ML16257A517

Biweekly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: September 27, 2016

ML16257A517
Person / Time
Site:	Hatch, Calvert Cliffs, Nine Mile Point, Palisades, Palo Verde, Indian Point, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Columbia, Vogtle, Duane Arnold, Farley, Cook, FitzPatrick
Issue date:	09/19/2016
From:	Boland A Division of Operating Reactor Licensing
To:
Burkhardt J
References
09/27/2016; 81FR66301, NRC-2016-0202
Download: ML16257A517 (49)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2016-0202]

Biweekly Notice Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations 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 August 30, 2016, to September 12, 2016. The last biweekly notice was published on September 13, 2016.

DATES: Comments must be filed by October 27, 2016. A request for a hearing must be filed by November 28, 2016.

2 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-2016-0202. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; e-mail: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.

• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN 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.

FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:

301-415-1384, e-mail: Janet.Burkhardt@nrc.gov.

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to NRC-2016-0202, facility name, unit number(s), plant docket number, application date, and subject 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:

3

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0202.

• NRCs 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 NRCs Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by e-mail 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 this document.

• NRCs PDR: You may examine and purchase copies of public documents at the NRCs PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments Please include Docket ID NRC-2016-0202 facility name, unit number(s), plant docket number, 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 they do not want to be publicly disclosed in their comment submission. Your request should

4 state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

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 Commissions 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

5 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 persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition to intervene (petition) with respect to issuance of the amendment to the subject facility operating license or combined license. Petitions shall be filed in accordance with the Commissions Agency Rules of Practice and Procedure in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the NRCs PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRCs regulations are accessible electronically from the NRC Library on the NRCs Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition 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 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 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

6 permitted with particular reference to the following general requirements: (1) the name, address, and telephone number of the petitioner; (2) the nature of the petitioners right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioners 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 petitioners interest. The petition must also set forth the specific contentions which the 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 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion to support its position on the issue. 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 petitioner to relief.

A 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 persons admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with the NRCs 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

7 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 finds 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 petitioners interest in the proceeding. The petition should be submitted to the Commission by November 28, 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 set forth in this section, except that under 10 CFR 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).

8 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. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

B. 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 (hereinafter petition), and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRCs E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail 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 petition (even in instances in

9 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 NRCs public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing the E-Submittal server are available on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRCs E-Filing system does not support unlisted software, and the NRC Electronic Filing Help Desk will not be able to offer assistance in using unlisted software.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a petition. Submissions should be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRCs public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the documents are submitted through the NRCs 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 e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRCs 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 petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

10 A person filing electronically using the NRCs adjudicatory E-Filing system may seek assistance by contacting the NRC Electronic Filing Help Desk through the Contact Us link located on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals.html, by e-mail to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 7 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 stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) first class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRCs 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

11 home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a petition 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.

The Commission will issue a notice or order granting or denying a hearing request or intervention petition, designating the issues for any hearing that will be held and designating the Presiding Officer. A notice granting a hearing will be published in the Federal Register and served on the parties to the hearing.

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 NRCs PDR. For additional direction on accessing information related to this document, see the Obtaining Information and Submitting Comments section of this document.

12 Energy Northwest, Docket No. 50-397, Columbia Generating Station (Columbia), Benton County, Washington Date of amendment request: July 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16196A419.

Description of amendment request: The amendment would change Technical Specification (TS) 5.5.6, Inservice Testing [IST] Program, to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, Inservice Test Frequency. This change, thereby, will then adopt Technical Specification Task Force (TSTF) TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing.

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

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 Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, Inservice Test Frequency, which has been approved for use at Columbia. 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

13 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 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

14 (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 licensees 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: William A. Horin, Esq., Winston & Strawn, 1700 K Street, N.W.,

Washington, D.C. 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

Energy Northwest, Docket No. 50-397, Columbia Generating Station (Columbia), Benton County, Washington Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A528.

Description of amendment request: The amendment would revise the current Columbia Emergency Plan Emergency Action Level scheme to one based on Nuclear Energy Institute (NEI) guidance established in NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6, which has been endorsed by the NRC.

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

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

Response: No.

The proposed amendment affects the Columbia Generating Station (Columbia) Emergency Plan (EP) and associated Emergency Action Levels (EALs); it does not alter the Operating License or the Technical Specifications. The proposed amendment does not change the design function of any system, structure, or component and does not change the way the plant is maintained or operated. The proposed amendment does not affect any accident mitigating feature or increase the likelihood of malfunction for plant structures, systems, and components.

The proposed amendment will not change any of the analyses associated with the Columbia Final Safety Analysis Report Chapter 15 accidents because plant operation, structures, systems, components, accident initiators, and accident mitigation functions remain unchanged.

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

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

Response: No.

The proposed amendment affects the Columbia EP and associated EALs; it does not change the design function of any system, structure, or component and does not change the way the plant is operated or maintained. The proposed amendment does not create a credible failure mechanism, malfunction, or accident initiator not already considered in the design and licensing basis.

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

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

Response: No.

Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public.

The proposed amendment does not impact operation of the plant and no accident analyses are affected by the proposed amendment. The

16 proposed amendment does not affect the Technical Specifications or the method of operating the plant. Additionally, the proposed amendment will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this amendment. The proposed amendment will not result in plant operation in a configuration outside the design basis. The proposed amendment does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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

The NRC staff has reviewed the licensees 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: William A. Horin, Esq., Winston & Strawn, 1700 K Street, N.W.,

Washington, D.C. 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3 (IP3), Westchester County, New York Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York Date of amendment request: August 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16230A308.

Description of amendment request: The amendment would transfer the beneficial interest in the Power Authority of the State of New York (PASNY) Master Decommissioning Trust (Master Trust), including all rights and obligations thereunder, held by PASNY for IP3 and FitzPatrick to Entergy Nuclear Operations, Inc. (ENO). ENO also requests the NRCs consent to

17 amendments to the Master Decommissioning Trust Agreement dated July 25, 1990, as amended (Master Trust Agreement), governing the Master Trust to facilitate this transfer.

Finally, ENO seeks approval of license amendments to modify the existing trust-related license conditions to reflect the proposed transfer of the Master Trust to ENO and to delete other conditions so as to apply the requirements of 10 CFR 50.75(h)(1). ENO and Exelon Generation Company, LLC. (Exelon), jointly filed an application for a direct license transfer of FitzPatrick to Exelon on August 18, 2016. A separate Federal Register notice details the NRCs consideration of approval for the FitzPatrick license transfer.

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

Response: No.

The requested changes delete certain license conditions pertaining to the decommissioning trust agreements currently in sections 2.Q to 2.X of the IP3 Operating License and sections 2.H to 2.O of the FitzPatrick Operating License. In addition, conforming changes to 2.W and 2.X of the IP3 Operating License and 2.P and 2.Q of the FitzPatrick Operating License are necessary [to] reflect the transfer of the Master Trust from PASNY to ENO.

The requested changes are consistent with the types of license amendments permitted in 10 CFR 50.75(h)(5).

The regulations of 10 CFR 50.75(h)(4) state that Unless otherwise determined by the Commission with regard to a specific application, the Commission has determined that any amendment to the license of a utilization facility that does no more than delete specific license conditions relating to the terms and conditions of decommissioning trust agreements involves no significant hazards consideration.

In addition the requested changes seek changes to the Master Trust agreement only to the extent that they replace PASNY, a non-licensee, with ENO, a licensee. No other changes to the Master Trust agreement are contemplated.

18 This request involves changes that are administrative in nature. No actual plant equipment or accident analyses will be affected by the proposed changes.

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

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

Response: No.

This request involves administrative changes to licenses that will be consistent with the NRCs regulations at 10 CFR 50.75(h) and to change the name of the entity responsible under the Master Trust for decommissioning from a non-licensee to a licensee.

No actual plant equipment or accident analyses will be affected by the proposed changes and no failure modes not bounded by previously evaluated accidents will be created.

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

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

Response: No.

The request involves administrative changes to the licenses that will be consistent with the NRCs regulations at 10 CFR 50.75(h) and to change the name of the entity responsible under the Master Trust for decommissioning from a non-licensee to a licensee.

Margin of safety is associated with confidence in the ability of the fission product barriers to limit the level of radiation doses to the public. No actual plant equipment or accident analyses will be affected by the proposed change. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety systems settings, or will not relax the bases for any limiting conditions of operation.

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

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

19 proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jeanne Cho, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, New York, 10601.

NRC Branch Chief: Travis L. Tate.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP),

Van Buren County, Michigan Date of amendment request: August 22, 2016, as supplemented by letter dated September 8, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16235A195 and ML16252A351, respectively.

Description of amendment request: The proposed amendment would replace existing license condition 2.C.(4) with a new license condition to state that technical specification (TS) surveillance requirement (SR) 3.1.4.3 is not required for control rod drive 13 (CRD-13) during cycle 25 until the next entry into Mode 3. In addition, the condition would state that CRD-13 seal leakage shall be repaired prior to entering Mode 2, following the next Mode 3 entry, and that the reactor shall be shut down if CRD-13 seal leakage exceeds two gallons per minute.

The proposed amendment also requests replacement of the obsolete note in TS SR 3.1.4.3 with a note to clarify that TS SR 3.1.4.3 is not required to be performed or met for CRD-13 during cycle 25 provided CRD-13 is administratively declared immovable, but trippable, and Condition D is entered for CRD-13.

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:

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

Response: No.

The proposed license amendment replaces an obsolete license condition concerning CRD-22 testing that applied only to operating cycle 21 with a new license condition to forgo the remaining two required surveillance tests of CRD-13 from the PNP TS surveillance requirement for partial movement every 92 days during cycle 25. Since CRD-13 remains trippable, the proposed license condition does not affect or create any accident initiators or precursors. As such, the proposed license condition does not increase the probability of an accident.

The proposed license amendment does not increase the consequences of an accident. The ability to move a full-length control rod by its drive mechanism is not an initial assumption used in the safety analyses. The safety analyses assume full-length control rod insertion, except the most reactive rod, upon reactor trip. The surveillance requirement performed during the last refueling outage verified control rod drop times are within accident analysis assumptions. ENO [Entergy Nuclear Operations] has determined that CRD seal leakage does not increase the likelihood of an untrippable control rod. The assumptions of the safety analyses will be maintained, and the consequences of an accident will not be increased.

Therefore, operation of the facility in accordance with the proposed license condition would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed license amendment does not involve a physical alteration of any structure, system or component (SSC) or change the way any SSC is operated. The proposed license condition does not involve operation of any required SSCs in a manner or configuration differently from those previously recognized or evaluated. No new failure mechanisms would be introduced by the requested SR interval extension.

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

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

Response: No.

21 The proposed license amendment does not affect trippability of the control rod. It will have the same capability to mitigate an accident as it had prior to the proposed license condition.

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

The NRC staff has reviewed the licensees 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: Jeanne Cho, Senior Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey; and Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1 (NMP1), Oswego County, New York Date of amendment request: August 1, 2016. A publicly-available version is in ADAMS under Accession No. ML16215A128.

Description of amendment request: The amendments would revise OCNGSs Technical Specification (TS) Section 2.1, Safety Limit - Fuel Cladding Integrity, and NMP1s TS Section 2.1.1, Fuel Cladding Integrity, to reduce the steam dome pressure.

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 edits in [brackets]:

22

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

Response: No.

The proposed change to the OCNGS TS for the reactor steam dome pressure in Reactor Core Safety Limits 2.1.A and 2.1.B does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. Additionally, the proposed change to NMP1 for the reactor steam dome pressure in Reactor Core Safety Limits 2.1.1.a and 2.1.1.b does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed change is in accordance with an NRC approved critical power correlation methodology, and as such, maintains required safety margins. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained.

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

Lowering the value of reactor steam dome pressure in the TS has no physical effect on plant equipment and therefore, no impact on the course of plant transients. The change is an analytical exercise to demonstrate the applicability of correlations and methodologies. There are no known operational or safety benefits.

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 reduction in the reactor dome pressure safety limit from 800 psia [pounds per square inch absolute] to 700 psia is a change based upon previously approved documents and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced. There are no hardware changes nor are there any changes in the method by which any plant systems perform

23 a safety function. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change.

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

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 margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. Evaluation of the 10 CFR Part 21 condition by GE [General Electric] determined that since the MCPR [minimum critical power ratio] improves during the PRFO

[pressure regulator failure-maximum demand (open)] transient, there is no decrease in the safety margin and therefore there is not a threat to fuel cladding integrity. The proposed change in reactor dome pressure supports the current safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged.

The proposed change to Reactor Core Safety Limits 2.1.A and 2.1.B is consistent with and within the capabilities of the applicable NRC approved critical power correlation for the fuel designs in use at OCNGS.

Additionally, the proposed change to Reactor Core Safety Limits 2.1.1.a and 2.1.1.b is consistent with and within the capabilities of the NRC approved critical power correlation for the fuel designs in use at NMP1.

No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

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

24 The NRC staff has reviewed the licensees 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: Shaun M. Anderson.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: July 21, 2016. A publicly-available version is in ADAMS under Accession No. ML16208A076.

Description of amendment request: The proposed changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-545, Revision 3, TS

[Technical Specification] Inservice Testing [IST] Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing. The proposed change would revise the TSs to eliminate the Section 5.5.6, Inservice Testing Program. A new defined term, INSERVICE TESTING PROGRAM, would be added to the TS Definitions section. TS SRs that currently refer to the Inservice Testing Program from Section 5.5.6 would be revised to refer to the new defined term, 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:

25

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 [Operation and Maintenance] Code, as clarified by Code Case OMN-20, Inservice Test Frequency. The remaining requirements in the Section 5.5.6 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 IST 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 IST 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.

26 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 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 TS 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 licensees 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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: David J. Wrona.

27 Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: August 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16224B122.

Description of amendment request: The amendment request proposes changes to plant-specific Tier 2 information incorporated into the Updated Final Safety Analysis Report (UFSAR),

and involves changes to combined license Appendix C (and corresponding plant-specific Tier 1 information). The proposed changes are to information identifying the frontal face area and screen surface area for the In-Containment Refueling Water Storage Tank (IRWST) screens, the location and dimensions of the protective plate located above the containment recirculation (CR) screens, and increasing the maximum Normal Residual Heat Removal System flowrate through the IRWST and CR screens. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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 the NRC staffs edits in square brackets:

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

Response: No.

The proposed changes to the location and dimensions of the protective plate continues to provide sufficient space surrounding the containment recirculation screens for debris to settle before reaching the screens as confirmed by an evaluation demonstrating that the protective plate continues to fulfill its design function of preventing debris from reaching the screens. In addition, the increase to the minimum IRWST screen size reinforces the ability of the screens to perform their design function with

28 the increased [Residual Heat Removal System (RNS)] maximum flowrate proposed. The proposed changes do not adversely affect any accident initiating component, and thus the probabilities of the accidents previously evaluated are not affected. The affected equipment does not adversely affect the ability of equipment to contain radioactive material. Because the proposed change does not affect a release path or increase the expected dose rates, the potential radiological releases in the UFSAR accident analyses are unaffected.

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

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

Response: No.

The proposed activity to change the location and dimensions of the protective plate above the containment recirculation screens, to change the minimum IRWST screen size, and to increase the maximum RNS flowrate through the IRWST and CR screens does not alter the method in which safety functions are accomplished. The analyses demonstrate that the screens are able to perform their functions in a similar manner and perform adequately in response to an accident, and no new failure modes are introduced by the proposed change.

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

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

Response: No.

The proposed change to the design does not change any of the codes or standards to which the IRWST screens, containment recirculation screens, and containment recirculation screen protective plate are designed as documented in the UFSAR. The containment recirculation screen protective plate continues to prevent debris from reaching the CR screens, and the IRWST and CR screens maintain their ability to block debris while at the proposed increase in RNS maximum flowrate.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes.

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

29 The NRC staff has reviewed the licensees 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: August 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16236A265.

Description of amendment request: The amendment request proposes changes to the Fire Pump Head and Diesel Fuel Day Tank. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Companys AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).

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

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

Response: No.

The increase in head pressure by the proposed change to the fire protection system (FPS) motor-driven and diesel-driven fire pumps

30 maintains compliance with National Fire Protection Association (NFPA)

Standard NFPA-14, Standard for the Installation of Standpipe, Private Hydrants, and Hose Systems, 2000 Edition, requirements by providing adequate pressure in the standpipe and automatic sprinkler system to maintain the ability to fight and/or contain a postulated fire. The proposed change to the diesel-driven fire pump fuel day tank volume maintains the availability of the diesel-driven fire pump for service upon failure of the electric motor-driven fire pump or a loss of offsite power by providing a fuel day tank that is reserved exclusively for the diesel-driven pump and meets the minimum capacity requirements of NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, 1999 Edition. These changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and

[components (SSCs)] accident initiator or initiating sequence of events.

These changes have no adverse impact on the support, design, or operation of mechanical and fluid systems. The response of systems to postulated accident conditions is not adversely affected by the proposed changes. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. Consequently, the plant response to previously evaluated accidents is not impacted, nor does the proposed change create any new accident precursors.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes to the fire pump performance specifications and fire pump fuel day tank volume do not affect any safety-related equipment, nor do they add any new interface to safety-related SSCs. No system or design function or equipment qualification is affected by this change. The changes do not introduce a new failure mode, malfunction, or sequence of events that could affect safety or safety-related equipment.

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

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

31 Response: No.

The proposed changes maintain compliance with the applicable Codes and Standards, thereby maintaining the margin of safety associated with these SSCs. The proposed changes do not alter any applicable design codes, code compliance, design function, or safety analysis.

Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change, thus the margin of safety is not reduced.

Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by these changes, no margin of safety is reduced.

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

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

Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16214A252.

Description of amendment request: The amendments would revise the technical specifications (TSs) at the Edwin I. Hatch Nuclear Plant, Units 1 and 2, to eliminate the lnservice Testing Program from TS 5.5, Programs and Manuals, and add a new defined term, INSERVICE

32 TESTING PROGRAM, to TS 1.1, Definitions. This request is submitted in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing.

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

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 lnservice Testing Program specification. Most requirements in the lnservice Testing Program are removed, as they are duplicative of requirements in the ASME OM [American Society of Mechanical Engineers Operation and Maintenance] Code, as clarified by Code Case OMN- 20, lnservice Test Frequency. The remaining requirements in the Section 5.5 IST

[Inservice Testing] 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. lnservice 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.

33 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 accident 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 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.

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

The NRC staff has reviewed the licensees 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 Inverness 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, Units 1 and 2, Houston County, Alabama Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16214A252.

Description of amendment request: The amendments would revise the technical specifications (TSs) at the Joseph M. Farley Nuclear Plant, Units 1 and 2, to eliminate the lnservice Testing Program from TS 5.5, Programs and Manuals, and add a new defined term, INSERVICE TESTING PROGRAM, to TS 1.1, Definitions. This request is submitted in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing.

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

35

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 lnservice Testing Program specification. Most requirements in the lnservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code [American Society of Mechanical Engineers Operation and Maintenance Code], as clarified by Code Case OMN-20, lnservice Test Frequency. The remaining requirements in the Section 5.5 IST [Inservice Testing] 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. lnservice 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

36 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 accident 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 SR 3.0.3 allowance to defer performance of missed in service 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 licensees 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.

37 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. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16214A252.

Description of amendment request: The amendments would revise the technical specifications (TSs) at the Vogtle Electric Generating Plant, Units 1 and 2, to eliminate the lnservice Testing Program from the TS 5.5, Programs and Manuals, section and to add a new defined term, INSERVICE TESTING PROGRAM, to the TS 1.1, Definitions, section. This request is submitted in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.

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

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 lnservice Testing Program specification. Most requirements in the lnservice Testing Program are removed, as they are duplicative of requirements in the ASME OM [American Society of Mechanical Engineers Operation and

38 Maintenance] Code, as clarified by Code Case OMN-20, lnservice Test Frequency. The remaining requirements in the Section 5.5 IST

[Inservice Testing] 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. lnservice 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 accident previously evaluated.

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

Response: No.

39 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 SR 3.0.3 allowance to defer performance of missed in service 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 licensees 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 Inverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

40 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 Commissions rules and regulations. The Commission has made appropriate findings as required by the Act and the Commissions 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 Commissions 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.

41 Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona Date of application for amendments: October 9, 2015, as supplemented by letter dated May 12, 2016.

Brief description of amendments: The amendments approve a revision to the emergency action levels from a scheme based on Nuclear Energy Institute (NEI) 99-01, Revision 5, Methodology for Development of Emergency Action Levels, to a scheme provided in the subsequent Revision 6 of NEI 99-01.

Date of issuance: September 8, 2016.

Effective date: As of the date of issuance and shall be implemented within 365 days from the date of issuance.

Amendment Nos.: Unit 1 - 198; Unit 2 - 198; Unit 3 - 198. A publicly-available version is in ADAMS under Accession No. ML16180A109; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Operating Licenses.

Date of initial notice in Federal Register: December 8, 2015 (80 FR 76318). The supplemental letter dated May 12, 2016, provided additional information that clarified the application, incorporated recent emergency preparedness frequently asked questions, 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 Commissions related evaluation of the amendments is contained in a Safety Evaluation dated September 8, 2016.

No significant hazards consideration comments received: No.

42 Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of amendment request: September 24, 2013, as supplemented by letters dated February 9, March 11, April 13, July 6, and August 13, 2015; and February 24 and April 22, 2016.

Brief description of amendments: These amendments modify the operating licenses and technical specifications (TSs) to incorporate a new fire protection licensing basis in accordance with 10 CFR 50.48(c). The amendments authorize the transition of the licensees fire protection program to a risk-informed, performance-based program based on the 2001 Edition of National Fire Protection Association Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.

Date of issuance: August 30, 2016.

Effective date: As of the date of issuance and shall be implemented in accordance with the schedule contained in the revised paragraph 2.E. and page 12 of Appendix C, Additional Conditions to the Renewed Facility Operating Licenses.

Amendment Nos.: 318 and 296. A publicly-available version is in ADAMS under Accession No. ML16175A359; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: August 5, 2014 (79 FR 45488). The supplemental letters dated February 9, March 11, April 13, July 6, and August 13, 2015; and February 24 and April 22, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original

43 proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 30, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York Date of amendment request: October 8, 2015, as supplemented by letter dated April 7, 2016.

Brief description of amendment: The amendments modified the technical specifications (TSs) to allow for brief, inadvertent simultaneous opening of redundant secondary containment personnel access doors during brief entry and exit conditions.

Date of issuance: August 31, 2016.

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

Amendment Nos.: 223 (Unit 1) and 157 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16197A486; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-63 and NPF-69: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: January 5, 2016 (81 FR 262). The supplemental letter dated April 7, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

44 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 31, 2016.

No significant hazards consideration comments received: No.

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: August 31, 2015, as supplemented by letters dated April 20 and July 15, 2016.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) consistent with Technical Specification Task Force Traveler 422, Revision 2, Change in Technical Specifications End States (CE NPSD-1186).

Date of issuance: August 30, 2016.

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

Amendment Nos.: 234 and 184. A publicly-available version is in ADAMS under Accession No. ML16210A374; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: November 24, 2015 (80 FR 73237). The supplemental letters dated April 20 and July 15, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

45 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 30, 2016.

No significant hazards consideration comments received: No.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC),

Linn County, Iowa Date of amendment request: May 18, 2016.

Brief description of amendment: The amendment revised the DAEC technical specifications (TSs) Section 2.1.1, Reactor Core [Safety Limits], to change the Safety Limit Minimum Critical Power Ratio (SLMCPR) for two recirculation loop operation and for single recirculation loop operation. The changes reflected the cycle-specific analysis. The amendment also removed an outdated historical footnote from TS Table 3.3.5.1-1.

Date of issuance: September 12, 2016.

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

Amendment No.: 297. A publicly-available version is in ADAMS under Accession No. ML16211A514; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 5, 2016 (81 FR 43665).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated September 12, 2016.

No significant hazards consideration comments received: No.

46 NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: August 18, 2015, as supplemented by letters dated January 29, April 14, and May 31, 2016.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.12, Primary Containment Leakage Rate Testing Program, to state that the program shall be in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J.

Date of issuance: August 30, 2016.

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

Amendment No.: 296. A publicly-available version is in ADAMS under Accession No. ML16210A008; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 2015 (80 FR 65814). The supplemental letters dated January 29, April 14, and May 31, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 30, 2016.

No significant hazards consideration comments received: No.

47 NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: June 26, 2013, as supplemented by letters dated September 16, 2013, July 29, August 28, September 25, November 14, December 19, 2014; January 16, May 12, August 26, 2015; and February 22, April 7, and May 3, 2016.

Brief description of amendments: The amendments authorized the transition of the Point Beach fire protection program to a risk-informed, performance-based program based on National Fire Protection Association Standard 805 (NFPA 805), Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, in accordance with 10 CFR 50.48(c).

Date of issuance: September 8, 2016.

Effective date: As of the date of issuance and shall be implemented as described in the Transition License Conditions.

Amendment Nos.: 256 and 260. A publicly-available version is in ADAMS under Accession No. ML16196A093; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 28580). The supplemental letters dated September 16, 2013, July 29, August 28, September 25, November 14, December 19, 2014; January 16, May 12, August 26, 2015; and February 22, April 7, and May 3, 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.

48 The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated September 8, 2016.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee Date of amendment request: December 15, 2015, as supplemented by letters dated May 4, 2016, and June 1, 2016.

Brief description of amendment: The amendment revised the Technical Specifications to allow implementation of the F* (F-star) alternate repair criterion for steam generator tubes.

Date of issuance: September 6, 2016.

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

Amendment No.: 2. A publicly-available version is in ADAMS under Accession No. ML16203A365; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-96: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 16, 2016 (81 FR 7844). The supplemental letters dated May 4, 2016, and June 1, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

49 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated September 6, 2016.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 19th day of September 2016.

For the Nuclear Regulatory Commission.

/RA/

Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.