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Notice

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

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Start Preamble

Background

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (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 May 30, 2013 to June 12, 2013. The last biweekly notice was published on June 11, 2013 (78 FR 35058).

ADDRESSES:

You may submit comment by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0134. Address questions about NRC dockets to Carol Start Printed Page 38079Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • 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.

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

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SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2013-0134 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly-available, by the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0134.
  • 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 pdr.resource@nrc.gov. 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-2013-0134 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 that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely 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 that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely 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 Section 50.92 of Title 10 of the Code of Federal Regulations (10 CFR), 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 “Agency Rules of Practice and Procedure” 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 Start Printed Page 38080petition 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's 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'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 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, Start Printed Page 3808111555 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. However, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. 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. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the following three factors in 10 CFR 2.309(c)(1): (i) The information upon which the filing is based was not previously available; (ii) the information upon which the filing is based is materially different from information previously available; and (iii) the filing has been submitted in a timely fashion based on the availability of the subsequent information.

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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3 (MPS-3), New London County, Connecticut

Date of amendment request: April 25, 2013.

Description of amendment request: The amendments would revise the peak calculated containment internal pressure (Pa) for the design basis loss of coolant accident (LOCA) described in Technical Specification (TS) 6.8.4.f, “Containment Leakage Rate Testing Program” for MPS-3.

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

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

Response: No.

The proposed change to Pa does not alter the assumed initiators to any analyzed event. The probability of an accident previously evaluated will not be significantly increased by this proposed change.

The change in Pa will not affect radiological dose consequence analyses. MPS-3 radiological dose consequence analyses assume a certain containment atmosphere leak rate based on the maximum allowable containment leakage rate, which is not affected by the change in peak calculated containment internal pressure. The Appendix J containment leakage rate testing program will continue to ensure that containment leakage remains within the leakage assumed in the offsite dose consequence analyses. The consequences of an accident previously evaluated will not be significantly increased by this proposed change.

Therefore, operation of the facility in accordance with the proposed change to Pa will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change provides a higher Pa than currently described in TS 6.8.4.f. This change is a result of an increase in the M&E [mass and energy] release input for the LOCA containment response analysis. The [Pa] remains below the containment design pressure of 45 psig [pounds per square inch gauge]. This change does not involve any alteration in the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Therefore, operation of the facility in accordance with the proposed change to TS 6.8.4.f would 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 [Pa] remains below the containment design pressure of 45 psig. Since the MPS3 radiological consequence analyses are based on the maximum allowable containment leakage rate, which is not being revised, the change in the [Pa] does not represent a significant change 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

Acting NRC Branch Chief: Robert Beall.

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: April 16, 2013.

Description of amendment request: The proposed amendments would remove superseded Technical Specification (TS) requirements McGuire Nuclear Station (MNS), Units 1 and 2. By letter dated May 28, 2010, Duke Energy submitted a license amendment request (LAR) to modify TS to allow the manual operation of the Containment Spray System in lieu of automatic actuation, and revise the minimum volume and low level setpoint on the Refueling Water Storage Tank. Because the associated modifications were implemented on a staggered basis for each MNS Unit during refueling outages, the deletion or modification of these TS requirements was accomplished via the use of temporary footnotes. This allowed the Start Printed Page 38082requirements to be either applicable or non-applicable, depending upon whether the modifications had not been implemented or implemented, respectively. The LAR contained a commitment for MNS to submit a follow-up administrative license amendment request to delete the superseded temporary TS requirements within 180 days of the installation of the associated modifications for the final MNS Unit. By letter dated September 12, 2011, the NRC issued amendments regarding the TS changes requested in the May 28, 2010 LAR. Installation of the associated modifications on the final MNS Unit was completed on October 18, 2012. This LAR satisfies the MNS commitment to delete the superseded temporary TS requirements described in the May 28, 2010 LAR. In addition, this LAR makes an administrative non-technical editorial correction by relocating NOTE 1 on TS page 3.3.2-15 to TS page 3.3.2-14. Relocating NOTE 1 back to TS page 3.3.2-14 is consistent with the reference to this NOTE in TS Table 3.3.2-1, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, Function 9, Containment Pressure Control System.

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.

This LAR proposes administrative non-technical changes only. These proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configurations of the facility. The proposed changes do not alter or prevent the ability of structures, systems and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

Given the above discussion, it is concluded the proposed amendment does not significantly increase 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.

This LAR proposes administrative non-technical changes only. The proposed changes will not alter the design requirements of any SSC or its function during accident conditions. No new or different accidents result from the changes proposed. The changes do not involve a physical alteration of the plant or any changes in methods governing normal plant operation. The changes do not alter assumptions made in the safety analysis.

Given the above discussion, it is concluded the proposed amendment 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 the margin of safety?

Response: No.

This LAR proposes administrative non-technical changes only. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

Given the above discussion, it is concluded the proposed 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: April 15, 2013.

Description of amendment request: The proposed change would revise Technical Specification 3.5.4, “Refueling Water Storage Tank (RWST)” such that the non-seismically qualified piping of the temporary Boric Acid Recovery System (BARS) may be connected to the seismic piping of the RWST. Operation of the BARS from the RWST will be under administrative controls for a limited period of time (i.e., 30 days for RWST filtration prior to each fuel cycle). This change will only be applicable until Refueling Outage R22 ends (Spring 2016).

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

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

Response: No.

The use of the non seismic Boric Acid Recovery System (BARS) to recirculate and filter the Refueling Water Storage Tank (RWST) water does not involve any changes or create any new interfaces with the reactor coolant system or main steam system piping. Therefore, the connection of the BARS Purification Loop to the RWST would not affect the probability of these accidents occurring. The BARS is not credited for safe shutdown of the plant or accident mitigation. Administrative controls ensure that the BARS can be isolated as necessary and in sufficient time to assure that the RWST volume will be adequate to perform the safety function as designed. Since the RWST will continue to perform its safety function and overall system performance is not affected, the consequences of the accident are not increased.

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

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

Response: No.

The design of the RWST and the SFP [spent fuel pool] Purification Loop has been revised to allow recirculation and purification using the BARS for a short period of time (not to exceed 30 days per fuel cycle) for the next two fuel cycles. The added BARS takes RWST water in and processes it out without additional connections that could affect other systems and without an impact from its installation. Procedures for the operation of the plant, including BARs, will not create the possibility of a new or different type of accident. Contingent upon manual operator action, a BARS line break will not result in a loss of the RWST safety function. Similarly, an active or passive failure in the BARS will not affect safety related structures, systems or components.

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

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

Response: No.

The SFP Purification Loop and recirculation and purification of the RWST water using the BARS is not credited for safe shutdown of the plant or accident mitigation. RWST volume will be maximized prior to purification and timely operator action can be taken to isolate the non seismic system from the RWST to assure it can perform its function. This will result in no significant reduction in the margin of safety.

Therefore the proposed change does not significantly reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this Start Printed Page 38083review, 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., 440 Hamilton Avenue, White Plains, NY 10601.

Acting NRC Branch Chief: Sean Meighan.

National Institute of Standards and Technology (NIST), Docket No. 50-184, Center for Neutron Research (NBSR), Montgomery County, Maryland

Date of amendment request: July 12, 2012, as supplemented on May 14, 2013.

Description of amendment request: The proposed amendments would revise NIST NBSR's Technical specifications, Sections 3.7, 4.7, and 6.8, pertaining to the environmental monitoring requirements and records retention which clarifies environmental sampling procedure and record retention processes.

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

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

Response: No.

The proposed amendment corrects a deficiency in the license issued in 2009 that created a disagreement in the periodicity of environmental sampling within the license Technical Specifications. Additionally, the proposed amendment aligns the record retention requirement (section 6.8) of the license technical specifications with the consensus standard ANSI/ANS 15.1. This standard has been endorsed by the NRC under Regulatory Guide 2.2. Neither of these proposed changes will have any influence or impact on reactor operations or previously analyzed accidents. There are no physical changes to the facility as a result of these administrative changes.

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

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

Response: No.

No accident of any kind would be created by the proposed administrative changes. The sample periodicity will not change from the sampling periodicity used by the facility for over 40 years. Records are maintained and summarized in facility annual reports and there would be no loss of information. There are no physical changes to the facility as a result of these administrative changes.

Therefore, the changes would 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.

Plant safety margins are established through limiting conditions of operation, limiting safety system settings, and safety limits specified in the Technical Specifications. The proposed changes do not alter any of the established safety margins and are administrative in nature.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Melissa J. Lieberman, Deputy Chief Counsel for NIST, National Institute of Standard and Technology, 100 Bureau Drive, Gaithersburg, MD 20899.

NRC Branch Chief: Alexander Adams, Jr.

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

Date of amendment request: April 2, 2013, as supplemented by a letter dated May 16, 2013.

Description of amendment request: The proposed amendments would revise the technical specification requirements regarding steam generator tube inspection and reporting as described in Technical Specification Task Force (TSTF)-510, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,” Revision 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:

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 Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.

Therefore, it is concluded that this 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 changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

Therefore, it is concluded that this 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 SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

Therefore, it is concluded that 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: J. Hagood Hamilton, Jr., South Carolina Electric & Start Printed Page 38084Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: April 3, 2013.

Description of amendment request: The proposed amendment would allow for the extension of the frequency of the containment leak rate test per Technical Specification 6.8.4(g) from 130-months (10.9 years) to 15 years.

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

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 permanent 15-year extension to the current interval for Type A containment testing. The current test interval of 130 months (10.9 years) would be extended to a permanent 15-year frequency from the last Type A test. The proposed extension does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the reactor containment itself and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated nor does it create the possibility of a new or different kind of accident.

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

Based on the above, the proposed extension does not involve a significant increase in the consequences of an accident previously evaluated.

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

Response: No.

The proposed revision to the TS involves a 15-year permanent extension to the current interval for Type A containment testing. The reactor containment and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve the prevention or identification of any precursors of an accident. The proposed TS change does not involve a physical change to the plant or the manner in which the plant is operated or controlled.

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

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

Response: No.

The proposed change to the TS involves a 15-year permanent extension to the current interval for Type A containment testing. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The specific requirements and conditions of the Primary Containment Leak Rate Testing Program, as defined in the TS, exist to ensure that the degree of reactor containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests. The proposed surveillance interval extension is bounded by the 15-year permanent extension currently authorized within NEI 94-01, Revision 3-A. Type B and C containment leak rate tests will continue to be performed at the frequency currently required by TS. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME, Section Xl and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by Type A testing.

The combination of these factors ensures that the margin of safety that is in plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards will continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A test interval. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Robert J. Pascarelli.

Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, Georgia

Date of amendment request: May 10, 2013.

Description of amendment request: The proposed change would amend Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4 by departing from VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by revising reference document APP-OCS-GEH-520, “AP1000 Plant Startup Human Factors Engineering Design Verification Plan,” from Revision B to Revision 1. APP-OCS-GEH-520 is incorporated by reference in the Updated Final Safety Analysis Report (UFSAR) as a means to implement the activities associated with the human factors engineering verification and validation.

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

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

Response: No.

The APP-OCS-GEH-520, document confirms aspects of the human system interface (HSI) and Operation and Control Centers Systems (OCS) design features that could not be evaluated in other Human Factors Engineering (HFE) verification and validation (V&V) activities. It also confirms that the as-built in the plant HSIs, procedures, and training conform to the design that resulted from the HFE program. Additionally, it confirms that all HFE-related issues (including human error discrepancies (HEDs)) documented in the SmartPlant Foundation (SPF) Human Factors (HF) Start Printed Page 38085Tracking System are verified as adequately addressed or resolved. Finally, it confirms the HFE adequacy for risk-important human actions in the local plant, including the ability for the tasks to be completed within the time window according to the Probabilistic Risk Assessment (PRA). The changes to the plan are to clarify the scope and amend the details of the methodology. The plan does not affect the plant itself. Changing the plan does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. The PRA is not affected. No safety-related Structure, System, or Component (SSC) or function is adversely affected. The document revision change does not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected. Because the changes to the plan do not involve any safety-related SSC or function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

Therefore, there is no 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.

APP-OCS-GEH-520, “AP1000 Plant Startup Human Factors Engineering Design Verification Plan” is the plan to confirm aspects of the HSI and OCS design features that could not be evaluated in other HFE V&V activities. The plan also confirms that the as-built in the plant HSIs, procedures, and training conform to the design that resulted from the HFE program. Additionally, it confirms that all HFE-related issues (including HEDs) documented in the SPF HF Tracking System are verified as adequately addressed or resolved. Finally, it confirms the HFE adequacy for risk-important human actions in the local plant, including the ability for the tasks to be completed within the time window according to the PRA. These functions support evaluating the HSI and OCS. Therefore, the changes do not affect the safety-related equipment itself, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected. No system or design function or equipment qualification will be adversely affected by the changes. This activity will not allow for a new fission product release path, nor will it result in a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment.

Therefore, this activity does not create the possibility of a new or different kind of accident than any accident previously evaluated.

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

Response: No.

APP-OCS-GEH-520, “AP1000 Plant Startup Human Factors Engineering Design Verification Plan” is the plan to confirm aspects of the HSI and OCS design features that could not be evaluated in other HFE V&V activities. The plan also confirms that the as-built in the plant HSIs, procedures, and training conform to the design that resulted from the HFE program. Additionally, it confirms that all HFE-related issues (including HEDs) documented in the SPF HF Tracking System are verified as adequately addressed or resolved. Finally, it confirms the HFE adequacy for risk-important human actions in the local plant, including the ability for the tasks to be completed within the time windows in the PRA. These functions support evaluating the HSI and OCS. The proposed changes to the plan do not affect the design or operation of safety-related equipment or equipment whose failure could initiate an accident, nor does the plan adversely affect the interfaces with safety-related equipment or fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes.

Therefore, the 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: Mr. M. Stanford Blanton, Blach & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

Acting NRC Branch Chief: Lawrence Burkhart.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: April 25, 2013.

Description of amendment request: The amendments would revise Technical Specification (TS) 5.1, “Site,” Figures 5.1-1 through 5.1-4 for South Texas Project (STP), Units 1 and 2, to remove identification of a Visitor's Center building, which has been demolished. The amendments also would revise Figures 5.1-1, 5.1-3, and 5.1-4 to remove references to the Emergency Operations Facility (EOF) within the Nuclear Training Facility, since the EOF was relocated to Center of Energy Development building located in Bay City, Texas, approximately 12.5 air miles from the plant site in 2009. The EOF was relocated offsite with an emergency plan change made by the licensee under 10 CFR 50.54(q), “Emergency plans,” by concluding that the change did not represent a decrease in effectiveness of the emergency plan. The amendments to remove references to the Visitor's Center Building and EOF from the TSs are administrative in nature.

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 is an administrative change to STP TS design features to remove reference to the Visitor's Center and onsite EOF. The design function of structures, systems and components (SSC) important to safety are not impacted by the proposed change. The proposed change will not initiate an event. The proposed change does not alter or prevent the ability of SSCs from performing their intended function to mitigate the consequences of an initiating event.

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 is an administrative change to STP TS design features to remove reference to the Visitor's Center and onsite EOF. The proposed change does not impact create the possibility of a new or different kind of accident from any accident previously evaluated. There are no new failure modes or mechanisms associated with the proposed change. This change does not involve any modification in operational limits or physical design of equipment important to safety.

Therefore, the proposed change does not involve a 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 is an administrative change to STP TS design features to remove reference to the Visitor's Center and onsite EOF. The proposed change does not impact TS safety limits, TS limiting safety system set points, or the results of any of the safety analyses.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that Start Printed Page 38086the request for amendments involves no significant hazards consideration.

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.

NRC Branch Chief: Michael T. Markley.

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.

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

Date of amendment request: May 22, 2013.

Brief description of amendment request: The proposed amendment would revise the WBN Unit 1 Technical Specifications (TSs) to allow a one-time extension to the Completion Time for TS Limiting Condition for Operation (LCO) 3.6.6 Required Action A.1 from 72 hours to 7 days for an inoperable Containment Spray (CS) Train B. This change is necessary to provide sufficient time to replace a leaking mechanical seal on CS Pump 1B-B. The pump repair is currently scheduled for the week of June 24, 2013. TVA requested this TS change under exigent circumstances and that the NRC expedites the review to support approval by June 22, 2013.

Date of publication of individual notice in Federal Register : June 3, 2013 (78 FR 33117).

Expiration date of individual notice: June 17, 2013 (public comments); August 2, 2013 (hearing requests).

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 pdr.resource@nrc.gov.

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

Date of application for amendment: September 6, 2012, as supplemented by letter dated December 7, 2012.

Brief Description of amendment: The amendment revised the Technical Specifications (TSs) to eliminate Function 14, Steam Generator Water Level-Low Coincident with Steam Flow/Feedwater Flow Mistmatch, from the HBRSEP TS Table 3.3.1-1, “Reactor Protection System Instrumentation.”

Date of issuance: May 29, 2013.

Effective date: As of date of issuance and shall be implemented prior exiting the scheduled fall 2013 refueling outage.

Amendment No.: 234.

Renewed Facility Operating License No. DPR-23: Amendment changed the license and TSs.

Date of initial notice in Federal Register: November 27, 2012 (77 FR 70840). The supplement dated December 7, 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.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 29, 2013.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit 2, New London County, Connecticut

Date of amendment request: July 21, 2010.

Description of amendment request: The proposed amendment revised the Technical Specification (TS) 3/4.9.3.1, “Decay Time” for Millstone Power Station, Unit 2 (MPS2). The proposed change revises TS 3/4.9.3.1 by reducing the minimum decay time for irradiated fuel prior to movement in the reactor vessel from 150 hours to 100 hours. The licensee requested a reduction in the minimum decay time requirement to provide additional flexibility in outage planning such that irradiated fuel can be moved from the reactor vessel to the spent fuel pool earlier in an outage.

Date of issuance: June 4, 2013.

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

Amendment No.: 315.

Renewed Facility Operating License No. DPR-65: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: April 2, 2013 (78 FR 19749). The supplemental letter dated July 19, 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 Start Printed Page 38087significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 4, 2013.

No significant hazards consideration comments received: No.

South Carolina Electric and Gas. Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 3 and 4, Fairfield County, South Carolina

Date of amendment request: February 14, 2013.

Brief description of amendment: The amendment authorizes a departure from the Virgil C. Summer Nuclear Station, Units 2 and 3 plant-specific Design Control Document (DCD) material incorporated into the Updated Final Safety Analysis Report (UFSAR) to revise Figure 3.8.8-1, Sheet 1, Note 2.

Date of issuance: May 23, 2013.

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

Amendment No.: Unit 2-3, and Unit 3-3.

Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14126).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 23, 2013.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 14th day of June 2013.

For The Nuclear Regulatory Commission.

John D. Monninger,

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

End Signature End Supplemental Information

[FR Doc. 2013-14880 Filed 6-24-13; 8:45 am]

BILLING CODE 7590-01-P