactors in the event of a high-energy line break or a fire in the turbine building, which is shared by all three units.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee provided its analysis of the issue of no significant hazards consideration. The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's analysis of the no significant hazards consideration is presented below:

Criterion 1:

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

Response: No.

The changes proposed include the construction of a new PSW building, which will have the equipment to receive electrical power from two independent sources and provide electrical power to important equipment located in the auxiliary building or the reactor containment building without being routed through the turbine building. Since certain high-energy line breaks (HELBs) or fires in the turbine building could adversely affect the power supplies to equipment needed to maintain the reactors in safe shutdown, the PSW System provides added assurances that safe shutdown can be achieved and maintained. The PSW system does not have any failure modes that would initiate the type of accidents previously evaluated, so there will be no increase in the probability of an accident previously evaluated. The PSW System modifications will be designed and installed in accordance with applicable quality standards such that there will be no significant increase in the probability of failure or malfunction of existing structures, systems, or components (SSCs) used to mitigate accidents. Since there will be no significant increase in the probability of malfunction of these SSCs, there also will be no significant increase in the consequences of accidents previously evaluated.

Criterion 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 modifications are designed to enhance the station's ability to achieve safe shutdown following a HELB or fire in the turbine building. As the new equipment will be designed and installed in accordance with applicable quality standards, there is reasonable assurance that it will not introduce new malfunctions or accident initiators different from the accidents that are already evaluated.

Criterion 3:

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

Response: No.

The addition of the PSW system improves the station's overall risk margin, therefore this change does not involve a significant reduction in a margin of safety.

Based on the NRC staff's 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: April 4, 2012.

Description of amendment request: The proposed amendment addresses the Arkansas Nuclear One, Unit No. 2 (ANO-2) revised fuel handling accident (FHA) based on the U.S. Nuclear Regulatory Commission (NRC) staff approved license amendment request regarding use of Alternate Source Terms (AST) (NRC safety evaluation dated April 26, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110980197)). As presented in the licensee's letter dated March 31, 2010 (ADAMS Accession No. ML100910241), the original FHA analysis assumed failure of 60 fuel rods in a single fuel assembly. The revised analysis assumes the failure of all fuel rods in two fuel assemblies (472 rods). The revised analysis was provided in the licensee's letter dated June 23, 2010 (ADAMS Accession No. ML102000199).

The changes necessary to support the revised FHA affect similar Technical Specifications (TSs) associated with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Travelers TSTF-51, Revision 2, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations”; TSTF-272, Revision 1, “Refueling Boron Concentration Clarification”; TSTF-268, Revision 2, “Operations Involving Positive Reactivity Additions”; and TSTF-471, Revision 1, “Eliminate use of Term Core Alterations in Actions and Notes.” Therefore, the licensee proposes to adopt these TSTFs in conjunction with changes necessary to support the revised FHA. Additionally, administrative and/or editorial errors noted during the review are also corrected (in relation to the TS pages affected by the aforementioned proposed changes).

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. Each of the five items described above is addressed individually under each of the three standards, as presented below:

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

Revised FHA

Response: No.

TS changes associated with the FHA analysis ensure the initial assumptions of the FHA are maintained and, therefore, act to minimize the consequences of an accident by ensuring TS required features are operable during the movement of fuel assemblies. The FHA analysis was recently accepted by the NRC during adoption of Alternate Source Terms for ANO-2. The probability of a fuel assembly drop (or any load drop) is unchanged by the revised analysis. Therefore, the revised FHA does not involve a significant increase in the probability of an accident previously evaluated.

The FHA analysis was recently accepted by the NRC during adoption of Alternate Source Terms for ANO-2. In addition, Licensee's has reviewed station procedures and controls in order to verify that no other loads, other than a new or irradiated fuel assembly, need be addressed with regard to a FHA (i.e., no other known load carried over irradiated fuel assemblies exists which would be expected to cause fuel damage if dropped). The proposed TS changes simply ensure required systems will be operable during operations that could lead to an FHA. Based on the above, the proposed FHA-related changes to the TSs do not result in a significant increase in the consequences of an accident previously evaluated.

TSTF-51 and TSTF 471

Response: No.

The only design basis accident assumed for ANO-2 related to the proposed changes is the FHA. The boron dilution event is evaluated, but considered an unlikely event due to the time available for operator response and the administrative controls that permit early detection of the event. The loss of SDC [shutdown cooling] event has little relationship and minimal impact with regard to a FHA. TSTF-51 and TSTF-471 simply replace the use of the previously defined “core alterations” term with requirements associated with the movement of fuel assemblies, since the drop of a fuel assembly is the only event that could reasonably lead to an FHA or a significant challenge to the plant.

The removal of all references to “core alterations” in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core) since such manipulation cannot result in an FHA, boron dilution event, or loss of SDC. In addition, manipulation of these other components cannot present a significant challenge to SDM [shutdown margin] because the TS required RCS [reactor coolant system] boron concentration for Mode 6 operation provides substantial margin to criticality.

Changes associated with TSTF-51 and TSTF-471 do not modify limitations in such a way that the consequences of an FHA would be greater than that assumed in the FHA analysis (i.e., 10 CFR 50.67 and General Design Criterion (GDC) 19 limitations are not exceeded following a FHA)).

Based on the above, the proposed changes associated with the adoption of TSTF-51 and TSTF-471 do not result in a significant increase in the probability or consequences of an accident previously evaluated.

TSTF-272

Response: No.

Changes associated with TSTF-272 simply place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the RCS when the RCS is in direct hydraulic communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are unrelated to any accident initiator and further prohibit any challenge to the fuel in the reactor vessel by ensure sufficient boron concentration is maintained during Mode 6 operations. Therefore, these changes do not result in a significant increase in the probability or consequences of an accident previously evaluated.

TSTF-286

Response: No.

Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. The activities that involve inventory makeup from sources with boron concentrations less than the current RCS concentration (i.e., boron dilution) need not be precluded in the TSs provided the required SDM is maintained for the worst-case overall effect on the core. Note that an unexpected boron dilution event is considered unlikely for ANO-2 due to the significant period of time for operator detection and response before SDM would be significantly challenged (reference ANO-2 SAR Section 15.1.4.3). In addition, while a boron dilution event is evaluated in the safety analysis, the only “accident” assumed for ANO-2 during Mode 6 operations is the FHA. Permitting RCS inventory and temperature adjustments is unrelated to any assumptions associated with a FHA. Therefore, these changes do not result in a significant increase in the probability an accident (or a boron dilution event) previously evaluated. Because an unexpected boron dilution event provides sufficient opportunity for detection and recovery, the proposed changes associated with TSTF-286 likewise do not result in a significant increase in the consequences of an accident (or boron dilution event) previously evaluated.

Enhancements and Administrative Changes

Response: No.

Enhancements and administrative changes proposed for specifications affected by the above revised FHA or TSTF adoptions are unrelated to any accident initiator. Administrative changes likewise cannot impact the consequences of any accident previously evaluated.

Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage. The proposed changes are neutral or more restrictive and, therefore, cannot increase the consequences of an accident previously evaluated.

Based on the above, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

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

Revised FHA

Response: No.

TS changes associated with the revised FHA involve no physical changes to the plant. These changes act to ensure required SSCs are operable when moving irradiated fuel assemblies or new fuel assemblies over irradiated fuel assemblies to limit any Control Room or offsite dose consequences to within acceptable limits. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-51 and TSTF-471

Response: No.

TS changes associated with the adoption of these TSTFs involve no physical changes to the plant. The removal of all references to “core alterations” in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core). Such manipulations cannot result in an FHA, boron dilution event, or loss of SDC. In addition, such manipulations cannot result in an appreciable change in core reactivity due to the high RCS boron concentration required during refueling operations by the TSs. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism.

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

TSTF-272

Response: No.

Changes associated with TSTF-272 place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the RCS when the RCS is in direct hydraulic communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are unrelated to any accident initiator and further prohibit any challenge to the fuel in the reactor vessel by ensure sufficient boron concentration is maintained during Mode 6 operations. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

TSTF-286

Response: No.

Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. No physical plant changes are related to these TS changes. The only accident or event that could be affected by this change is the boron dilution event, which has been previously evaluated. The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Enhancements and Administrative Changes

Response: No.

Enhancements and administrative changes proposed for specifications affected by the above revised FHA or TSTF adoptions are unrelated to any accident initiator and involve no physical changes to the plant.

Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage.

The proposed changes do not introduce a new accident initiator, accident precursor, or accident-related malfunction mechanism. Based on the above, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Revised FHA

Response: No.

TS changes associated with the revised FHA act to ensure required SSCs [structures, systems, and components] are operable when moving irradiated fuel assemblies or new fuel assemblies over irradiated fuel assemblies to limit any Control Room or offsite dose consequences to within acceptable limits. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

TSTF-51 and TSTF-471

Response: No.

The removal of all references to “core alterations” in favor of restrictions associated with the movement of fuel assemblies eliminates current restrictions associated with the manipulation of other core components (i.e., sources or reactivity control components within the core). Such manipulations cannot result in an FHA, boron dilution event, or loss of SDC. In addition, such manipulations cannot result in an appreciable change in core reactivity due to the high RCS boron concentration required during refueling operations by the TSs. Changes associated with TSTF-51 and TSTF-471 do not modify limitations in such a way that the consequences of an FHA would be greater than that assumed in the FHA analysis (i.e., 10 CFR 50.67 and GDC 19 limitations are not exceeded following a FHA). Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

TSTF-272

Response: No.

Changes associated with TSTF-272 place additional restrictions on Mode 6 operations by ensuring the boron concentration of the water in the refueling canal meets the same TS limits required for the RCS when the RCS is in direct hydraulic communication with the refueling canal (i.e., reactor vessel head removed and refueling canal filled). These changes are more restrictive than the current specification and therefore do not involve a significant reduction in a margin of safety.

TSTF-286

Response: No.

Changes associated with TSTF-286 permit operator control of RCS inventory and temperature when certain TS requirements are not met, provide the overall required SDM of the RCS is maintained. The only accident or event that could be affected by this change is the boron dilution event, which has been previously evaluated. While the margin between existing boron concentration and that required to meet SDM requirements may be reduced, margin is gained by permitting operators to take corrective action to maintain RCS inventory and temperature within limits during periods when such operations are otherwise prohibited. While not quantifiable, the changes associated with TSTF-286 have a general balanced effect in relation to the margin of safety. Because an unexpected boron dilution event provides sufficient opportunity for detection and recovery, the proposed changes associated with TSTF-286 do not involve a significant reduction in a margin of safety.

Enhancements and Administrative Changes

Response: No.

Enhancements and administrative changes proposed for specifications affected by the above revised FHA or TSTF adoptions are unrelated to any accident initiator or mitigation strategy. Enhancements associated with the Containment Purge system radiation instrumentation ensure Surveillance testing is performed when the system is in service, regardless if an actual Purge is taking place. In addition, the proposed changes ensure appropriate testing is performed prior to placing the system in service each refueling outage. Based on the above, these proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Joseph A. Aluise, Associate General Council—Nuclear, Entergy Services, Inc., 639 Loyola Avenue New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa

Date of amendment request: May 1, 2012.

Description of amendment request: The proposed amendment would revise the Duane Arnold Energy Center (DAEC) Technical Specifications (TS) on a one-time basis by adding a note to TS Table 3.3.5.1-1, Function 1d, Modes 4 and 5, specifying that Function 1d is not required to be met during Refueling Outage (RFO) 23 in Modes 4 and 5.

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 any accident previously evaluated?

Response: No.

The proposed amendment would revise the DAEC TS on a one-time basis by adding a note to TS Table 3.3.5.1-1, Function 1d, Modes 4 and 5, specifying that Function 1d is not required to be met during RFO 23 in Modes 4 and 5. Accidents are initiated by the malfunction of plant equipment, or the catastrophic failure of plant structures, systems, or components.

The low pressure Emergency Core Cooling System (ECCS) subsystems are designed to inject to reflood or to spray the core after any size break up to and including a design basis Loss of Coolant Accident (LOCA). The proposed change to the Core Spray System Operability requirements does not change the operating configurations or minimum amount of operating equipment assumed in the safety analysis for accident mitigation. The change does not require any change in safety analysis methods or results. Also, it does not change the amount of core spray provided to the core in the accident analyses. No changes are proposed to the manner in which the ECCS provides plant protection or which would create new modes of plant operation. The proposed change does not result in any new or affect the probability of any accident 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. This change will only apply when the plant is in MODES 4 and 5 where LOCAs are not postulated to occur. In MODES 4 and 5, the CS function is to mitigate OPDRVs [Operations with the Potential for Draining the Reactor Vessel].

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 previously evaluated?

Response: No.

This change does not affect the method by which any plant systems perform a safety function. It does not introduce any new equipment, or hardware changes, which could create a new or different kind of accident. No new release pathways or equipment failure modes are created. No new accident scenarios failure mechanisms or limiting single failures are introduced as a result of this request. This request does not affect the normal methods of plant operation. The Core Spray System retains its ability to function following any accident previously evaluated and provide the proper flow rate to the core. This change will only apply when the plant is in MODES 4 and 5 where LOCAs are not postulated to occur. In MODES 4 and 5, the CS function is to mitigate OPDRVs. Strict administrative and procedural controls, operator training, and use of human performance tools will be essential to preventing these types of consequential human errors. Furthermore, both CS subsystems will be guarded and no work or testing will be permitted on either of the CS subsystems during RFO 23 when both CS subsystems are needed to be Operable to meet the requirements of LCO 3.5.2.

Therefore, the implementation of the proposed change will not create a possibility for an accident of a new or different type than those previously evaluated.

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

Response: No.

The ECCS are designed with sufficient redundancy such that if a Core Spray subsystem were unavailable, or did not provide the required flowrate, the remaining Core Spray subsystem is capable of providing water and removing heat loads to satisfy the Updated Final Safety Analysis Report requirements for accident mitigation. A minimum of two low pressure ECCS subsystems continue to be required to be OPERABLE in MODES 4 and 5, except with the spent fuel storage pool gates removed and water level ≥ 21 ft 1 inch over the top of the reactor pressure vessel flange. There is no change in the Limiting Conditions for Operation. For these reasons, the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. Mitchell S. Ross, P.O. Box 14000 Juno Beach, FL 33408-0420.

NRC Acting Branch Chief: Istvan Frankl.

Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: March 22, 2012.

Brief description of amendment request: The proposed amendments would revise the technical specification for the Vogtle Electric Generating Plant, Units 1 and 2, associated with the “Steam Generator (SG) Program” allowing the exclusion of portions of the SG tubes below the top of the tube sheet from periodic SG tube inspections during the remaining licensed operations of the plant. Furthermore, the amendment requests to remove the interim SG alternative inspection criteria that had been previously approved.

Date of publication of individual notice in Federal Register: May 25, 2012 (77 FR 31402).

Expiration date of individual notice: July 24, 2012.

Notice of Issuance of Amendments to Facility Operating Licenses and Combined 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.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, 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 Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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 PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to [e-mail].

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

Date of application for amendment: November 22, 2011, as supplemented by letter dated May 11, 2012.

Brief description of amendment: The amendments remove duplicate Technical Specification (TS) requirements and unit-specific references that are no longer needed. In addition, the administrative changes correct typographical errors and provide clarification to ensure understanding of the required actions of some of the TSs. The changes include corrective actions from the Unit 2 event described in Licensee Event Report (LER) 50-529/2011-001. The changes are administrative or editorial in nature, and would not result in any change to operating requirements. These administrative changes are for TS 3.3.1, “Reactor Protective System (RPS) Instrumentation—Operating”; TS 3.3.2, “Reactor Protective System (RPS) Instrumentation—Shutdown”; TS 3.3.5, “Engineered Safety Features Actuation System (ESFAS) Instrumentation”; TS 3.5.5, “Refueling Water Tank (RWT)”; TS 3.3.9, “Control Room Essential Filtration Actuation Signal (CREFAS)”; TS 3.7.11, “Control Room Essential Filtration System (CREFS)”; TS 5.4, “Procedures”; and TS 5.5.16, “Containment Leakage Rate Testing Program.”

Date of issuance: June 18, 2012.

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

Amendment No.: Unit 1—189; Unit 2—189; Unit 3—189.

Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: January 24, 2012 (77 FR 3510). The supplemental letter dated May 11, 2012, 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 on January 24, 2012 (77 FR 3510).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 18, 2012.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2), York County, South Carolina; Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2), Mecklenburg County, North Carolina; Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3 (Oconee 1, 2, and 3), Oconee County, South Carolina

Date of application for amendments: December 15, 2009, as supplemented by letter dated September 22, 2011.

Brief description of amendments: The amendments consist of changes to the Technical Specifications (TSs) associated with Reactor Coolant System (RCS) Specific Activity and the deletion of the TS definition of E Bar (average disintegration energy) consistent with Revision 0 to TS Task Force (TSTF) Standard Technical Specification Change Document TSTF-490, “Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.”

Date of issuance: June 25, 2012.

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

Amendment Nos.: Catawba: Unit 1—268 and Unit 2—264; McGuire: Unit 1—266 and Unit 2—246; Oconee: Unit 1—380, Unit 2—382, and Unit 3—381.

Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-17, DPR-38, DPR-47, and DPR-55: Amendments revised the licenses.

Date of initial notice in Federal Register: March 23, 2010 (75 FR 13789). The September 22, 2011, supplement 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 June 25, 2012.

No significant hazards consideration comments received: No.

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 application for amendments: October 21 and December 14, 2010, as supplemented by letters dated December 21, 2010, January 7, 2011, January 28, February 22, March 3, March 9 (two letters), March 16 (two letters), March 23, March 25, March 31 (two letters), April 14 (two letters), April 22 (2 letters), April 26, April 28 (2 letters), April 29, May 11, May 18, May 19 (two letters), May 26 (two letters), June 7, June 9, June 21 (two letters), July 7 (two letters), July 22, July 29, August 5, August 11, August 16 (two letters), August 19, August 25 (two letters), August 29, September 14, September 16, September 30 (two letters), October 6, October 12 (two letters), October 14, October 15, November 9, December 22 (2 letters), December 31, 2011, January 10, 2012, January 16 (two letters), January 17, January 19, January 23 (two letters), January 25, January 31, February 3, February 15, February 23 (two letters), and March 15, 2012.

Brief description of amendments: The proposed amendments would increase the licensed core power level for Turkey Point, Units 3 and 4 from 2300 megawatts thermal (MWt) to 2644 MWt. This represents a net increase in the core thermal power of approximately 15 percent, including a 13-percent power uprate and a 1.7 percent measurement uncertainty recapture, over the current licensed thermal power level and is defined as an extended power uprate. The proposed amendments would change the renewed facility operating licenses, the technical specifications (TSs) and licensing bases to support operation at the increased core thermal power level, including changes to the maximum licensed reactor core thermal power, reactor core safety limits, reactor protection system and engineered safety feature actuation system limiting safety system settings, and emergency diesel generator surveillance start voltage and frequency. Additional TS changes include reactor coolant system heatup and cooldown limitations, pressurizer safety valve settings, accumulator and refueling water storage tank boron concentrations, main steam safety valve maximum allowable power level and lift settings, new main feedwater isolation valves, and core operating limits report references. A complete list of the proposed TS changes and the licensee's basis for change can be found in Attachment 1 of the licensee's application (Agencywide Documents and Management System Accession No. ML103560167).

Date of issuance: June 15, 2012.

Effective date: Unit 3—This license amendment is effective as of its date of issuance and shall be implemented prior to Unit 3 startup from the spring 2012 refueling outage. Unit 4—This license amendment is effective as of its date of issuance and shall be implemented prior to Unit 4 startup from the fall 2012 refueling outage.

Amendment Nos.: Unit 3—249 and Unit 4—245.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: May 9, 2011 (76 FR 26771). The supplemental letters 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 June 15, 2012.

No significant hazards consideration comments received: No.

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 application for amendments: August 17, 2011, as supplemented by letters dated October 14, and December 1, 2011.

Brief description of amendments: The amendments revised items in Technical Specification (TS) 3.3.3.3, Table 3.3-5, Accident Monitoring Instrumentation, High Range-Noble Gas Effluent Monitors, Main Steam Lines, Instrument 19d, and TS 4.3.3.3, Table 4.3-4 related to the need to have High Range-Noble Gas Effluent Monitors for the Main Steam Lines. The changes relocated the TSs and surveillance requirements for this instrument to the Updated Final Safety Analysis Report and related procedures.

Date of issuance: June 15, 2012.

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

Amendment Nos.: Unit 3—250 and Unit 4—246.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the TSs and Surveillance Requirements.

Date of initial notice in Federal Register. October 18, 2011 (76 FR 64393). The supplements dated October 14 and December 1, 2011, 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 June 15, 2012.

No significant hazards consideration comments received: No.

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 application for amendments: May 25, 2011.

Brief description of amendments: The amendments relocate Technical Specifications (TSs) in Section 5.2—“Containment,” Section 5.4—“Reactor Coolant System,” and Section 5.6—“Component Cyclic or Transient Limit,” to the Updated Final Safety Analysis Report. TS 5.3.3 regarding spent fuel storage pool capacity would be revised to a total pool capacity limit only.

Date of issuance: June 21, 2012.

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

Amendment Nos.: Unit 3-251 and Unit 4-247.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the TSs.

Date of initial notice in Federal Register: October 18, 2011 (76 FR 64392).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2012.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 19, 2012.

Brief description of amendment: The NRC issued Amendment No. 239, Departure from a Method of Evaluation for the Auxiliary Building Overhead Crane (FHCR-5), on December 27, 2011. Amendment No. 239 was approved to be implemented within 180 days of issuance of the amendment. By letter dated March 19, 2012, the licensee requested extending the implementation period for Amendment 239 to allow for installation and testing of the new single failure proof FHCR-5. This amendment approved additional time to complete the implementation of Amendment No. 239 from 180 days to, “Implementation shall be completed 90 days prior to moving a spent fuel shipping cask with FHCR-5.”

Date of issuance: June 26, 2012.

Effective date: As of the date of issuance.

Amendment No.: 241.

Facility Operating License No. DPR-72: Amendment approved a revision to the Amendment No. 239 implementation schedule.

Date of initial notice in Federal Register: April 17, 2012 (77 FR 22814).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 26, 2012.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 29th day of June 2012.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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