Skip to Content

Notice

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

Document Details

Information about this document as published in the Federal Register.

Published Document

This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

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 December 15, 2011 to December 28, 2011. The last biweekly notice was published on December 27, 2011 (76 FR 80972).

Addresses: Please include Docket ID NRC-2011-0303 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-2011-0303. 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, or by fax 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 the 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. 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-2011-0303.

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 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. 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. The 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 should be in Portable Document Format (PDF) in accordance with the 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 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 pdr.resource@nrc.gov.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: August 22, 2011.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 6.9.1.6, “Core Operating Limits Report,” to add plant-specific methodology, ANP-3011 (P), “Harris Nuclear Plant Unit 1 Realistic Large Break LOCA [Loss-of-Coolant Accident] Analysis,” Revision 1, that implements AREVA's NRC-approved topical report, EMF-2103(P)(A), “Realistic Large Break LOCA Methodology for Pressurized Water Reactors,” and add EMF-2103(P)(A), “Realistic Large Break LOCA Methodology for Pressurized Water Reactors,” Revision 2 or higher upon approval of the specific revision by the NRC, to the TS 6.9.1.6.2 listing of analytical methods used to determine the core operating limits, and eliminates extraneous detail in TS 6.9.1.6 that cross references each method to the applicable TS Section 3.0 specifications and parameters.

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 TR [topical report] underlying the proposed HNP [Shearon Harris Nuclear Power Plant] methodology has been reviewed and approved by the NRC for use in determining core operating limits and for evaluation of LBLOCA [large break loss-of-coolant accident]. The core operating limits to be developed using the new methodologies for HNP will be established in accordance with the applicable limitations as documented in the NRC SE [safety evaluation]. In the April 9, 2003, NRC SE, the NRC concluded that the S-RELAP5 RLBLOCA [realistic large break loss-of-coolant accident] methodology is acceptable for referencing in licensing applications in accordance with the stated limitations.

The proposed change enables the use of new methodology to re-analyze a LBLOCA. It does not, by itself, impact the current design bases. Revised analysis may either result in continued conformance with design bases or may change the design bases. If design basis changes result from a revised analysis, the specific design changes will be evaluated in accordance with HNP design change procedures and 10 CFR 50.59.

The proposed change does not involve physical changes to any plant structure, system, or component (SSC). Therefore, the probability of occurrence for a previously analyzed accident is not significantly increased.

The consequences of a previously analyzed accident are dependent on the initial conditions assumed for the analysis, the behavior of the fission product barriers during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated.

The proposed methodologies will ensure that the plant continues to meet applicable design and safety analyses acceptance criteria. The proposed change does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. As a result, no analysis assumptions are impacted and there are no adverse effects on the factors that contribute to offsite or onsite dose as a result of an accident. The proposed change does not affect setpoints that initiate protective or mitigative actions. The proposed change ensures that plant SSCs are maintained consistent with the safety analysis and licensing bases.

Therefore, this amendment does not involve a significant increase in the probability or consequences of a previously analyzed accident.

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

Response: No.

The proposed change does not involve any physical alteration of plant SSCs. No new or different equipment is being installed and no installed equipment is being operated in a different manner. There is no change to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result, no new failure modes are being introduced.

Therefore, the proposed change will 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.

There is no impact on any margin of safety resulting from the incorporation of this new TR into the TS or deletion of cross-reference information from the description of the COLR [core operating limit report]. If design basis changes result from a revised analysis that uses these new methodologies, the specific design changes will be evaluated in accordance with HNP design change procedures and 10 CFR 50.59. Any potential reduction in the margin of safety would be evaluated for that specific design change.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: October 28, 2011.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) Table 3.2.B to increase the condensate storage tank low water level setpoint for the interlock to high-pressure coolant injection (HPCI) pump suction valves. The proposed amendment would also correct typographical errors in TS numbering and referencing that were introduced in previous license amendment requests.

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 increasing of the setpoint for the Condensate Storage Tank (CST) low water level High Pressure Coolant Injection (HPCI) System automatic suction transfer to the Suppression Pool is not a precursor to any accident previously evaluated. The CST is not utilized to mitigate the consequences of any accident previously evaluated. The increase in the setpoint provides for HPCI pump performance with the required flow to mitigate the accident conditions. The proposed corrections to typographical errors incurred in the prior License Amendments provide correct references to the applicable existing Specifications, which is an administrative change.

The proposed changes do not involve a change to the safety function of the HPCI system operation. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions.

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

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

Response: No.

The increasing of the setpoint for the Condensate Storage Tank (CST) low water level High Pressure Coolant Injection (HPCI) System automatic suction transfer to the Suppression Pool is not a precursor to any accident previously evaluated. The CST is not utilized to mitigate the consequences of any accident previously evaluated. The increase in the setpoint provides for HPCI pump performance with the required flow to mitigate the accident conditions. The proposed corrections to typographical errors incurred in the prior License Amendments provide correct references to the applicable existing Specifications, which is an administrative change.

The proposed changes do not change the safety function of the HPCI and RCIC [reactor core isolation cooling] systems. There is no alteration to the parameters within which the plant is normally operated. The increase in the setpoint is not a precursor to new or different kinds of accidents and do not initiate new or different kinds of accidents. The impact of these changes have been analyzed and found to be acceptable within the design limits and plant operating procedures. As a result, no new failure modes are being introduced.

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 margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event and design basis accidents. The proposed change increases the setpoint at which protective actions are initiated, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The corrections to the typographical errors introduced in prior License Amendments do not impact the safety margin.

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

NRC Branch Chief: Nancy Salgado.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Unit 2, LaSalle County, Illinois

Date of amendment request: October 26, 2011.

Description of amendment request: The proposed amendment revises license condition 2.C.(32) to require the installation of NETCO-SNAP-IN® inserts to be completed no later than December 31, 2012, for LaSalle County Station (LSCS) Unit 2. In addition, license condition 2.C.(31) is revised to apply until March 31, 2012, and a new license condition 2.C.(34) is being proposed to prohibit fuel storage after March 31, 2012, in spent fuel pool (SFP) storage rack cells that have not been upgraded with the NETCO-SNAP-IN® inserts.

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 the LSCS Unit 2 Operating License to accelerate the timeline for installation of the NETCO-SNAP-IN® inserts in the LSCS Unit 2 SFP, and limit the time period under which BORAFLEXTM is credited as the neutron absorbing material in the Unit 2 SFP. There are no changes to the SFP criticality analysis associated with the proposed change. The SFP criticality analysis was previously approved by the NRC and continues to demonstrate that the effective neutron multiplication factor, Keff, is less than or equal to 0.95 if the SFP is fully flooded with unborated water. No physical changes to the plant are proposed, no new plant equipment is being installed, and there are no changes to the manner in which the plant is operated. Rather, the proposed change is administrative because it involves accelerating the timeline for installing the NETCO-SNAP-IN® inserts and limiting the time period under which BORAFLEXTM is credited as the neutron absorbing material in the Unit 2 SFP.

The probability that a fuel assembly would be dropped is unchanged by the proposed change. These events involve failures of administrative controls, human performance, and equipment failures that are unaffected by the proposed change. The proposed change does not result in a significant increase in the consequence of an accident previously analyzed. The criticality analysis that demonstrates adequate margin to criticality for spent fuel storage rack cells with rack inserts in the LSCS Unit 2 SFP, and adequate criticality margin for assemblies accidentally dropped onto the spent fuel storage racks, is not being changed. The consequences of dropping a fuel assembly onto any other fuel assembly or other structure are unaffected by the proposed change.

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

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

Response: No.

The proposed change revises the LSCS Unit 2 Operating License to accelerate the timeline for installation of the NETCO-SNAP-IN® inserts in the LSCS Unit 2 SFP, and limit the time period under which BORAFLEXTM is credited as the neutron absorbing material in the Unit 2 SFP. There are no changes to the SFP criticality analysis associated with the proposed change. No physical changes to the plant are proposed, and there are no changes to the manner in which the plant is operated. Rather, the proposed change is administrative because it involves accelerating the timeline for installing the NETCO-SNAP-IN® inserts and limiting the time period under which BORAFLEXTM is credited as the neutron absorbing material in the Unit 2 SFP.

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

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

Response: No.

The proposed change revises the LSCS Unit 2 Operating License to accelerate the timeline for installation of the NETCO-SNAP-IN® inserts in the LSCS Unit 2 SFP, and limit the time period under which BORAFLEXTM is credited as the neutron absorbing material in the Unit 2 SFP. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in Technical Specifications. The proposed change does not alter these established safety margins. For SFP criticality, the required safety margin is 5% including a conservative margin to account for engineering and manufacturing uncertainties. The proposed change does not alter the criticality analysis for the SFP and does not affect the SFP criticality safety margin.

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 requested amendments involve no significant hazards consideration.

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

NRC Branch Chief: Jacob I. Zimmerman.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of amendment request: May 27, 2011.

Description of amendment request: The proposed amendment would modify Technical Specifications (TSs) to allow the BVPS-1 containment spray additive, sodium hydroxide (NaOH), to be replaced by sodium tetraborate (NaTB). Also, an administrative change to the BVPS-2 license is required.

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.

Use of NaTB in lieu of NaOH would not involve a significant increase in probability of a previously evaluated accident because the containment spray additive is not an initiator of any analyzed accident. The NaTB would be stored and delivered by a passive method that does not have potential to affect plant operations. Any existing NaOH delivery system equipment which remains in place but is removed from service would meet existing seismic and electrical requirements. Therefore the change in additive, including removal of NaOH equipment from service, would not result in any failure modes that could initiate an accident.

The spray additive is used to mitigate the consequences of a LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu of NaOH would not involve a significant increase in the consequences of a previously evaluated accident because the amount of NaTB specified in the proposed TS would achieve a pH of 7 or greater, consistent with the current licensing basis. This pH is sufficient to achieve long-term retention of iodine by the containment sump fluid for the purpose of reducing accident related radiation dose following a LOCA.

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

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

Response: No.

Regarding the proposed use of NaTB in lieu of NaOH, the NaTB would be stored and delivered by a passive method that does not have potential to affect plant operations. Any existing NaOH delivery system equipment that is removed from service would meet existing seismic and electrical requirements. Hydrogen generation would not be significantly impacted by the change.

Therefore, no new failure mechanisms, malfunctions, or accident initiators would be introduced by the proposed change, and it would 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.

Since the quantity of NaTB specified in the amended TS would reduce the potential for undesirable chemical effects while achieving radiation dose reductions, corrosion control and hydrogen generation effects that are comparable to NaOH, the proposed change does not involve a significant reduction in a margin of safety. The primary function of an additive is to reduce LOCA consequences by controlling the amount of iodine fission products released to containment atmosphere from reactor coolant accumulating in the sump during a LOCA. Because the amended [TS] would achieve a pH of 7 or greater using NaTB, dose related safety margins would not be significantly reduced. Use of NaTB reduces the potential for undesirable chemical effects that could interfere with recirculation flow through the sump strainers. Any existing NaOH delivery system equipment that remains in place but is removed from service would meet existing seismic and electrical requirements and would not interfere with operation of the existing containment or containment spray system.

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: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Nancy L. Salgado.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 1, Rockingham County, New Hampshire

Date of amendment request: November 17, 2011.

Description of amendment request: The proposed change would revise the applicability of the figures in the Technical Specifications for the reactor coolant system (RCS) pressure-temperature limits and the cold overpressure protection setpoints. The proposed change revises the applicability of the figures from 20 effective full-power years (EFPY) to 23.7 EFPY.

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

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

The proposed change does not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions. The proposed change does not alter or prevent the ability of operable SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. The change does not affect the integrity of the RCS pressure boundary. The proposed change to the applicability of the RCS pressure-temperature limits and the cold overpressure protection setpoints continues to protect the integrity of the RCS pressure boundary.

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

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

The proposed change, which revises the applicability of the RCS pressure-temperature limits and the cold overpressure protection setpoints, will not impact the accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a significant change in the method of plant operation, or new operator actions. The proposed change will not introduce failure modes that could result in a new accident. The RCS pressure-temperature limits and the cold overpressure protection setpoints are not accident initiators. The change does not alter assumptions made in the safety analysis.

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

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

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change does not involve a significant change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The proposed change to the applicability of the RCS pressure-temperature limits and the cold overpressure protection setpoints continues to protect the integrity of the RCS pressure boundary.

Therefore, 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC.

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

NRC Branch Chief: Harold K. Chernoff.

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 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 NRC's PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by email to pdr.resource@nrc.gov.

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

Date of application for amendment: March 31, 2011, as supplemented by letter dated August 12, 2011.

Brief description of amendment: The amendments relocated certain surveillance frequencies to a licensee-controlled program (the Surveillance Frequency Control Program) in accordance with Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF (Risk Informed Technical Specification Task Force) Initiative 5b.” The amendments also approved two deviations from TSTF-425, Revision 3: an administrative change which would allow it to retain a definition that also appears in a portion of the plants' technical specifications that are not subject to TSTF-425, and TS Bases changes recommended by the NRC to the TSTF in a letter dated April 14, 2010.

Date of issuance: December 15, 2011.

Effective date: As of the date of issuance and shall be implemented within 180 days from the date of issuance.

Amendment No.: Unit 1—188; Unit 2—188; Unit 3—188.

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: June 14, 2011 (76 FR 34765). The supplemental letter dated August 12, 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 December 15, 2011.

No significant hazards consideration comments received: No.

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of application for amendments: July 12, 2011.

Brief Description of amendments: The license amendments revised Brunswick Steam and Electric Plant, Units 1 and 2 Technical Specification (TS) 3.4.5, “RCS Leakage Detection Instrumentation,” consistent with the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-514, “Revise BWR [Boiling Water Reactor] Operability Requirements and Actions for RCS [Reactor Coolant System] Leakage Instrumentation,” Revision 3. The availability of this TS improvement was announced in the Federal Register on December 17, 2010 (75 FR 79048) as part of the consolidated line item improvement process.

Date of issuance: December 21, 2011.

Effective date: Date of issuance, shall be implemented within 60 days of the effective date.

Amendment Nos.: Unit 1—260 and Unit 2—288.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments revised the technical specifications.

Date of initial notice in Federal Register: September 6, 2011 (76 FR 55127).

The Commission's related evaluation of the amendments is contained in a safety evaluation dated December 21, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 6, 2011.

Brief description of amendment: The amendments modify the actions to be taken when the containment atmosphere gaseous radioactivity monitoring system and the primary containment pressure and temperature monitoring system are the only operable reactor coolant leakage detection monitoring systems. The modified actions require additional, more frequent monitoring of other indications of Reactor Coolant System (RCS) leakage and provide appropriate time to restore another monitoring system to operable status. This change is consistent with the U.S. Nuclear Regulatory Commission-approved safety evaluation on Technical Specification Task Force (TSTF) Traveler, TSTF-514-A, Revision 3, “Revised [Boiling Water Reactor] BWR Operability Requirements and Actions for RCS Leakage Instrumentation,” dated November 24, 2010.

Date of issuance: December 19, 2011.

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

Amendment Nos.: 205 and 167.

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

Date of initial notice in Federal Register: August 9, 2011 (76 FR 48911).

The Commission's related evaluation of the amendment is contained in Safety Evaluation dated December 19, 2011.

No significant hazards consideration comments received: No.

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.

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

Date of application for amendment: December 20, 2010, as supplemented by letters dated July 20, September 1, and October 5, 2011. The July 20, 2011, submittal entirely replaced the licensee's submittal dated December 20, 2010.

Brief description of amendment: Florida Power Corporation (the licensee) will be constructing and operating an onsite independent spent fuel storage installation, under its general license, in order to maintain full-core offload capacity in the spent fuel pools located in the CR-3 auxiliary building (AB). In support of future dry shielded canister/transfer cask loading operation, the licensee is replacing the AB overhead crane. This amendment approved departure from a method for evaluating the replaced AB overhead crane, revisions to the CR-3 Final Safety Analysis Report (FSAR), and changes to the associated commitments in the FSAR.

Date of issuance: December 27, 2011.

Effective date: Date of issuance, to be implemented within 180 days. The FSAR changes shall be implemented in the next periodic update made in accordance with 10 CFR 50.71(e).

Amendment No.: 239.

Facility Operating License No. DPR-72: Amendment approved revisions to the FSAR Sections 5.1.1.1.h, 9.6.1.5.a.5, and 9.6.3.1 as indicated in the NRC's safety evaluation dated December 27, 2011.

Date of initial notice in Federal Register: September 6, 2011 (76 FR 55129). The supplements dated September 1 and October 5, 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.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of application for amendment: May 27, 2009, as supplemented on August 28, 2009, December 23, 2009, February 19, 2010, April 16, 2010, May 7, 2010, June 3, 2010, June 30, 2010, July 9, 2010, July 30, 2010, September 16, 2010, October 8, 2010, October 28, 2010, November 5, 2010, December 10, 2010, December 13, 2010, January 19, 2011, January 31, 2011, February 4, 2011, March 23, 2011, May 9, 2011, June 13, 2011, July 15, 2011, August 5, 2011, August 19, 2011, September 23, 2011, October 27, 2011, and November 1, 2011.

Brief description of amendment: The amendment changes the NMP2 Technical Specifications to increase the maximum steady-state reactor core power level from 3,467 megawatts thermal (MWt) to 3,988 MWt, which is an increase from the current license of approximately 15 percent. The proposed increase in power level is considered an extended power uprate.

Date of issuance: December 22, 2011.

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

Amendment No.: 140.

Renewed Facility Operating License No. NPF-69: The amendment revises the License and TSs.

Date of initial notice in Federal Register: October 10, 2009 (74 FR 53778). The supplemental letters dated August 28, 2009, December 23, 2009, February 19, 2010, April 16, 2010, May 7, 2010, June 3, 2010, June 30, 2010, July 9, 2010, July 30, 2010, September 16, 2010, October 8, 2010, October 28, 2010, November 5, 2010, December 10, 2010, December 13, 2010, January 19, 2011, January 31, 2011, February 4, 2011, March 23, 2011, May 9, 2011, June 13, 2011, July 15, 2011, August 5, 2011, August 19, 2011, September 23, 2011, October 27, 2011, and November 1, 2011, provided additional information that clarified the application and did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 29th day of December 2011.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

[FR Doc. 2012-124 Filed 1-9-12; 8:45 am]

BILLING CODE 7590-01-P