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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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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 July 11, 2013, to July 23, 2013. The last biweekly notice was published on July 23, 2013, (78 FR 44167).

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-2013-0175. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; Start Printed Page 47786email: 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: 3WFN, 06A44M, 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-2013-0175 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-2013-0175.
  • 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-2013-0175 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 petition must also identify the specific Start Printed Page 47787contentions 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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by email to 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: Start Printed Page 47788Rulemaking 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 three factors in 10 CFR 2.309(c)(1)(i)(iii).

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

Date of amendment request: May 28, 2013.

Description of amendment request: The amendment would modify safety limits (SL) in Technical Specification (TS) 2.1.1, “Reactor Core SLs,” to reduce the minimum reactor dome pressure associated with the critical power correlation from 785 pounds per square inch gauge (psig) to 685 psig. The RBS has evaluated the critical power correlation for the General Electric Nuclear Energy advanced fuel designs (i.e., GE14 and GNF2 fuels) used at the facility which will allow for a lower-bound pressure. The change will provide a greater pressure margin such that the reactor remains above the proposed low SL of 685 psig in the event of a Pressure Regulator Maximum Demand Open transient.

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.

Decreasing the reactor dome pressure limit in TS Safety Limits 2.1.1 for reactor Rated Thermal Power range effectively expands the validity range for the GEXL 14 and GEXL 17 correlations and the calculation of Minimum Critical Power Ratio Safety Limit (MCPR). The MCPR rises during the pressure reduction following the scram that terminates the Pressure Regulator Failure Open (PRFO) transient. Since the change does not involve a modification of any plant hardware, the probability and consequence of the PRFO transient are essentially unchanged. The reduction in the reactor dome pressure safety limit from 785 psig to 685 psig provides greater margin to accommodate the pressure reduction during the transient within the revised TS limit.

The proposed change will continue to support the validity range for the GEXL correlations applied at RBS and the calculation of MCPR as approved. The proposed TS revision involves no significant changes to the operation of any systems or components in normal, accident or transient operating conditions.

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

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

Response: No.

The proposed reduction in the reactor dome pressure safety limit from 785 psig to 685 psig is a change based upon previously approved documents and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced.

Therefore, the change does not introduce a new or different kind of accident from those previously evaluated.

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

Response: No.

The margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. The proposed change in reactor dome pressure enhances the safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged. The available pressure range is expanded by the change, thus offering greater margin for pressure reduction during the transient.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: April 5, 2013.

Description of amendment request: The proposed amendment would revise the Pilgrim Technical Specifications (TSs) to reduce the reactor steam dome pressure from 785 pounds per square inch, gauge (psig) to 685 psig specified in TS Reactor Core Safety Limits 2.1.1 and 2.1.2. The proposed amendment is intended to address the potential to exceed the low pressure TS safety limit associated with a pressure regulator failure open (PRFO)—maximum Start Printed Page 47789demand abnormal operation occurrence, as identified by General Electric Nuclear Energy in its report, “10 CFR 21 Reportable Condition Notification: Potential to Exceed Low Pressure Technical Specification Safety Limit,” MFN 05-021, dated March 29, 2005.

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 the NRC's edits in square brackets:

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

Response: No.

Decreasing the reactor dome pressure in Technical Specification Safety Limits 2.1.1 and 2.1.2 for reactor Rated Thermal Power ranges effectively expands the validity range for GEXL [GE critical quality-boiling length correlation] and the calculation of Minimum Critical Power Ratio Safety Limit (MCPR). MCPR rises during the pressure reduction following the scram that terminates the PRFO transient. Since the change does not involve a modification of any plant hardware, the probability and consequence of the PRFO transient are essentially unchanged. The reduction in the reactor dome pressure value in the safety limit from 785 psig to 685 psig provides adequate margin to accommodate the pressure reduction during the transient within the revised TS limit.

The expanded GEXL correlation range supports Pilgrim's revised low pressure safety limit of 685 psig. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident or transient operating conditions.

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

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

Response: No.

The proposed reduction in the reactor dome pressure value in the safety limit from 785 psig to 685 psig reflects a wider range of applicability for the GEXL correlation which is approved by the NRC for fuels in use at Pilgrim and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced.

Therefore, the [proposed] change does not [create the possibility of] a new or different kind of accident from any [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 the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. The proposed change in reactor dome pressure restores the safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged. The reduction in the reactor dome pressure value in the safety limit from 785 psig to 685 psig provides adequate margin to accommodate the pressure reduction during the transient within the revised TS limit.

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

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

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

NRC Acting Branch Chief: Robert Beall.

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

Date of amendment request: May 14, 2013.

Description of amendment request: The proposed amendment would revise the Vermont Yankee Technical Specifications (TSs) to reduce reactor pressure associated with the fuel cladding integrity safety limits (SLs) from 800 pounds per square inch, absolute (psia) to 700 psia in SLs 1.1.A and 1.1.B. The proposed change is intended to address the potential to exceed the low pressure TS SL associated with a pressure regulator failure-maximum demand open (PRFO) transient as reported by General Electric Nuclear Energy in its Part 21 Communication, “Potential to Exceed Low Pressure Technical Specification Safety Limit,” SC05-03, dated March 29, 2005.

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 change to the reactor pressure in Fuel Cladding Integrity Safety Limits 1.1.A and 1.1.B does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed change is in accordance with NRC approved critical power correlation methodologies and as such maintains required safety margins. The proposed change does not adversely affect accident initiators or precursors nor does it alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained.

The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not require any physical change to any plant SSCs nor does it require any change in systems or plant operations. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

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

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

Response: No.

There are no hardware changes nor are there any changes in the method which any plant systems perform a safety function. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change.

The proposed change does not introduce any new accident precursors, nor does it involve any physical plant alterations or changes in the methods governing normal plant operation. Also, the change does not impose any new or Start Printed Page 47790different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis.

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

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

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. Evaluation of the 10 CFR Part 21 issue by General Electric determined that the PRFO transient provides additional margin to the Minimum Critical Power Ratio Safety Limit and is not a threat to fuel cladding integrity.

The proposed change to Fuel Integrity Cladding Safety Limits 1.1.A and 1.1.B is consistent with, and within the capabilities of the applicable NRC approved critical power correlations, and thus continues to ensure that valid critical power calculations are performed. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

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

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

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

NRC Acting Branch Chief: Robert Beall.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie Plant, Unit 1, St. Lucie County, Florida

Date of amendment request: May 10, 2013.

Description of amendment request: The amendment will revise the Technical Specifications (TSs) to allow the use of M5® fuel rod cladding material at St. Lucie Plant, Unit 1. The current acceptable fuel rod cladding material is identified in TS 5.3.1, Reactor Core, Fuel Assemblies. The proposed change would revise TS 5.3.1 to add M5® to the approved fuel rod cladding materials and TS 6.9.1.11 to add Framatome (AREVA) topical report BAW-10240(P)(A), Revision 0, “Incorporation of M5® Properties in Framatome ANP Approved Methods,” to the analytical methods used to determine the core operating limits previously reviewed and approved by the NRC.

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 would allow the use of M5® fuel rod cladding in the St. Lucie Unit 1 reactor. The topical report BAW-10240(P)—A prepared by Framatome, currently known as AREVA, has been approved by the NRC for use with M5® fuel cladding. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of M5® fuel cladding, which has essentially the same properties as currently licensed Zircaloy, has been shown to meet all 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident.

The proposed change to Technical Specification 6.9.1.11 (Core Operating Limits Report (COLR)) enables the use of the appropriate methodology to analyze accidents for cores containing fuel with M5® cladding to ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed change to the list of NRC-approved methodologies listed in Technical Specification 6.9.1.11 has no impact on plant operation and configuration. The list of methodologies in Technical Specification 6.9.1.11 does not impact either the initiation of an accident or the mitigation of its consequences.

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

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

Response: No.

Use of M5® clad fuel will not result in changes in the operation or configuration of the facility. The material properties of M5® are similar to those of Zircaloy. Therefore, M5® fuel rod cladding will perform similarly to those fabricated from Zircaloy, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident. The proposed change to Technical Specification 5.3.1, to add M5® as a fuel clad material, does not create any new accident initiators.

The proposed change to the list of NRC-approved methodologies listed in Technical Specification 6.9.1.11, to add BAW-10240(P)—A, has no impact on any plant configuration or system performance. There is no change to the parameters within which the plant is normally operated, and thus the possibility of a new or different type of accident is not created.

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

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

Response: No.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the M5® are not significantly different from those of Zircaloy. The M5® is expected to perform similarly to Zircaloy for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, plant-specific LOCA analyses using M5® properties demonstrate that the acceptance criteria of 10 CFR 50.46 have been satisfied.

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: James Petro, Managing Attorney—Nuclear, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Jessie F. Quichocho.Start Printed Page 47791

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

Date of amendment request: June 6, 2013.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.7.10, “Control Room Ventilation System (CRVS),” and TS 5.6.5, “Core Operating Limits Report (COLR),” to incorporate editorial changes. Specifically, the proposed amendments delete footnote (1), which expired on December 10, 2012, and is no longer applicable, from TS 3.7.10 Condition A Completion Time, and corrects inconsistent wording between TS 5.6.5a.4 and TS 3.2.1, between TS 5.6.5a.5, and TS 3.2.2, and between TS 5.6.5a.9 and TS 3.4.1.

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 editorial changes do not involve any physical changes to structures, systems or components. The proposed editorial change to TS 3.7.10 deletes a footnote that is no longer applicable. The proposed editorial changes to TS 5.6.5 correct administrative discrepancies in the TS to provide consistency with the existing TS Sections 3.2.1, 3.2.2 and 3.4.1.

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

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

Response: No.

The proposed editorial changes to TS 3.7.10 and TS 5.6.5 do not involve an accident.

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

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

Response: No.

The proposed editorial changes to TS 3.7.10 and TS 5.6.5 do not impact accident analyses, fission product barriers, or margin of safety.

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

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

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: April 3, 2013.

Description of amendment request: The proposed amendment would add an exception to Technical Specification 3.0.4 in Technical Specification 3/4.7.6, Control Room Emergency Filtration System (CREFS).

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 adds an exception to the provisions of Specification 3.0.4 in Technical Specification 3/4.7.6, “Control Room Emergency Filtration System (CREFS)” that was previously included in this Technical Specification prior to Amendment 180. The proposed change would allow entry into the applicable Modes of Technical Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying on the actions. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does 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. The proposed change does not alter the Technical Specification Limiting Condition for Operation, Applicability, or remedial Actions that provide for the safe operation of the plant when the Limiting Condition for Operation is not met. The Actions in Technical Specification 3/4.7.6 Action statement b. continue to ensure the safe operation of the plant in the same manner as before. In addition, the proposed change does not affect the Surveillance Requirements of Technical Specification 3/4.7.6. As such, the Surveillance Requirements continue to provide the same level of assurance as before that the CREFS and control room boundary will perform their required safety functions to mitigate the consequences of events within the assumed acceptance limits.

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 adds an exception to the provisions of Specification 3.0.4 in Technical Specification 3/4.7.6, “Control Room Emergency Filtration System (CREFS)” that was previously included in this Technical Specification prior to Amendment 180. The proposed change would allow entry into the applicable Modes of Technical Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying on the actions. The proposed change does not alter the operability requirements or remedial Actions of Technical Specification 3/4.7.6, nor does the change affect the CREFS or control room boundary function during accident conditions. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The change does not alter assumptions made in the applicable safety analyses. As such, the proposed change does not impact the safety analyses assumptions and is consistent with current plant operating practices.

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 adds an exception to the provisions of Specification 3.0.4 in Technical Specification 3/4.7.6, “Control Room Emergency Filtration System (CREFS)” Start Printed Page 47792that was previously included in this Technical Specification prior to Amendment 180. The proposed change would allow entry into the applicable Modes of Technical Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying on the actions. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by the change. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. As such, the CREFS and control room boundary will continue to provide the same level of safety as before.

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, Inc. Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

Date of amendment request: June 19, 2013.

Description of amendment request: The proposed changes would amend Combined License numbers NPF-91 and NPF-92 for Vogtle Electric Generating Plant Units 3 and 4 by departing from the plant-specific design control document Tier 2 and Tier 2* material contained within the updated final safety analysis report (UFSAR) related to the design of structural wall modules used to construct containment internal structures and portions of the auxiliary building. The proposed changes would revise requirements for design spacing of shear studs and the design of structural elements in order to address interferences and obstructions other than wall openings.

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 design function of the containment structural modules is to support the reactor coolant system components and related piping systems and equipment. The design functions of the affected structural modules in the auxiliary building are to provide support and protection for new and spent fuel and the equipment needed to support fuel handling, cooling, and storage in the spent fuel racks, and to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located outside the containment building.

The design function of the shear studs is to enable the concrete and steel faceplates to act in a composite manner and transfer loads into the concrete of the structural modules. The structural modules are seismic Category I structures and are designed for dead, live, thermal, pressure, safe shutdown earthquake loads, and loads due to postulated pipe breaks. The loads and load combinations applicable to the structural modules in the auxiliary building are the same as for the containment internal structures except that there are no design basis accident loadings due to the automatic depressurization system or pressure loads due to pipe breaks. The proposed changes to the UFSAR are to include types of interferences other than wall openings and penetrations that may cause a change in the design spacing of shear studs and the design and spacing of wall module trusses in a local area. The proposed changes clarify that the stud spacing is specified as a design value and add the tolerance for stud spacing. The revised spacing including the tolerance continues to be in conformance with the design and analysis requirements identified in the UFSAR. The proposed changes also include clarification of a requirement for a complete joint penetration weld. The thickness, geometry, and strength of the structures are not adversely altered. The material of the steel plates is not altered. The properties of the concrete included in the structural modules are not altered. As a result, the design function of the containment structural modules is not adversely affected by the proposed change. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor does the change described create any new accident precursors.

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?

Response: No.

The proposed changes to the UFSAR acknowledge types of interferences (other than wall openings and penetrations) that may cause a change in the typical design spacing of shear studs and the design and spacing of wall module trusses in a local area. The proposed changes clarify that the stud spacing is specified as a design value and provide the tolerance for stud spacing. The revised spacing, including the tolerance, continues to be in conformance with the design and analysis requirements identified in the UFSAR. Stud spacing and sizing are evaluated to demonstrate that stud loadings and shear transfer capability are within acceptable limits and that the structural module acts in a composite manner. An additional proposed change is to clarify a requirement for a complete joint penetration weld. The thickness, geometry, and strength of the structures are not adversely altered. The materials of the steel plates are not altered. The properties of the concrete included in the structural modules are not altered. The changes to the internal design of the structural modules do not create any new accident precursors. As a result, the design function of the modules is not adversely affected by the proposed changes.

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?

Response: No.

The criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690 provide a margin of safety to structural failure. The design of the shear studs and wall trusses for the structural wall modules conforms to applicable criteria and requirements in ACI 349 and AISC N690 and, therefore, maintain the margin of Start Printed Page 47793safety. The proposed changes to the UFSAR acknowledge types of interferences (other than wall openings and penetrations) that may cause a change in the typical design spacing of shear studs and the design and spacing of wall module trusses in a local area. The proposed changes clarify that the stud spacing is specified as a design value and add the tolerance for stud spacing. The revised spacing including the tolerance continues to be in conformance with the design and analysis requirements identified in the UFSAR. An additional proposed change is to clarify a requirement for a complete joint penetration weld. There is no change to the capacity of the weld or to the design requirements of the modules. There is no change to the method of evaluation from that used in the design basis calculations.

Therefore, the proposed amendment does not result in 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.

NRC Branch Chief: Lawrence Burkhart.

ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station (ZNPS), Units 1 and 2, Lake County, Illinois

Date of amendment request: June 18, 2012, and supplemented June 5, 2013.

Description of amendment request: The proposed amendments would revise the Physical Security Plan associated with the transfer and storage of spent fuel at the Independent Spent Fuel Storage Installation (ISFSI).

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

Response: No.

The proposed amendment, which incorporates ISFSI security functions, does not reduce the ability of the Security organization to prevent attempts of radiological sabotage and, therefore, does not increase the probability or consequences of a radiological release previously evaluated. The proposed ZNPS ISFSI Physical Security Plan will not affect any important-to-safety systems or components, their mode of operation or operating strategies. The changes have no effect on accident initiators or mitigation.

Therefore, the proposed amendment will 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.

The proposed amendment incorporating ISFSI security functions does not affect the operation of systems that are important-to-safety. The ZNPS ISFSI Physical Security Plan amendment does not affect any of the parameters or conditions that could contribute to the initiation of any accident. No new accident scenarios are created as a result of the ZNPS ISFSI Physical Security Plan. In addition, the design functions of equipment important to safety are not altered as a result of the proposed ZNPS ISFSI Physical Security Plan.

Therefore, the proposed ISFSI Security Plan will not create the possibility of a new or different accident from any previously evaluated.

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

Response: No.

Implementation of the proposed amendment incorporating ISFSI security functions will not reduce a margin of safety as detailed in the Technical Specifications, as there are no Technical Specification requirements associated with the physical security system. Specifically, the proposed ZNPS ISFSI Physical Security Plan does not represent a change in initial conditions, system response time, or any other parameter affecting the course of an accident analysis supporting the Bases of any Technical Specification. The proposed amendment does not reduce the effectiveness of any security/safeguards measures currently in place at the ZNPS.

Therefore, the proposed ZNPS ISFSI Physical Security Plan will 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: Russ Workman, Deputy General Counsel, EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 84101.

NRC Branch Chief: Bruce Watson.

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 Commission'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 Start Printed Page 47794problems 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.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of application for amendment: December 21, 2012.

Brief description of amendment:

The amendment revises Fermi 2 operating license to change its name on the license to “DTE Electric Company.” This name change is purely administrative in nature. Detroit Edison is a wholly owned subsidiary of DTE Energy Company, and this name change is part of a set of name changes of DTE Energy subsidiaries to conform their names to the “DTE” brand name. No other changes are contained within this amendment. This change does not involve a transfer of control over or of an interest in the license for Fermi 2.

Date of issuance: July 12, 2013.

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

Amendment No.: 193.

Facility Operating License No. NPF-43: Amendment revised the operating license.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14131).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 12, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 21, 2013.

Description of amendment request: The amendments revised the divider barrier seal test coupons' tensile strength in Technical Specification Surveillance Requirement 3.6.14.4 from “> 39.7 psi” to “> 39.7 lbs.” This change is an administrative change to correct an error where the wrong units were used when Catawba and McGuire converted to Standard Technical Specifications in 1998 using NUREG-1431, Revision 1.

Date of issuance: July 16, 2013.

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

Amendment Nos.: 270, 266, 270 and 250.

Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: May 14, 2013 (78 FR 28251).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 16, 2013.

No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, Texas

Date of amendment request: July 12, 2012, as supplemented by letter dated October 23, 2012.

Brief description of amendments: The amendments revised Technical Specification (TS) 5.7.1, “High Radiation Areas with Dose Rates not Exceeding 1.0 rem [roentgen equivalent man]/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation,” and 5.7.2, “High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation,” to allow entry into high radiation areas by personnel continuously escorted by individuals qualified in radiation protection procedures and to require a pre-job briefing prior to entry into such areas. In addition, the amendment incorporates an editorial change to TS Table 3.3.3-1, “Post Accident Monitoring Instrumentation.” The typographical error in the title of TS Table 3.3.1-1 column “CONDITION REFERENCED FROM REQUIRED ACTION E.1,” is corrected to read, “CONDITION REFERENCED FROM REQUIRED ACTION D.1,” to reflect that the Required Actions for Condition D of TS 3.3.3, “Post Accident Monitoring (PAM) Instrumentation” are listed in the table.

Date of issuance: July 11, 2013.

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

Amendment Nos.: Unit 1—159; Unit 2—159.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: November 13, 2012 (77 FR 67683).

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

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: September 18, 2012.

Brief description of amendment: The amendment revises the MNGP Technical Specifications (TS) Sections 3.1.6, “Rod Pattern Control,” and 3.3.2.1, “Control Rod Block Instrumentation,” to allow MNGP to reference an optional Banked Position Withdrawal Sequence (BPWS) shutdown sequence in the TS Bases. In addition, a footnote is revised in TS Table 3.3.2.1-1, “Control Rod Block Instrumentation,” to allow operators to bypass the rod worth minimizer if conditions for the optional BPWS shutdown process are satisfied. The changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-476, Revision 1, “Improved BPWS Control Rod Insertion Process (NEDO-33091).”

Date of issuance: July 15, 2013.

Effective date: This license amendment is effective as of the date of its date of issuance and shall be implemented within 180 days after start-up from the 2013 Refueling Outage.

Amendment No.: 173.

Renewed Facility Operating License No. DPR-22: Amendment revises the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: December 11, 2012.

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

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Units 1 and 2, Salem County, New Jersey

Date of application for amendments: July 17, 2012, as supplemented on January 28, 2013, and March 22, 2013.

Brief description of amendments: The amendment revised Salem Nuclear Generating Station Technical Specification 3.7.6.1 (Unit 1) and 3.7.6 (Unit 2), “Control Room Emergency Air Conditioning System,” to eliminate the separate action statements for securing an inoperable Control Area Air Conditioning System and Control Room Start Printed Page 47795Emergency Air Conditioning System isolation damper in the closed position and entering the actions for an inoperable control room envelope boundary.

Date of issuance: July 17, 2013.

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

Amendment Nos.: 304 and 286.

Renewed Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 2, 2013 (78 FR 19754).

The supplemental letter dated March 22, 2013, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 17, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 14, 2012, as supplemented by letters dated February 28, April 19, and June 24, 2013.

Brief description of amendment request: The amendments revised Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR),” to reference and allow use of Westinghouse WCAP-16045-P-A, Addendum 1-A, “Qualification of the NEXUS Nuclear Data Methodology,” (Reference 1 of Enclosure 1) to determine core operating limits. The non-proprietary version is WCAP-16045-NP-A, Addendum 1-A (Reference 2 of Enclosure 1).

Date of issuance: July 17, 2013.

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

Amendment Nos: 191 and 187.

Facility Operating License Nos. NPF-2 and NPF-8: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: October 9, 2012 (77 FR 61440).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 17, 2013.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 29th day of July, 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. 2013-18851 Filed 8-5-13; 8:45 am]

BILLING CODE 7590-01-P