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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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AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

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 to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from May 10, 2016, to May 23, 2016. The last biweekly notice was published on May 24, 2016 (81 FR 32800).

DATES:

Comments must be filed by July 7, 2016. A request for a hearing must be filed by August 8, 2016.

ADDRESSES:

You may submit comments 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-2016-0107. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individuals listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

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FOR FURTHER INFORMATION CONTACT:

Kay Goldstein, Office of Nuclear Reactor Regulation, telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov and Lynn Ronewicz, Office of Nuclear Reactor Regulation, telephone: 301-415-1927, email: Lynn.Ronewicz@nrc.gov. Both are staff of the U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.

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SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0107 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

  • Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0107.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection 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. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section of this document.
  • 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-2016-0107, facility name, unit number(s), application date, and subject in your comment submission.

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 will post all comment submissions at http://www.regulations.gov as well as enter 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 Start Printed Page 36614they 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 into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 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 if 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. If the Commission takes 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. If the Commission makes 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.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

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's 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 within 60 days, 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 set forth 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 with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.

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). If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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 unless the Commission Start Printed Page 36615finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by August 8, 2016. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by August 8, 2016.

Electronic Submissions (E-Filing)

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 ten 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/​getting-started.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 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 NRC'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 public 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: 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 Start Printed Page 36616expedited 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, in some instances, 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.

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: March 22, 2016. A publicly-available version is in ADAMS under Accession No. ML16082A309.

Description of amendment request: The proposed amendment would allow for permanent extension of the Type A primary containment integrated leak rate test interval to 15 years and extension of the Type C test interval up to 75 months. The amendment also proposes two administrative changes to remove text that is no longer applicable. The first change revises technical specification (TS) 5.5.12 to remove a one-time extension of the Type A test frequency. The second change would revise the Fermi 2 Operating License, Section D, to remove a reference to an exemption regarding Appendix J testing of containment air locks.

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

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

Response: No.

The proposed amendment to the TS involves the extension of Fermi 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The current Type A test interval of 10 years would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The primary containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. RG [Regulatory Guide] 1.174 [sic] [ADAMS Accession No. ML023240437] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF [core damage frequency] below 1.0E-06/yr and increases in LERF [large early release frequency] below 1.0E-07/yr. Since the ILRT [integrated leak rate test] does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is very conservatively estimated as 1.27E-08/yr using the EPRI [Electric Power Research Institute] guidance as written. As such, the estimated change in LERF is determined to be “very small” using the acceptance guidelines of RG 1.174.

RG 1.174 also states that when the calculated increase in LERF is in the range of 1.0E-06 per reactor year to 1.0E-07 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.0E-05 per reactor year. An additional assessment of the impact from external events was also made. In this case, the total LERF increase was conservatively estimated (with an external event multiplier of 15) as 1.90E-07 for Fermi 2 (the baseline total LERF for this case is 7.88E- 06/yr). This is well below the RG 1.174 acceptance criteria for total LERF of 1.0E-05.

The change in Type A test frequency to once per 15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 1.14E-4 person-rem/yr (a 0.00184% increase). EPRI Report No. 1009325, Revision 2-A, states that a very small population dose is defined as an increase of ≤1.0 person-rem per year or ≤1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. Moreover, the risk impact when compared to other severe accident risks is negligible.

The increase in the CCFP [conditional containment failure probability] from the three in 10 year [sic] interval to one in 15 year interval is 0.73%. EPRI Report No. 1009325, Revision 2-A, states that increases in CCFP of less than or equal to 1.5 percentage points are very small. Therefore, this increase judged to be very small.

The other two changes, to TS 5.5.12, item a, and Operating License, Provision D, are administrative in nature to remove old text that is no longer applicable.

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 amendment to the TS involves the extension of the Fermi 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (e.g., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

The other two changes to TS 5.5.12, item a, and Operating License, Provision D, are administrative in nature to remove old text that is no longer needed. Therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed amendment to TS 5.5.12 involves the extension of the Fermi 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the Start Printed Page 36617degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed surveillance interval extension is bounded by the 15 year ILRT interval and the 75 month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and Type C testing detects a large percentage of containment leakage paths and the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections preformed in accordance with ASME [American Society of Mechanical Engineers] Section XI, Maintenance Rule, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods, and acceptance criteria for Type A, Type B, and Type C containment leakage tests specified in applicable codes and standards would continue to be met with the acceptance of this proposed change since these are not affected by the changes to the Type A and Type C test intervals.

The other two changes to TS 5.5.12, item a, and Operating License, Provision D, are administrative in nature to remove old text that is no longer needed. Therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.

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: Jon P. Christinidis, DTE Energy, Expert Attorney—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

NRC Branch Chief: David J. Wrona.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: March 24, 2016. A publicly available version is in ADAMS under Accession No. ML16089A228.

Description of amendment request: The amendments would modify Technical Specification 3.6.13, “Ice Condenser Doors,” to revise Condition B for an ice condenser lower inlet door invalid open alarm to preclude plant shutdown caused by an invalid “OPEN” alarm from the “Inlet Door Position Monitoring System.”

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

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

Response: No.

The proposed change will not increase the probability of accident previously evaluated. The Ice Condenser performs an entirely mitigative function. The proposed change does not result in any physical change to the plant which would affect any accident initiators. No structures, systems, or components (SSCs) involved in the initiation of postulated accidents will be operated in any different manner. The probability of occurrence of a previously evaluated accident will not be significantly increased. The proposed change involves use of an alternate method of verifying that the lower inlet doors to the ice condenser are closed. This proposed change has no effect on the ability of the ice condenser to perform its 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 does not alter the design function or operation of any SSC that may be involved in the initiation of an accident. The Ice Condenser will not become the source of a new type of accident. No new accident causal mechanisms will be created. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators.

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 the margin of safety?

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The proposed change involves use of a method to verify the lower inlet doors to the ice condenser are closed when an invalid alarm is providing indication of an open door. This proposed change has no effect on the ability of the ice condenser to perform its function. Hence, the proposed change will not affect containment barriers. Nor does the proposed change have any effect on fuel cladding or the reactor coolant pressure boundary.

Therefore, existing safety margins will be preserved, and 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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendment request: April 13, 2016. A publicly-available version is in ADAMS under Accession No. ML16111B203.

Description of amendment request: The amendments would revise the Allowable Values (AVs) of Surveillance Requirements (SRs) contained in Technical Specification 3.3.8.2, “RPS Electric Power Monitoring,” by amending the Reactor Protection System electric power monitoring assembly AVs for overvoltage and undervoltage contained within SRs 3.3.8.2.2 and 3.3.8.2.3.

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

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

Response: No

The proposed change to the Allowable Values of Surveillance Requirements contained in Technical Specifications 3.3.8.2 does not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SCCs [sic] perform their design function. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The Electrical Protection Assemblies are not accident initiators. They operate in response to off-normal voltage conditions on Class 1E buses to protect the connected loads. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.Start Printed Page 36618

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 change to the Allowable Values of Surveillance Requirements contained in Technical Specifications 3.3.8.2 does not require any modification to the plant (i.e., other than the setpoint changes) or change equipment operation or testing. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators.

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

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

Response: No

The proposed change to the Allowable Values of Surveillance Requirements contained in Technical Specifications 3.3.8.2 does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, Appendix A will continue to be met.

Therefore, the proposed change does not involve any 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: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, M/C DEC45A, Charlotte NC 28202.

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: March 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16088A186.

Description of amendment request: The amendment will revise the Technical Specifications (TSs) to eliminate TS 6.5.8, “Inservice Testing Program.” A new defined term, “Inservice Testing [IST] Program,” will be added to TS 1.0, “Definitions,” section. The licensee has noted that while the request is consistent with TS Task Force (TSTF)-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” there are various deviations from the TSTF-545, Revision 3. ANO-2 TSs are of an older standard version and have not been converted to the improved standard TSs (ISTSs) based on NUREG 1432, “Standard Technical Specifications—Combustion Engineering Plants,” Revision 4. As such, Entergy stated there are several administrative-type variations (TS numbering, wording, etc.) but these variations do not result in any technical conflict with the intent of TSTF-545, Revision 3 or the associated model safety evaluation.

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, with NRC edits in [brackets], 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 TS Chapter 6, “Administrative Controls,” Section 6.5, “Programs and Manuals,” by eliminating the “Inservice Testing Program” specification. Most requirements in the IST Program are removed, as they are duplicative of requirements in the ASME [American Society of Mechanical Engineers] OM Code [ASME Code for Operation and Maintenance of Nuclear Power Plants], as clarified by Code Case OMN-20, “Inservice Test Frequency.” The remaining requirements in the Section 6.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, “Inservice Testing Program,” is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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 does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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 eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS Surveillance Requirement (SR) 4.0.3 (referenced as SR 3.0.3 in the ISTS) allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

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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: Meena K. Khanna.

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

Date of amendment request: March 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16088A181.

Description of amendment request: The amendment would revise the Technical Specifications (TSs) to eliminate TS Section 5.5.8, “Inservice Testing [IST] Program.” A new defined term, “Inservice Testing Program,” will be added to TS 1.1, “Definitions.” This amendment request is consistent with TS Task Force (TSTF)-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” under the consolidated line item improvement process.

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, with NRC edits in [brackets], 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 TS Chapter 5, “Administrative Controls,” Section 5.5, “Programs and Manuals,” by eliminating the “Inservice Testing Program” specification. Most requirements in the IST Program are removed, as they are duplicative of requirements in the ASME [American Society of Mechanical Engineers] OM Code [ASME Code for Operation and Maintenance of Nuclear Power Plants], as clarified by Code Case OMN-20, “Inservice Test Frequency.” The remaining requirements in the Section 5.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, “Inservice Testing Program,” is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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 does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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 eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS Surveillance Requirement (SR) 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or 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 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: Meena K. Khanna.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of amendment request: March 24, 2016, as supplemented by letter dated May 11, 2016. A publicly-available version is in ADAMS under Accession Nos. ML16084A567 and ML16132A440.

Description of amendment request: The amendments would revise the frequency for cycling of the recirculation pump discharge valves as specified in Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.5. Specifically, SR 3.5.1.5 requires verification that each recirculation pump discharge valve cycles through one complete cycle of full travel or is de-energized in the closed position. Currently, this SR needs to be performed once each plant startup prior to exceeding 23 percent rated thermal power (RTP), if the SR had not been performed within the previous 31 days. The amendments would change the frequency for the SR such that it is performed in accordance with the Inservice Testing Program.

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:

Start Printed Page 36620

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 frequency for cycling the recirculation pump discharge valves from “Once each startup prior to exceeding 23% RTP,” as modified by a Note stating, “Not required to be performed if performed within the previous 31 days” to “In accordance with the Inservice Testing Program”. Testing of the recirculation pump discharge valves is not an initiator of any accident previously evaluated. As the recirculation pump discharge valves are still required to be Operable, the ability to mitigate any accident previously evaluated is not affected. The proposed change does not adversely affect 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) from performing their intended function.

Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change revises the frequency for cycling the recirculation pump discharge valves from “Once each startup prior to exceeding 23% RTP,” as modified by a Note stating, “Not required to be performed if performed within the previous 31 days” to “In accordance with the Inservice Testing Program”. This revision will not impact the accident analysis. The change will not alter the methods of operation of the recirculation pump discharge valves. No new or different accidents result. 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 safety analysis.

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

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

Response: No.

The proposed change revises the frequency for cycling the recirculation pump discharge valves from “Once each startup prior to exceeding 23% RTP,” as modified by a Note stating, “Not required to be performed if performed within the previous 31 days” to “In accordance with the Inservice Testing Program.” 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 this change. The proposed change will not result in plant operation in a configuration outside the design basis. The frequency of testing the recirculation pump discharge valves will be consistent with the frequency of testing other valves in the Emergency Core Cooling System.

Therefore, this 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 60555.

NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No.1, DeWitt County, Illinois

Date of amendment request: April 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16095A285.

Description of amendment request: The proposed changes would revise technical specification (TS) limiting condition for operation (LCO) 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than 200 degrees Fahrenheit as a consequence of in-service leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an in-service leak or hydrostatic test, while considering operational conditions to be in Mode 4.

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.

Technical Specifications currently allow for operation at greater than 200 degrees F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated.

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.

Technical Specifications currently allow for operation at greater than 200 degrees F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice

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.

Technical Specifications currently allow for operation at greater than 200 degrees F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an in-service leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. 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: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: G. Ed Miller (Acting)

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

Date of amendment request: April 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16095A275.

Description of amendment request: The amendments would revise the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) Start Printed Page 36621system actuation instrumentation Technical Specification (TS) requirements.

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

Response: No.

The proposed changes involve the addition of clarifying footnotes to the HPCI and RCIC actuation instrumentation TS to reflect the as-built plant design and operability requirements of HPCI and RCIC instrumentation as described in the LGS Updated Final Safety Analysis Report (UFSAR).

HPCI and RCIC are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. In addition, the automatic start of HPCI on high drywell pressure, and the manual initiation of HPCI and RCIC, are not credited to mitigate the consequences of design basis accidents, transients or special events within the current LGS design and licensing basis.

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

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

Response: No.

The proposed changes do not alter the protection system design, create new failure modes, or change any modes of operation. The proposed changes do not involve a physical alteration of the plant, and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident.

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

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

Response: No.

The proposed changes have no adverse effect on plant operation. The plant response to the design basis accidents does not change. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses.

There is no change being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes.

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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Acting Branch Chief: Andrew Hon.

Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: April 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16125A253.

Description of amendment request: The amendments would revise Appendix B (Environmental Protection Plan (EPP)) of the Unit 1 and Unit 2 Operating Licenses to incorporate the revised Section 8.4, “Terms and Conditions” of the currently applicable Biological Opinion issued by the National Marine Fisheries Service (NMFS) on March 24, 2016. In addition, the amendments would clarify in the EPP that the licensee must adhere to the currently applicable Biological Opinion. This clarification would preclude the need for a new license amendment in the event that NMFS issues a new Biological Opinion.

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. Operation of the Facility in Accordance With the Proposed Amendments Would Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The changes are administrative in nature and would in no way affect the initial conditions, assumptions, or conclusions of the St. Lucie Unit 1 or Unit 2 accident analyses. In addition, the proposed changes would not affect the operation or performance of any equipment assumed in the accident analyses. Based on the above information, we conclude that the proposed changes would not significantly increase the probability or consequences of an accident previously evaluated.

2. Use of the Modified Specification Would Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated

The changes are administrative in nature and would in no way impact or alter the configuration or operation of the facilities and would create no new modes of operation. We conclude that the proposed changes would not create the possibility of a new or different kind of accident.

3. Use of the Modified Specification Would Not Involve a Significant Reduction in a Margin of Safety

The changes are administrative in nature and would in no way affect plant or equipment operation or the accident analysis. We conclude that the proposed changes would 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/JB, Juno Beach, FL 33408-0420.

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: April 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16099A097.

Description of amendment request: The proposed amendment would revise technical specification (TS) 3.8.4, “DC Sources—Operating,” Surveillance Requirement (SR) 3.8.4.2 to increase the required 125 Volt (V) Direct Current (DC) subsystems battery charger output current and to remove the second method specified to perform the surveillance. The first proposed change is to increase the required 125 Volt VDC battery charger output current specified as the first option under SR 3.8.4.2 to resolve a non-conservative TS condition. The second proposed change is to remove from SR 3.8.4.2 an alternative option for meeting the surveillance requirement. This alternative requires verifying each battery charger can recharge the battery to the fully charged state within the required time period, 24 hours for the 250 VDC and 8 hours for the 125 VDC subsystems, respectively, while supplying the largest combined continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the Start Printed Page 36622licensee 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 TS changes revise the battery charger surveillance requirements in SR 3.8.4.2. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Updated Safety Analysis Report (USAR). Rather, the DC electrical power system supports operation of equipment used to mitigate accidents. Operation in accordance with the proposed TS continues to ensure that the DC electrical power system is capable of performing its specified safety functions as described in the USAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis.

Accidents are initiated by the malfunction of plant equipment, or the catastrophic failure of plant structures, systems, or components (SSCs). Performance of battery testing is not a precursor to any accident previously evaluated, nor does it change the manner in which the batteries and battery chargers are operated. The proposed testing requirements will not contribute to the failure of the batteries nor any plant SSC. NSPM has determined that the proposed TS changes provide an equivalent level of assurance that the batteries and battery chargers are capable of performing their intended safety functions. Thus, the proposed changes do not affect the probability of an accident previously evaluated.

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 DC electrical power system, including the associated battery chargers, is not an initiator of any accident sequence analyzed in the USAR. The proposed TS changes do not involve operation of the DC electrical power system in a manner or configuration different from those previously evaluated. Performance of battery testing is not a precursor to any accident previously evaluated. NSPM has determined that the proposed TS changes provide an equivalent level of assurance that the batteries and battery chargers are capable of performing their intended safety functions. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed in the safety analyses.

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 equipment design, the 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 do not adversely affect operation of plant equipment. The proposed TS changes do not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment continues to be ensured. The equipment fed by the DC electrical sources will continue to provide adequate power to safety-related loads in accordance with safety analysis assumptions.

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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

South Carolina Electric & 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 7, 2016. A publicly-available version is in ADAMS under Accession No. ML16104A027.

Description of amendment request: The amendment would revise the Emergency Feedwater System pump performance testing requirements in Technical Specification (TS) 3/4.7.1.2, “Emergency Feedwater System,” Surveillance Requirements 4.7.1.2.a.1 and 4.7.1.2.a.2.

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

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

Response: No.

The proposed change deletes an allowed outage time that is no longer applicable and revises the Surveillance Requirements (SRs) that confirm the Emergency Feedwater (EFW) pump performance to be more consistent with the STS [Standard Technical Specifications—Westinghouse Plants]. The change has been determined not to adversely affect the safe operation of the plant. The affected TS requirements are not initiating conditions for any accident previously evaluated. In addition, changes that are consistent with the STS have been previously evaluated by plants adopting the STS and found not to adversely affect the safe operation of Westinghouse NSSS [Nuclear Steam Supply System] plants. Based on the conclusions of the plant specific evaluation associated with the change and the evaluations performed in developing the STS, the proposed change does not result in operating conditions that will significantly increase the probability of initiating an analyzed event. The proposed change was also evaluated to assure that it does not alter the safety analysis assumptions relative to mitigation of an accident or transient event and that the resulting TS requirements continue to ensure the necessary equipment is operable consistent with the safety analyses or that the plant is placed in an operating Mode where the system is no longer required operable. As such the proposed change also does not result in operating conditions that will significantly increase the consequences of an analyzed event.

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

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

Response: No.

The proposed change includes the deletion of an expired allowed outage time extension and the revision of the SRs that confirm the EFW pump performance to be more consistent with the corresponding STS SR. Consistent with the STS SR, the proposed change would remove the specific pump head and flow values from the current SRs and require that the SR be performed in accordance with the Inservice Testing Program. The removal of the specific pump head and flow values from the SR is necessary to support the implementation of a plant modification that would change the current EFW pump head and flow values in the SR. The plant modification is being performed under the provisions of 10CFR50.59. The proposed TS change does not involve a change in the methods governing normal plant operation. The proposed change also does not change any system functions nor does the proposed TS change affect any safety analysis or design basis requirements. The proposed TS change will continue to ensure the EFW System is operable in a similar manner as before. As such, the proposed change does not create new failure modes or mechanisms that are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed changes do [sic] not create the possibility of a new or different kind of accident from any previously evaluated.Start Printed Page 36623

3. Does this [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 proposed change does not physically alter safety-related systems, nor does it affect the way in which safety related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed change. Therefore, in a similar manner as before, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The proposed change results in TS requirements that are consistent with the plant safety analyses. As such, the change does not result in operating conditions that significantly reduce any margin of safety.

Therefore, the proposed changes do [sic] 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: Michael T. Markley.

Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

Date of amendment request: August 11, 2015, as supplemented by letters dated March 16, 2014, and April 4, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15226A276, ML16076A453, and ML16095A373, respectively.

Description of amendment request: The amendments would revise the technical specification (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 Cell Parameters.” A new battery monitoring and maintenance program is being proposed for Section 5.5, “Administrative Controls—Programs and Manuals.”

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 any 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 inoperability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Final Safety Analysis Report (FSAR). 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 updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the FSAR. 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 FSAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis.

The relocation of preventive maintenance surveillances, and certain operating limits and actions, to a licensee-controlled Battery Monitoring and Maintenance Program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance that are consistent with industry standards will continue to be performed. 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 FSAR 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 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 FSAR. 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 updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the FSAR. 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 FSAR.

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 the 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 Monitoring and Maintenance 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 FSAR 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, Inc., 40 Iverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: April 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16120A294.

Description of amendment request: The license proposed three changes to Start Printed Page 36624modifications specified in the March 10, 2015, NFPA [National Environmental Policy Act]-805 amendment, Attachment S, Table S-2, “Plant Modifications Committed.” The three proposed modifications are: (1) Delete Fire Area 1-041 information from Table S-2, (2) add information on item 11, Pyro Panel modification, and, (3) change cable 2VCHAL07P to cable 2VCFARK2P.

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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The licensee's analysis is presented below:

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

Response: No.

The proposed amendment updates Attachments M, S, and W of the previously approved NFPA-805 LAR [license amendment request] submittal for FNP. The attachment revisions are based on the three changes to Table S-2 proposed in this LAR. One of the changes is justified based on negligible risk impact to Core Damage Frequency or Large Early Release Frequency associated with not performing the committed modification. The other two changes have no impact on accident analysis as they are clarifying or administrative in nature.

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not increase the probability or consequence of an accident as verified by the risk analysis performed.

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

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 amendment updates Attachments M, S, and W of the previously approved NFPA-805 LAR submittal for FNP. The attachment revisions are based on the three changes to Table S-2 proposed in this LAR. One of the changes is justified based on negligible risk impact to Core Damage Frequency or Large Early Release Frequency associated with not performing the committed modification. The other two changes have no impact on accident analysis as they are clarifying or administrative in nature. The proposed change relates to the availability of fire PRA [probabilistic risk analysis] credited component in given fire scenarios.

Therefore, this proposed change 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 proposed amendment updates Attachments M, S, and W of the previously approved NFPA-805 LAR submittal for FNP. The attachment revisions are based on the three changes to Table S-2 proposed in this LAR. One of the changes is justified based on negligible risk impact to Core Damage Frequency or Large Early Release Frequency associated with not performing the committed modification. The other two changes have no impact on accident analysis as they are clarifying or administrative in nature.

The proposed change does not increase the probability or consequence of an accident and does not reduce the margin of safety as verified by the risk analysis performed.

Therefore, this 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.

NRC Branch Chief: Michael T. Markley.

III. 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 can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 29, 2015.

Brief description of amendment: The amendment approved a change to the Waterford Steam Electric Station, Unit 3, Cyber Security Plan Implementation Schedule Milestone 8 full implementation date and a related change to the existing operating license physical protection license condition.

Date of issuance: May 10, 2016.

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

Amendment No.: 247. A publicly-available version is in ADAMS under Accession No. ML16077A270; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-38: The amendment revised the facility operating license.

Date of initial notice in Federal Register : September 1, 2015 (80 FR 52805).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 10, 2016.

No significant hazards consideration comments received: No.Start Printed Page 36625

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: August 18, 2015, as supplemented by letter dated April 14, 2016.

Brief description of amendments: The amendments revised the reactor steam dome pressure specified in the technical specification safety limits.

Date of issuance: May 11, 2016.

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

Amendment Nos.: 209, 250, 243, 262, and 257. A publicly-available versions is in ADAMS under Accession No. ML16111A104. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. : NPF-62, DPR-19, DPR-25, DPR-29, and DPR-30. Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : October 27, 2015 (80 FR 65812). The supplemental letter dated April 14, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a safety evaluation dated May 11, 2016.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 16, 2015.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.4.1, “RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to delete current Tables 3.4.1-1, “Reduction in Percent RATED THERMAL POWER for Reduced RCS Flow Rate, Unit 1,” and 3.4.1-2, “Reduction in Percent RATED THERMAL POWER for Reduced RCS Flow Rate, Unit 2,” and add RCS thermal design flow (TDF) values to the requirements of TS 3.4.1. The change also relocates the RCS minimum measured flow (MMF) values to the DCPP, Units 1 and 2, core operating limits reports (COLR) with a reference to the MMF values in TS 3.4.1 and Surveillance Requirements 3.4.1.3 and 3.4.1.4. Figure 2.1.1-1, “Reactor Core Safety Limit,” has been revised to delete a footnote with references to Tables 3.4.1-1 and 3.4.1-2. The change is consistent with NUREG-1431, Volume 1, Revision 4.0, “Standard Technical Specifications, Westinghouse Plants,” April 2012; NRC-approved Technical Specification Task Force (TSTF) Change Traveler 339-A, Revision 2, “Relocate TS Parameters to COLR,” dated June 13, 2000; and NRC-approved WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report,” January 1999.

The change is necessary to correct a non-conservative TS 3.4.1 total RCS flow rate value for DCPP, Unit 1. The change also ensures that the TS stays conservative, if the cycle-specific minimum RCS flow is higher than the minimum TDF.

Date of issuance: May 19, 2016.

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

Amendment Nos.: Unit 1—226; Unit 2—228. A publicly-available version is in ADAMS under Accession No. ML16117A252; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and TSs.

Date of initial notice in Federal Register : November 10, 2015 (80 FR 69714).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 19, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 27, 2014, as supplemented by letters dated October 31, 2014; February 12, May 12, September 10, and November 5, 2015; and January 14 and March 4, 2016.

Brief description of amendment: The amendment approved a change to the Virgil C. Summer Nuclear Station licensing basis to incorporate a supplemental analysis for the steam generator tube rupture accident.

Date of issuance: May 16, 2016.

Effective date: As of the date of issuance and shall be implemented within120 days of issuance.

Amendment No.: 205. A publicly-available version is in ADAMS under Accession No. ML15231A605; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-12: Amendment revised the Facility Operating License.

Date of initial notice in Federal Register : October 14, 2014 (79 FR 61661). The supplemental letters dated October 31, 2014; February 12, May 12, September 10, and November 5, 2015; and January 14 and March 4, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 16, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 31, 2015, as supplemented by letters dated January 28, 2016, and March 11, 2016.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.4.14, “RCS Pressure Isolation Valve (PIV) Leakage,” to eliminate the requirements for the residual heat removal system suction valve auto closure interlock function.

Date of issuance: May 17, 2016.

Effective date: As of the date of issuance and shall be implemented as follows: Unit 1—prior to the first entry into Mode 4, following the end-of-cycle refueling outage 27 (scheduled for fall 2016), and Unit 2—prior to the first entry into Mode 4, following the end-of-cycle refueling outage 25 (scheduled for fall 2017).

Amendment Nos.: 201 (Unit 1) and 197 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16083A265; documents related Start Printed Page 36626to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register : October 27, 2015 (80 FR 65815). The supplemental letters dated January 28, 2016, and March 11, 2016, 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 May 17, 2016.

No significant hazards consideration comments received: No.

Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station (SSES), Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: October 27, 2014, as supplemented by letters dated July 2, 2015; September 21, 2015; November 11, 2015; and January 29, 2016.

Brief description of amendments: The amendments modified the SSES technical specifications (TSs). Specifically, the amendments modified the TSs by relocating specific surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program, with implementation of Nuclear Energy Institute (NEI) 04-10, Revision 1, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.” The changes are consistent with NRC-approved Technical Specification Task Force Improved Standard Technical Specifications Change Traveler (TSTF)-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF Initiative 5b.” The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of this TSTF improvement and included a model no significant hazards consideration and safety evaluation (SE).

This license amendment request was submitted by PPL Susquehanna, LLC; however, on June 1, 2015, the NRC staff issued an amendment changing the name on the SSES license from PPL Susquehanna, LLC to Susquehanna Nuclear, LLC (ADAMS Accession No. ML15054A066). These amendments were issued subsequent to an order issued on April 10, 2015, to SSES, approving an indirect license transfer of the SSES license to Talen Energy Corporation (ADAMS Accession No. ML15058A073).

Date of issuance: May 20, 2016.

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

Amendment Nos.: 266 (Unit 1) and 247 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16005A234; documents related to these amendments are listed in the SE enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-14 and NPF-22: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register : March 3, 2015 (80 FR 11479). The supplemental letters dated July 2, 2015; September 21, 2015; November 11, 2015; and January 29, 2016, 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 an SE dated May 20, 2016.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 27th day of May, 2016.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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

End Signature End Supplemental Information

[FR Doc. 2016-13255 Filed 6-6-16; 8:45 am]

BILLING CODE 7590-01-P