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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 (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 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 April 4, 2013 to April 17, 2013. The last biweekly notice was published on April 16, 2013 (78 FR 22563).

ADDRESSES:

Please refer to Docket ID NRC-2013-0074 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, using any of the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0074. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov.
  • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
  • Fax comments to: RADB at 301-492-3446.

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

End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2013-0074 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-2013-0074.
  • 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-2013-0074 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 Start Printed Page 25311publication 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 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 the 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 Start Printed Page 25312participant 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 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, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. 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.

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

Date of amendment request: February 7, 2013.

Description of amendment request: Entergy Operations, Inc. (the licensee), proposes to revise RBS Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” Surveillance Requirements (SRs) 3.8.4.2 and 3.8.4.5. The changes to the SRs will add new acceptance criteria to address possible non-conservative conditions when the battery connection resistances are at maximum values.

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 changes are to the surveillance requirements only. The ability of the TS surveillance to ensure that the batteries have the capacity to perform their specified safety functions with regard to accident mitigation or meeting their licensing design basis requirements is not reduced/diminished.

There are no design changes associated with this TS amendment. The DC power system/batteries will retain adequate independency, redundancy, capacity and testability to permit the functioning required of the engineered safety features. The batteries will each continue to independently Start Printed Page 25313provide this capacity assuming a failure of a single active component. The proposed change will not affect accident initiators or precursors, or adversely alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated. The proposed change will not alter the ability of structures, systems and components to perform their intended functions to mitigate the consequences of an initiating event. The proposed change does not physically alter safety related systems nor affect the way in which safety related systems perform their function.

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 involves only surveillance test acceptance criteria. The ability of the TS surveillance to ensure that the batteries have the capacity to perform their specified safety functions with regard to accident mitigation or meeting their licensing design basis requirements is not reduced/diminished.

There are no proposed design changes, nor are there any changes in the method by which any safety related plant structure, system, or component (SSC) performs its specified safety function. The proposed change will not affect the normal method of plant operation or change any operating parameters. Equipment performance necessary to fulfill safety analysis missions will be unaffected. The proposed change will not alter any assumptions required to meet the safety analysis acceptance criteria. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced because of this amendment. There will be no adverse effect or challenges imposed on any safety related system because of this amendment.

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

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

Response: No.

The proposed change does not reduce the ability of the TS surveillance requirements to ensure that the station batteries have adequate capacity to perform their engineered safety features functions with regard to accident mitigation and meeting their licensing design basis requirements. The lower battery inter-cell connection resistance values are more restrictive, consistent with design basis calculations and appropriately identified in maintenance procedures. The proposed changes do not physically alter safety related systems. There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. The applicable radiological dose consequence acceptance criteria will continue to be met.

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: January 28, 2013.

Description of amendment request: The ANO-1 Technical Specification (TS) requirements are revised from requirements on battery cells to requirements on the battery. This focuses the requirements on the assumed safety function of the battery. The proposed amendment would revise TS requirements related to direct current (DC) electrical systems in TS Limiting Condition for Operation (LCO) 3.8.4, “DC Sources—Operating,” LCO 3.8.5, “DC Sources—Shutdown,” and LCO 3.8.6, “Battery Parameters.” A new “Battery Monitoring and Maintenance Program” is being proposed for Section 5.5, “Administrative Controls—Programs and Manuals.”

These changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.” The availability of this TS improvement was announced in the Federal Register on September 1, 2011 (76 FR 54510), as part of the consolidated line item improvement process (CLIIP).

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 changes restructure the Technical Specifications (TS) for the direct current (DC) electrical power system and are consistent with TSTF-500, Revision 2. The proposed changes modify TS Actions relating to battery and battery charger operability requirements. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Safety Analysis Report (SAR). Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure TS and change surveillances for batteries and chargers to incorporate the applicable updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the SAR. Operation in accordance with the proposed TS would ensure that the DC electrical power system is capable of performing its specified safety function as described in the SAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis. A new licensee-controlled Battery Monitoring and Maintenance Program will ensure appropriate monitoring and maintenance that is consistent with industry standards. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC electrical power system.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the SAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes.

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

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

Response: No.

The proposed changes involve restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the SAR. Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure the TS and change surveillances for batteries and chargers to incorporate the applicable updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the SAR. Administrative and mechanical controls are in place to ensure the design and operation of the DC systems continues to meet the plant design basis described in the SAR.

Therefore, operation of the facility in accordance with this 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.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The equipment margins will be Start Printed Page 25314maintained in accordance with the plant-specific design bases as a result of the proposed changes. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new Battery Maintenance and Monitoring Program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety-related loads in accordance with analysis assumptions.

TS changes made in accordance with TSTF-500, Revision 2, maintain the same level of equipment performance stated in the SAR and the current TSs.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: January 28, 2013.

Description of amendment request: The ANO-2 Technical Specification (TS) requirements are revised from requirements on battery cells to requirements on the battery. This focuses the requirements on the assumed safety function of the battery. The proposed amendment would revise the TS requirements related to direct current (DC) electrical systems in TS Limiting Condition for Operation (LCO) 3.8.2.3, “DC Distribution—Operating,” and LCO 3.8.2.4, “DC Distribution—Shutdown.” Because ANO-2 is a custom TS plant, a new TS 3.8.3, “Battery Parameters,” would be created to capture the intent of Standard TS (STS) LCO 3.8.6, “Battery Parameters,” as modified by TSTF-500. A new “Battery Monitoring and Maintenance Program” is also being proposed for Section 6.5, “Administrative Controls—Programs and Manuals.”

These changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.” The availability of this TS improvement was announced in the Federal Register on September 1, 2011 (76 FR 54510), as part of the consolidated line item improvement process (CLIIP).

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 changes restructure the Technical Specifications (TS) for the direct current (DC) electrical power system and are consistent with TSTF-500, Revision 2. The proposed changes modify TS Actions relating to battery and battery charger operability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Safety Analysis Report (SAR). Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure TS and change surveillances for batteries and chargers to incorporate the applicable updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the SAR. Operation in accordance with the proposed TS would ensure that the DC electrical power system is capable of performing its specified safety function as described in the SAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis. A new licensee-controlled Battery Monitoring and Maintenance Program will ensure appropriate monitoring and maintenance that is consistent with industry standards. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC electrical power system.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the SAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes.

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

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

Response: No.

The proposed changes involve restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the SAR. Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure the TS and change surveillances for batteries and chargers to incorporate the applicable updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the SAR. Administrative and mechanical controls are in place to ensure the design and operation of the DC systems continues to meet the plant design basis described in the SAR.

Therefore, operation of the facility in accordance with this 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.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The equipment margins will be maintained in accordance with the plant-specific design bases as a result of the proposed changes. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new Battery Maintenance and Monitoring Program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety-related loads in accordance with analysis assumptions.

TS changes made in accordance with TSTF-500, Revision 2, maintain the same level of equipment performance stated in the SAR and the current TSs.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: March 20, 2013.

Description of amendment request: Due to the pending corporate name Start Printed Page 25315change for the licensee of CR-3, the licensee is requesting that an amendment be made to this license to reflect the change in the name of the licensee from Florida Power Corporation to Duke Energy Florida, Inc.

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?

The proposed amendment involves a change of the corporate name from Florida Power Corporation to Duke Energy Florida, Inc. The proposed amendment does not involve any change in the technical qualifications of the licensee or the plant's design, configuration, or operation. All Limiting Conditions for Operation, Limiting Safety System Settings and Safety Limits specified in the CR-3 Improved Technical Specifications remain unchanged. Also, the Physical Security Plan and related plans, the Operator Training and Requalification Program, the Quality Assurance Program, and the Emergency Plan will not be materially changed by the proposed name change. The corporate name change amendment will not affect the executive oversight provided by the Chief Nuclear Officer and his staff.

Therefore, the proposed amendment 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?

The proposed amendment does not involve any change in the plant's design, configuration, or operation. The current plant design, design bases, and plant safety analysis will remain the same.

The Limiting Conditions for Operations, Limiting Safety System Settings and Safety Limits specified in the CR-3 Improved Technical Specifications are not affected by the proposed corporate name change. As such, the plant conditions for which the design basis accident analysis was performed remain valid.

The proposed amendment does not introduce a new mode of plant operation or new accident precursors, does not involve any physical alterations to plant configuration, or make changes to system set points that could initiate a new or different kind of accident.

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

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

The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not affect either the way in which the plant structures, systems, and components perform their safety function or its design and licensing bases.

Plant safety margins are established through Limiting Conditions for Operation, Limiting Safety System Settings and Safety Limits specified in the Technical Specifications. Because there is no change to the physical design of the plant, there is no change to any of these margins.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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, 550 South Tryon Street, Charlotte NC 28202.

NRC Branch Chief: Jessie Quichocho.

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

Date of amendment request: March 13, 2013.

Description of amendment request: The proposed amendment will revise the Seabrook Technical Specifications (TSs). Specifically, the proposed amendment will modify the circuitry that initiates high-head safety injection (SI) by adding a new permissive, cold leg injection permissive. This permissive prevents opening of the high head SI valves until the reactor coolant system pressure decreases to the cold leg injection permissive set point.

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, along with NRC edits in square brackets:

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

The proposed change adds an additional permissive before high head safety injection is initiated to assist the operators in mitigating the consequences of an inadvertent initiation of the emergency core cooling system (ECCS). This change in the ECCS actuation circuitry does not increase the probability of any accident previously evaluated because:

  • There is no effect on any of the systems, structures, or components that are used for normal operation of the plant,
  • There is no effect on any of the fission product barriers,
  • This change will not affect the normal operating procedures,

The revised circuitry will delay the initiation of high head SI until reactor coolant pressure is below the CLIP [cold leg injection permissive] setpoint; however, the proposed change does not significantly increase the consequences of accidents previously evaluated. The proposed change does not alter ECCS flow. The delayed opening of the high head SI valves has been evaluated for the effect on the consequences of the following:

  • Mass and energy release for steam line break accidents,
  • Steam line break—UFSAR section 15.1.5 (specifically hot zero-power conditions)
  • Feedwater line break—UFSAR section 15.2.8
  • Inadvertent operation of emergency core cooling system during power operation—UFSAR section 15.5.1
  • Chemical and volume control system malfunction that increases reactor coolant inventory—UFSAR section 15.5.2

For all of the above evaluated accidents, the analysis results continue to meet all the safety limits. For the inadvertent initiation of ECCS event, the proposed change assists the operators in mitigating the event by significantly extending the time for the pressurizer to fill. Additional evaluations of small break LOCA [loss-of-coolant accident], best estimate large break LOCA, long term cooling, LOCA forces, cold overpressure mitigation/low temperature over pressure protection, steam generator tube rupture, and LOCA mass and energy release were performed and it was concluded that they were not affected by this change.

In addition evaluations were performed for the centrifugal charging pumps and reactor vessel internals; and for the NSSS [nuclear steam supply system] design transients to determine if the change in the timing of the high head injection would have an effect and it was concluded that these components and transients are not adversely affected.

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 adds new components to the process protection racks and solid state protection system similar to the components and configurations that are already installed. The sequence of operation of equipment used to mitigate the consequences of an accident is changed; however, it does not add any different types of equipment. The proposed change is a change to the protection circuitry for the plant and not to the system or equipment used for normal operation of the plant. It does not alter any fluid flow paths or fission product barriers and does not change the method of control of any plant systems used for normal operations. The proposed change does not alter or prevent the ability of the ECCS to perform its specified function to mitigate the consequences of an initiating event within assumed acceptance limits. The evaluation of the centrifugal charging pumps, reactor internals, control systems and NSSS design transients confirmed that new failure modes were not created.Start Printed Page 25316

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 changes do 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 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 involve a change in the timing of the mitigation of inadvertent ECCS actuation and steam line break.

This change provides additional time for mitigating the inadvertent operation of emergency core cooling system during power operation event prior to filling the pressurizer water solid, by preventing the injection of high head safety injection when it is not required.

This change delays the injection of high head safety injection on a steam line break. The delay has no effect on the steamline break mass and energy releases and the limiting analysis of record hot zero power steam line break as discussed below.

An evaluation was performed to address the impact of the CLIP modification on the steamline break (SLB) mass and energy release stretch power uprate (SPU) analyses, the current analysis of record. For the steamline break mass and energy analyses, the CLIP modification has the potential to delay initiation of ECCS injection by inhibiting auto-open of the cold leg injection valves until both an S-signal and a CLIP signal are present. There are three parts to the evaluation: Part 1 addresses the licensing-basis cases for steamline break mass and energy release inside containment, part 2 addresses the licensing-basis cases for steamline break mass and energy release outside containment, and part 3 addresses steamline break s smaller than those analyzed for the updated final safety analysis report (UFSAR) for which there may be an S-signal but no signal associated with the CLIP.

Steamline break inside containment. For these breaks, two different break types are analyzed: double-ended ruptures and split breaks. All cases from the SPU analysis were reviewed with respect to the timing of SI flow actuation from the analysis of record and when SI flow delivery with CLIP occurs.

In the SPU steamline break mass and energy release analysis for double-ended ruptures, the first signal is low steam pressure for all cases. Using the SPU analysis output results, the assumed time of SI flow delivery is compared to the time when SI flow delivery with CLIP occurs. The results are that for all of the double-ended ruptures, SI flow delivery with CLIP is not reached until after the time assumed for SI flow delivery in the SPU analysis. Although an increase in safety injection delay is considered nonconservative, a sensitivity calculation was specifically performed to evaluate the impact of safety injection and the results show that mass and energy releases are not impacted by the increased delay time for safety injection. These results were expected as the ECCS injection occurs at relatively low flow rates due to high reactor coolant system pressure, and boron injection occurs long after the return to power has been mitigated by increasing reactor coolant system temperature. Any delay in initiation of ECCS injection has a negligible effect on core cooling throughout the event and core reactivity during the initial return to power.

In the SPU steamline break mass and energy release analysis for split breaks, the first signal is the time of the first high containment pressure setpoint. Using the SPU analysis output results, the assumed time of SI flow delivery is compared to the time when SI flow delivery with CLIP occurs. The results are that for all of the split breaks, SI flow delivery with CLIP is not reached until after the time assumed for SI flow delivery in the SPU analysis. Although an increase in safety injection delay is considered non-conservative, a sensitivity calculation was specifically performed to evaluate the impact of safety injection and the results show that mass and energy releases are not impacted by the increased delay time for Safety Injection. These results were expected as the ECCS injection occurs at relatively low flow rates due to high reactor coolant system pressure, and boron injection occurs long after the return to power has been mitigated by increasing reactor coolant system temperature. Any delay in initiation of ECCS injection has a negligible effect on core cooling throughout the event and core reactivity during the initial return to power.

Steamline break outside containment. The SPU analysis for the steamline break mass and energy release outside containment was also evaluated for the CLIP modification. Each steamline break case actuated ECCS flow on a low-low pressurizer pressure S-signal. The CLIP modification requires an S-signal and a CLIP signal. The results show that the credited S-signal is much later than the CLIP signal. The results from the SPU analysis remain valid and bounding for the CLIP modification.

Smaller Steamline breaks. For the condition involving an S-signal actuation with pressurizer pressure above the CLIP setpoint, sensitivity cases varying the start time for ECCS injection, including no ECCS injection have concluded that the instantaneous and integrated mass and energy releases are insensitive to the injection start time. These results were expected as the ECCS injection occurs at relatively low flow rates due to high reactor coolant system pressure, and boron injection occurs long after the return to power has been mitigated by increasing reactor coolant system temperature. Any delay in initiation of ECCS injection has a negligible effect on core cooling throughout the event and core reactivity during the initial return to power.

The hot zero-power steamline break event remains bounding for operation at the current uprate conditions. The CLIP modification does not impact the limiting case for hot zero-power steamline break results because the cold leg injection valves will be fully open before the as-modeled high head safety injection flow starts. In addition, sensitivity studies confirm that the maximum break size remains bounding for the hot zero-power steamline break event with the CLIP modification.

The above evaluation shows that the installation of a CLIP would not impact the Seabrook steamline break mass and energy release licensing basis or the hot zero-power steam line break results.

The feedline break (FLB) has been reanalyzed with the additional conservatism, with respect to the SPU FLB analysis, of assuming no safety injection flow. The results of the analysis show that all the safety limits continue to be met even with the additional conservatism of no safety injection assumed. The assumption that operator action is required to mitigate the consequences of a chemical and volume control malfunction is not changed by this modification. Before CLIP, the event was bounded by the inadvertent ECCS actuation event and its associated operator action. With CLIP, the event requires operator action to terminate charging and seal injection flows. As discussed above, the consequences of the other accidents evaluated remain bounded by the analyses of record. The results of analyses and evaluations supporting the proposed change demonstrate acceptance criteria continue to be met.

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 no significant hazards consideration.

Attorney for licensee: Mr. James Petro, Managing Attorney, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Meena Khanna.

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

Date of amendment request: March 27, 2013.

Description of amendment request: The proposed amendment will revise the Seabrook Technical Specifications (TSs). Specifically, the proposed amendment will modify TS requirements regarding steam generator tube inspections and reporting as described in TS Task Force (TSTF)-510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,” using the Consolidated Line Item Improvement Process (CLIIP). The Start Printed Page 25317changes are consistent with Industry/TSTF Standard Technical Specification Change Traveler, TSTF-510. The availability of this TS improvement was announced in the Federal Register on October 27, 2011 (76 FR 66763), as part of the CLIIP.

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 is 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Meena Khanna.

Virginia Electric and Power Company, Docket No. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Date of amendment request: February 22, 2013.

Description of amendment request: The proposed change will allow the sequence and overlap limits to be exceeded and TS 3.1.6.C Action entered if a failure is identified during the performance of Surveillance Requirement (SR) 3.1.4.2, which verifies control rod freedom of movement. This will align the sequence and overlap limit of Condition A with the control bank insertion limit Condition B. The control bank insertion limit of Condition B was modified with this same change in Amendments 179 and 160. The subsequent change to Improved Technical Specifications (ITS) added the Condition for sequence and overlap limits but failed to include the exception if a failure is identified during control rod freedom of movement testing.

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

The proposed amendment would modify the North Anna Power Station current licensing basis by increasing the time that a single rod bank may be permitted to be outside of sequence and overlap limits. The new allowance only applies to minor sequence and overlap limit differences. The proposed change will result in a small increase in the probability that, at any given time, a control bank will be inserted outside of sequence and overlap limits. However, the probability of occurrence of previously evaluated accidents is not affected, since the existing TS already permit a similar deviation with respect to insertion limit. Only the allowed duration of the sequence and overlap limits' exceedance is being changed.

The allowed misalignment is not a malfunction of equipment important to safety; therefore, the probability of such a malfunction is not increased. A single rod bank's position within 18 steps of its sequence and overlap limits does not significantly increase the probability of a malfunction of a component important to safety. This change does not impact the requirement that the rod bank shall be operable (i.e., trippable); as such, it remains able to fulfill its safety function. Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident.

Therefore, neither the probability of occurrence nor the consequences of an accident previously evaluated is significantly increased.

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

The proposed amendment does not create any new allowances for operating the plant. Only the duration of an existing allowance is being lengthened, with additional restrictions being applied during the extended allowance. No physical changes are being made to any portion of the plant, so no new accident causal mechanisms are being introduced. The proposed change does not result in any new mechanisms that could initiate damage to the reactor or its principal safety barriers (i.e., fuel cladding, reactor coolant system, or primary containment).

Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created.

Criterion 3—Does this change involve a significant reduction in a margin of safety?

The proposed amendment does not affect the inputs or assumptions of any of the design basis analyses that demonstrate the integrity of the fuel cladding, reactor coolant system, or containment during accident conditions. Operation within the proposed limits will not cause unacceptable core radial peaking factors that could result in exceeding departure from nucleate boiling (DNB) limits. Operation within the sequence and overlap limit differences will not result in shutdown margins lower than assumed in the accident analyses. Control and Shutdown rods will remain fully operable (i.e., trippable) during the duration of the proposed extended allowance.

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

Start Printed Page 25318

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

NRC Branch Chief: Robert Pascarelli.

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

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

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

Southern California Edison, Docket No. 50-361, San Onofre Nuclear Generating Station, Unit 2, San Diego County, California

Date of amendment request: April 5, 2013, as supplemented by letter dated April 9, 2013.

Brief description of amendment request: The proposed amendment makes a temporary change to the steam generator management program and the license condition for maximum power. For the duration of Unit 2, Cycle 17, the proposed amendment would change the terms “full range of normal operating conditions” and “normal steady state full power operation” and restricts operation to 70 percent of the maximum authorized power level. “Full range of normal operating conditions” and “normal steady state full power operation” shall be based upon the steam generators being operated under conditions associated with reactor core power levels up to 70 percent Rated Thermal Power (2406.6 megawatts thermal).

Date of publication of individual notice in Federal Register: April 16, 2013 (78 FR 22576).

Expiration date of individual notice: May 16, 2013 (public comments) and June 17, 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 online 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.

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: June 22, 2011, as supplemented by letters dated December 9, 2011, January 27, 2012, and January 30, 2013.

Brief description of amendment: The amendments revised Technical Specification (TS) 3.7.4, “Atmospheric Dump Valves (ADVs).” Specifically, the amendments revised the Limiting Condition for Operation for TS 3.7.4, with corresponding revisions to the TS Conditions, Required Actions, and Completion Times associated with one or more inoperable ADV lines.

Date of issuance: April 11, 2013.

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

Amendment No.: Unit 1-191; Unit 2-191; Unit 3-191.

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: October 4, 2011 (76 FR 61394). The supplemental letters dated December 9, 2011, January 27, 2012, and January 30, 2013, 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 April 11, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 22, 2012.

Brief Description of amendment: The amendment revised the Technical Specification (TS) surveillance requirements for addressing missed surveillances, and is consistent with the Nuclear Regulatory Commission approved Revision 6 of Technical Specification Task Force (TSTF) Standard TSs Change Traveler TSTF-358, “Missed Surveillance Requirements.”

Date of issuance: April 11, 2013.

Effective date: As of date of issuance and shall be implemented within 90 days.Start Printed Page 25319

Amendment No.: 141.

Renewed Facility Operating License No. NPF-63: Amendment revised the TSs.

Date of initial notice in Federal Register: November 27, 2012 (77 FR 70839).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 11, 2013.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit 3, Westchester County, New York

Date of application for amendment: May 23, 2012, as supplemented on August 3, 2012.

Brief description of amendment: The amendment revises Technical Specification 3.7.4, “Atmospheric Dump Valves (ADVs),” Limiting Condition for Operation 3.7.4 to require four operable ADVs instead of three.

Date of issuance: April 15, 2013.

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

Amendment No.: 251.

Facility Operating License Nos. DPR-26 and DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: September 14, 2012 (77 FR 56880).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 15, 2013.

No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VYNPS), Vernon, Vermont

Date of amendment request: March 5, 2012, as supplemented on November 20, 2012, March 26, March 29, and April 5, 2013.

Brief description of amendment: The amendment revised the VYNPS License Condition 3.P and 3.Q to clarify that the information in the updated final safety analysis report (UFSAR) supplement submitted pursuant to Section 54.21(d) of Title 10 of the Code of Federal Regulations (10 CFR), as revised during the license renewal application review process, and as supplemented by commitments of Appendix A of Supplement 2 of NUREG-1907, can be incorporated as part of the UFSAR and may be changed without prior NRC approval provided the requirements of 10 CFR 50.59 have been previously satisfied.

Date of Issuance: April 17, 2013.

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

Amendment No.: 256.

Facility Operating License No. DPR-28: The amendment revised the License.

Date of initial notice in Federal Register: April 3, 2012 (77 FR 20074). The supplemental correspondence dated November 20, 2012, March 26, March 29, and April 5, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 17, 2013.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

Date of application for amendment: March 22, 2012, as supplemented by letter dated. December 3, 2012.

Brief description of amendment: The amendment modifies technical specification (TS) requirements regarding steam generator tube inspections and reporting as described in Technical Specifications Task Force (TSTF)-510, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,” with proposed variations and deviations.

Date of issuance: March 21, 2013.

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

Amendment Nos.: 172 and 172, respectively.

Facility Operating License Nos. NPF-72 and NPF-77: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: (77 FR 31660; May 29, 2012).

The December 3, 2012, supplement did not increase the scope of the application and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 21, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 20, 2012.

Description of amendment request: The amendment revised the Seabrook Technical Specifications (TS) TS 6.7.6.m, “Reactor Coolant Pump Flywheel Inspection Program.” The amendment extends the reactor coolant pump (RCP) motor flywheel examination frequency from the currently approved 10-year inspection interval, to an interval not to exceed 20 years. The changes are consistent with Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-421, “Revision to RCP Flywheel Inspection Program (WCAP-15666).” The availability of this TS improvement was announced in the Federal Register on October 22, 2003, as part of the consolidated line item improvement process (CLIIP).

Date of issuance: April 4, 2013.

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

Amendment No.: 134.

Facility Operating License No. NPF-86: The amendment revised the License and TS.

Date of initial notice in Federal Register: January 22, 2013 (78 FR 4473).

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

No significant hazards consideration comments received: No.

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

Date of amendment request: November 17, 2011, as supplemented by letters dated December 3, 2012, and January 9, 2013.

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

Date of issuance: April 15, 2013.

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

Amendment No.: 135.

Facility Operating License No. NPF-86: The amendment revised the TS and the License.

Date of initial notice in Federal Register: January 10, 2012 (77 FR 1519). The supplements dated December 3, 2012, and January 9, 2013, provided additional information that clarified the application, did not expand Start Printed Page 25320the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 15, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: November 30, 2011, as supplemented by letters dated August 16 and December 7, 2012, and March 3, 2013.

Brief description of amendment: The amendment revised the Technical Specification 3.8.1, “AC [Alternating Current] Sources—Operating,” Surveillance Requirements related to Diesel Generator test loads, voltage, and frequency. The changes correct non-conservative Diesel Generator load values that are currently under administrative controls.

Date of issuance: April 11, 2013.

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

Amendment No.: 204.

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

Date of initial notice in Federal Register: June 12, 2012 (77 FR 35078). The supplemental letters dated August 16 and December 7, 2012, and March 3, 2013, 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 April 11, 2013.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 22nd day of April 2013.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

End Signature End Supplemental Information

[FR Doc. 2013-10020 Filed 4-29-13; 8:45 am]

BILLING CODE 7590-01-P