ML16271A092

Biweekly Federal Register Notice - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: October 11, 2016

ML16271A092
Person / Time
Site:	Hatch, Monticello, Browns Ferry, Mcguire, Nine Mile Point, Catawba, Harris, Saint Lucie, Point Beach, Summer, Seabrook, Brunswick, Surry, Turkey Point, Vogtle, Duane Arnold, Robinson, Clinton, FitzPatrick, McGuire
Issue date:	09/28/2016
From:	Boland A Division of Operating Reactor Licensing
To:
Goldstien K, NRR/DORL/LPL1-1, 415-1506
References
NRC-2016-0207
Download: ML16271A092 (59)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2016-0207]

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 September 13, 2016 to September 26, 2016. The last biweekly notice was published on September 27, 2016.

DATES: Comments must be filed by November 10, 2016. A request for a hearing must be filed by December 12, 2016.

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

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0207. 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: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:

301-415-1506, e-mail: Kay.Goldstein@nrc.gov.

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0207, 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:

• Federal rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0207.

• 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-2016-0207, facility name, unit number(s), application date, and subject in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

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 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 the action. 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 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 proceeding. 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 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) through (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 December 12, 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).

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 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.

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 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.

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, New Hill, North Carolina Date of amendment request: May 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16151A001.

Description of amendment request: The amendment would revise the Shearon Harris Nuclear Power Plant, Unit 1, technical specifications (TSs) to institute a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, the specifics of which will be contained in a licensee-controlled document. It also relocates to this program the current TS surveillance requirements (SRs) for evaluating diesel fuel oil, along with the SRs for the draining, sediment removal, and cleaning of each main fuel oil storage tank at least once every 10 years. In addition, an exception is proposed to Regulatory Guide (RG) 1.137, Revision 1, Fuel Oil Systems for Standby Diesel Generators, for the allowance of performing sampling of new fuel oil offsite prior to its addition to the fuel oil storage tanks.

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

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

Response: No.

The proposed amendment institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program. The specifics of this program will be contained in a licensee-controlled document. The current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks will be relocated to this program. The American Society for Testing and Materials (ASTM) standard references pertaining to new and stored fuel oil will be relocated to the aforementioned program; however, requirements to perform testing in accordance with applicable ASTM standards are retained in the TS. Requirements to perform surveillances of both new and stored diesel fuel oil are also retained in the TS.

Evaluations of future changes to the licensee-controlled document will be conducted pursuant to the requirements of 10 CFR 50.59. A more rigorous testing of water and sediment content is added to the clear and bright test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. These changes will not affect nor degrade the ability of the emergency diesel generators (DGs) to perform their specified safety functions as the diesel fuel oil continues to be properly evaluated.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems or components from performing their intended function to mitigate the consequences on an initiating event with the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure.

Therefore, the proposed changes do 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 institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, of which the current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks are relocated, including pertinent ASTM standard references. A more rigorous testing of water and sediment content is added to the clear and bright test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. These changes do not alter the way any structure, system, or component functions and does not modify the manner in which the plant is operated. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs.

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

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

Response: No.

The proposed amendment institutes a new administrative program TS for the establishment, implementation, and maintenance of a Diesel Fuel Oil Testing Program, the specifics of which will be contained in a licensee-controlled document. The current TS SR for evaluating new and stored diesel fuel oil and the cleaning of the fuel oil storage tanks will be relocated to this program, along with the pertinent ASTM standard references. Changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59, thereby providing an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there is no significant reduction in a margin of safety.

A more rigorous testing of water and sediment content is added to the clear and bright test used to establish the acceptability of new fuel oil for use prior to its addition to the fuel oil storage tanks. Additionally, an exception to RG 1.137 is proposed to allow for the performance of new fuel oil sampling offsite. The margin of safety provided by the DGs is unaffected by the proposed changes since there continue to be TS requirements to ensure fuel oil is of the appropriate quality and reliability for emergency DG use. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies, while maintaining sufficient controls to preserve the current margins of safety.

Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of no significant hazards consideration is justified.

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: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Business Services, 550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Acting Branch Chief: Jeanne A. Dion.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16242A332.

Description of amendment request: The amendment would revise technical specification (TS) 5.5.6, Primary Containment Leak Rate Testing Program. These revisions would extend the Type A Primary Containment Integrated Leak Rate Test interval to 15 years and extend the Type C Local Leak Rate Test testing interval up to 75 months.

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

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

Response: No.

The proposed amendment to the TS involves the extension of the JAF

[James A. FitzPatrick Nuclear Power Plant] Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0087 person rem/year. EPRI [Electric Power Research Institute] Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of 1.0 person-rem per year, or 1 % of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of the risk assessment for this amendment meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The JAF Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and; (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers] Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

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

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

Response: No.

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

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is operated.

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

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

Response: No.

The proposed amendment to TS 5.5.6 involves the extension of the JAF Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

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

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for JAF. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any margin of 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. 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: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No.1, DeWitt County, Illinois Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A300.

Description of amendment request: The proposed changes supports changes to the organization, staffing, and training requirements contained in Section 5.0 of the technical specifications (TSs) after the license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor pressure vessel.

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

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

Response: No.

The proposed changes would not take effect until CPS has permanently ceased operation and entered a permanently defueled condition. The proposed changes would revise the CPS TS by deleting or modifying certain portions of the TS administrative controls described in Section 5.0 of the TS that are no longer applicable to a permanently shutdown and defueled facility.

The proposed changes do not involve any physical changes to plant structures, systems, and components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting control settings, limiting conditions for operation, surveillance requirements, or design features.

The deletion and modification of provisions of the facility administrative controls do not affect the design of SSCs necessary for safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the Spent Fuel Pool (SFP). The proposed changes are administrative in nature and do not affect any accidents applicable to the safe management of spent irradiated fuel or the permanently shutdown and defueled condition of the reactor.

In a permanently defueled condition, the only credible accidents are the Fuel Handling Accident (FHA), Postulated Radioactive Releases Due to Liquid Radwaste Tank Failures, and Cask Drop Accident. Other accidents such as Loss of Coolant Accident, Loss of Feedwater, and Reactivity and Power Distribution Anomalies will no longer be applicable to a permanently defueled reactor plant.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a permanently defueled condition will be the only operation allowed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

The proposed changes in the administrative controls do not affect the ability to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations. The proposed changes narrow the focus of nuclear safety concerns to those associated with safely maintaining spent nuclear fuel. These changes remove the implication that CPS can return to operation once the final certification required by 10 CFR 50.82(a)(1)(ii) is submitted to the NRC.

Any event involving safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the SFP would evolve slowly enough that no immediate response would be required to protect the health and safety of the public or station personnel. Adequate communications capability is provided to allow facility personnel to safely manage storage and handling of irradiated fuel. As a result, no changes to radiological release parameters are involved. There is no effect on the type or amount of radiation released, and there is no effect on predicted offsite doses in the event of an accident.

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

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

Response: No.

The proposed changes to delete and/or modify certain TS administrative controls have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The proposed changes do not result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled and CPS will no longer be authorized to operate the reactor.

The proposed changes will continue to require proper control and monitoring of safety significant parameters and activities. The proposed changes do not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers in support of maintaining the plant in a permanently shutdown and defueled condition (e.g., fuel cladding and SFP cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

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

Response: No.

The proposed changes involve deleting and/or modifying certain TS administrative controls once the CPS facility has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for CPS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). As a result, the occurrence of certain design basis postulated accidents are no longer considered credible when the reactor is permanently defueled. The only remaining credible accidents are the FHA, the Postulated Radioactive Releases Due to Liquid Radwaste Tank Failures, and the Cask Drop Accident. The FHA is the limiting Chapter 15 dose event for CPS in its decommissioned state.

The proposed changes do not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA. The proposed changes are limited to those portions of the TS administrative controls that are not related to the safe storage and maintenance of spent irradiated fuel.

These proposed changes do not directly involve any physical equipment limits or parameters. The requirements that are proposed to be revised and/or deleted from the CPS TS are not credited in the existing accident analysis for the remaining applicable postulated accidents; therefore, they do not contribute to the margin of safety associated with the accident analysis. Certain postulated DBAs [design-basis accidents] involving the reactor are no longer possible because the reactor will be permanently shut down and defueled and CPS will no longer be authorized to operate the reactor.

Therefore, the proposed changes do 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: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear,.

4300 Winfield Road, Warrenville, IL 60555.

Acting NRC Branch Chief: G. Edward Miller.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa 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 NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire 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 Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16214A276.

Description of amendment request: The amendments would revise the Technical Specifications (TS) consistent with Technical Specifications Task Force Traveler 545, Revision 3, TS Inservice Testing [IST] Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing (ADAMS Accession No. ML15294A555).

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 eliminates Technical Specifications (TS) Section 5.5.6 and 5.5.7, Inservice Testing Program, for Duane Arnold and Point Beach, respectively, and eliminates TS Section 6.8.4.i, Inservice Testing Program for St. Lucie Units 1 and 2. The proposed change eliminates the requirements regarding [IST] from TS 4.0.5 in the Seabrook and Turkey Point TS. Most requirements in the [IST] 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, Inservice Test Frequency. The remaining requirements related to the 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 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 [IST] 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.

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 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 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 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 Blair, Managing Attorney - Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

Acting NRC Branch Chief: Jeanne A. Dion.

Northern States Power Company - Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: July 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A030.

Description of amendment request: The proposed amendment would eliminate technical specification (TS), Section 5.5.5, Inservice Testing [IST] Program, to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, Inservice Test Frequency.

A new defined term, Inservice Testing Program, is added to TS Section 1.1, Definitions. The proposed change to the TS is consistent with TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [surveillance requirement] Usage Rule Application to Section 5.5 Testing. TS SRs that currently refer to the IST 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:

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 OM Code, as clarified by Code Case OMN-20, Inservice Test Frequency. The remaining requirements in the Section 5.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to the regulations. A new defined term, Inservice Testing Program, is added to the TS, which references the requirements of 10 CFR 50.55a(f).

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

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

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

Response: No.

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

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

3. Does the proposed change involve a significant reduction 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 frequency, and will instead 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc.,

414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos.52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: August 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16225A437.

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 (RNS) 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:

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 the increased 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 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.

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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue, NW, Washington, DC, 20004-2514.

NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos.52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South Carolina Date of amendment request: September 8, 2016. A publicly-available version is in ADAMS under Accession No. ML16252A200.

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 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?

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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.

NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No.

50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16243A463.

Description of amendment request: The amendment would remove the administrative controls associated with the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.5.4, Refueling Water Storage Tank.

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

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

Response: No.

The proposed change removes an administrative note added by Amendment No. 192. The administrative control applied by Amendment No. 192 was issued to prevent or reduce the risk for drainage of the Reactor Water Storage Tank (RWST) when aligned to the non-safety, non-seismic purification system. The station has implemented a modification that qualifies the interconnection of the RWST to the purification system. The installed design prevents the RWST being drained below the current Technical Specifications minimum volume requirement due to a failure in the non-safety purification system. The RWST will continue to perform its safety function and the overall system performance has not been affected [by] this proposed amendment.

Assumptions previously made in evaluating the consequences of the accident are not altered, and the consequences of the accident are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation.

Therefore, the proposed change has insignificant impact on the probability and consequences of an accident previously evaluated. A combination of design and administrative controls ensure that the Purification Loop maintains RWST boron concentration and water volume requirements whenever the contents of the RWST are processed through the system. The RWST is operated under System Operating Procedure for the Spent Fuel Cooling System and is protected by maintaining the isolation valve for the lower return line locked closed in modes 1 through 4.

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 introduce a new or different accident previously evaluated. The station implemented a qualified design that prevents the RWST from being drained below the current TS 3.5.4.a minimum volume requirement. The proposed change does not alter the design requirements of the RWST or any Structure, System or Component or its function during accident conditions. The changes do not alter assumptions made in the safety analysis and the current TS LCO are maintained. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation. The proposed change removes a note added by Amendment No. 192 that applied an administrative control to manage the risk of a postulated RWST drainage scenario by the purification system.

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 removes a note added by Amendment No. 192.

The proposed change does not alter the safety limits, limiting safety system settings or limiting conditions for operation of the RWST. The modification preserved the current licensing and design bases of the RWST, therefore the margin of safety for the RWST are not affected. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The Purification Loop supports the Spent Fuel System and is not credited for safe shutdown of the plant or accident mitigation.

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: Kathryn M. Sutton, Morgan, Lewis & Bockius LLP, 1111 Pennsylvania Avenue, NW, Washington, DC 20004.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC); Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant (HNP), Unit No. 2, Appling County, Georgia Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16245A257.

Description of amendment request: The amendment would revise the values for the reactor core Safety Limit 2.1.1.2 for Minimum Critical Power Ratios for both single and dual recirculation loop operation.

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

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

Response: No.]

The Safety Limit Minimum Critical Power Ratio (SLMCPR) ensures that 99.9% of the fuel rods in the core will not be susceptible to boiling transition during normal operation or the most limiting postulated design-basis transient event. The new SLMCPR values preserve the existing margin to the onset of transition boiling; therefore, the probability of fuel damage is not increased as a result of this proposed change. The determination of the revised HNP Unit 2 SLMCPRs has been performed using NRC-approved methods of evaluation. These plant-specific calculations are performed each operating cycle and may require changes for future cycles. The revised SLMCPR values do not change the method of operating the plant; therefore, they have no effect on the probability of an accident, initiating event, or transient:

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

Response: No.]

The proposed changes result only from a specific analysis for the HNP Unit 2 core reload design. These changes do not involve any new or different methods for operating the facility. No new initiating events or transients result from these changes.

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

Response: No.]

The new SLMCPRs have been calculated using NRC-approved methods of evaluation with plant and cycle-specific input values for the fuel and core design for the upcoming cycle of operation. The SLMCPR values ensure that 99.9% of the fuel rods in the core will not be susceptible to boiling transition during normal operation or the most limiting postulated design-basis transient event. The operating MCPR limit is set appropriately above the safety limit value to ensure adequate margin when the cycle-specific transients are evaluated. Accordingly, the margin of safety is maintained with the revised values.

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, 40 Iverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc., Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: July 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16211A436.

Description of amendment request: The amendment request proposes to add to License Condition 2.D.(1) of the VEGP Units 3 and 4 combined licenses an Interim Amendment Request process for changes during construction when emergent conditions are present.

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 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 amendment would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs [combined licenses] to allow construction to continue, at SNC's [Southern Nuclear Operating Company] own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted.

The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which 1) identifies the proposed change; 2) evaluates whether emergent conditions are present; 3) evaluates whether the change would result in any material decrease in safety; and 4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue a determination that construction could continue pending SNCs initiation of the COL-ISG-025 PAR [preliminary amendment request] / LAR [license amendment request] process. The requirement to include a Nuclear Construction Safety Assessment ensures that the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. If the continued construction would result a material decrease in safety, then continued construction would not be authorized.

The proposed amendment does not modify the design, construction, or operation of any plant structures, systems, or components (SSCs), nor does it change any procedures or method of control for any SSCs. Because the proposed amendment does not change the design, construction, or operation of any SSCs, it does not adversely affect any design function as described in the Updated Final Safety Analysis Report.

The proposed amendment does not affect the probability of an accident previously evaluated. Similarly, because the proposed amendment does not alter the design or operation of the nuclear plant or any plant SSCs, the proposed amendment does not represent a change to the radiological effects of an accident, and therefore, does not involve an increase in the consequences of an accident previously evaluated.

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 would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow construction to continue, at SNC's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which 1) identifies the proposed change; 2) evaluates whether emergent conditions are present;

3) evaluates whether the change would result in any material decrease in safety; and 4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would NRC issue a determination that construction could continue pending SNC's initiation of the COL-ISG-025 PAR/LAR process.

The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment only adds a new screening process and does not change the design, construction, or operation of the nuclear plant or any plant operations.

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

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

Response: No.

The proposed amendment would add an Interim Amendment Request process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow construction to continue, at SNC's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SNC to submit a Nuclear Construction Safety Assessment which 1) identifies the proposed change; 2) evaluates whether emergent conditions are present;

3) evaluates whether the change would result in any material decrease in safety; and 4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue determination that construction could continue pending SNC's initiation of the COL-ISG-025 PAR/LAR process.

The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment does not alter any design function or safety analysis.

Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed amendment, thus the margin of safety is not reduced. The only impact of this activity is the addition of an Interim Amendment Request process.

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: 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.52-025 and 52-026, Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia Date of amendment request: September 9, 2016. A publicly-available version is in ADAMS under Accession No. ML16253A412.

Description of amendment request: The amendment request proposes changes to update the Protection and Safety Monitoring System (PMS) design, specifically the description of the roles of the Qualified Data Processing System (QDPS) and the safety displays. The proposed changes add Main Control Room (MCR) safety-related display divisions A and D to plant-specific Tier 1 (and associated COL Appendix C) and the Updated Final Safety Analysis Report (UFSAR), and correct the name of the QDPS in the UFSAR by referring to the QDPS as a system, rather than a subsystem. Because, this proposed change requires a departure from Tier 1 information in 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 proposed change to the roles of the qualified data processing system (QDPS) and safety-related displays, as well as the change to add Division A and Division D of the main control room (MCR) safety-related displays to the listing of PMS equipment, as identified in Combined License (COL) Appendix C (and plant-specific Tier 1) Table 2.5.2-1 and Updated Final Safety Analysis Report (UFSAR) Table 3.11-1 and 3l.6-2 do not alter any accident initiating component/system failure or event, thus the probabilities of the accidents previously evaluated are not affected.

The proposed changes do not adversely affect safety-related equipment or a radioactive material barrier, and this activity dos not involve the containment of radioactive material.

The radioactive material source terms and release paths used in the safety analysis are unchanged, thus the radiological releases in the UFSAR accident analysis are not affected.

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 change to the roles of the QDPS and safety-related displays, as well as the change to add Division A and Division D of the MCR safety-related displays to the listing of PMS equipment, as identified in COL Appendix C (and plant-specific Tier 1) Table 2.5.2-1 and UFSAR Table 3.11-1 and 3l.6-2 does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not alter the design or capability of any sensors which provide input to the QDPS. The functionality of the QDPS to process the input obtained from sensors into data to be sent to the safety displays is not affected by the proposed changes. The proposed changes do not affect any functions performed by the safety displays, nor do the proposed changes affect the capability of the safety displays to display the data received from the QDPS.

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.

There is no safety-related structure, system or component (SSC) or function adversely affected by the proposed change to the roles of the QDPS and safety-related displays, nor by the change to add Division A and Division D of the MCR safety-related displays to the listing of Protection and Safety Monitoring System (PMS) equipment. The proposed changes do not alter the mechanisms by which system components are actuated or controlled. Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed 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.52-025 and 52-026, Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia Date of amendment request: September 9, 2016. A publicly-available version is in ADAMS under Accession No. ML16253A204.

Description of amendment request: The amendment request proposes changes to revise plant-specific Tier 1, plant-specific Tier 2, and combined license (COL) Appendix C information concerning the details of the Class 1E direct current and uninterruptible power supply system (IDS), specifically adding seven Class 1E fuse panels to the IDS design. These proposed changes provide electrical isolation between the non-Class 1E IDS battery monitors and their respective Class 1E battery banks. 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 with NRC staff 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 revise plant-specific Tier 1, COL Appendix C, and

[Updated Final Safety Analysis Report (UFSAR)] information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect the design functions of the IDS, including the Class 1E battery banks and the battery monitors.

These proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits as described in the current licensing basis do not have an adverse effect on any of the design functions of any plant systems. The proposed changes do not adversely affect any plant electrical system and do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors.

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

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

Response: No.

The proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect any plant electrical system and do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors.

No design function described in the UFSAR is adversely affected by the proposed changes.

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

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

Response: No.

There is no safety-related [structure, system, and component (SSC)] or function adversely affected by the proposed change to add IDS fuse isolation panels to non-Class 1E IDS battery monitors and Class 1E IDS circuits. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes and no margin or safety is reduced.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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.52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: September 13, 2016. A publicly-available version is in ADAMS under.

Accession No. ML16257A711.

Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* information. The proposed departure consists of changes to Tier 2* information in the UFSAR to change the provided minimum reinforcement area in the column line 7.3 wall from elevation 82-6 to elevation 100-0.

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.

As indicated in the UFSAR Subsection 3H.5.1.2, the wall at column line 7.3 is a shear wall that connects the shield building and the nuclear island exterior wall at column line I. Deviations were identified in the constructed wall from the design requirements. The wall was repaired in accordance with American Concrete Institute (ACI) 349-01. This change impacts UFSAR Table 3H.5-5. For the south face of the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel for wall section 11 for the vertical reinforcement from the wall segment of elevation 82-6 to 100-0 is decreased from 3.12 in2/ft to 3.08 in2/ft. The change of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building for postulated loads. The criteria and requirements of ACI 349-01 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349-01 and therefore maintains the margin of safety. This change does not involve any accident initiating components or events, thus leaving the probabilities of an accident unaltered. The reduced margin does not adversely affect any safety-related structures or equipment nor does the reduced margin reduce the effectiveness of a radioactive material barrier. Thus, the proposed change would not affect any safety-related accident mitigating function served by the containment internal structures.

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 reduction of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building. As demonstrated by the continued conformance to the applicable codes and standards governing the design of the structures, the wall withstands the same effects as previously evaluated. There is no change to the design function of the wall, and no new failure mechanisms are identified as the same types of accidents are presented to the wall before and after the 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 of the provided versus required vertical reinforcing steel, identified in UFSAR Table 3H.5-5, is not a significant reduction in the margin of safety. For the south face of the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel for wall section 11 for the vertical reinforcement from the wall segment of elevation 82-6 to 100[-0 is decreased from 3.12 in2/ft to 3.08 in2/ft.

The change of the provided versus required vertical reinforcing steel does not change the performance of the affected portion of the auxiliary building for postulated loads. The criteria and requirements of ACI 349-01 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349-01 and therefore maintains the margin of safety. The reduction in margin does not alter any design function, design analysis, or safety analysis input or result, and sufficient margin exists to justify departure from the Tier 2* requirements for the wall. As such, because the system continues to respond to design basis accidents in the same manner as before without any changes to the expected response of the structure, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes. Accordingly, no significant safety margin is reduced by the change.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama Date of amendment request: August 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16225A663.

Description of amendment request: The amendments would modify the Technical Specifications (TSs) for Units 1, 2, and 3 by revising TS 4.3.1.2, Fuel Storage Criticality, to preclude the placement of fuel in the new fuel storage vaults. This TS change would remove the existing TS 4.3.1.2 criticality criteria wording in its entirety, and replaces it with language that specifically restricts the placement of fuel in the new fuel storage vaults.

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

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

Response: No.

The proposed amendment does not change the fuel handling processes, the fuel handling equipment, or require alteration of the plant fuel storage systems. The amendment places a restriction on use of the new fuel storage vaults, requiring that new fuel be placed only in the spent fuel pool racks. Because no changes to fuel handling equipment, fuel storage systems, or fuel handling processes are involved, the proposed amendment does not increase the probability or consequences of a fuel handling accident.

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

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 modification to the Technical Specifications does not require changes to the plant hardware or alter the operating characteristics of any plant system. As a result, no new failure modes are being introduced. Therefore, the change does not introduce a new or different kind of accident from those previously evaluated.

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

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

Response: No.

The proposed change to TS 4.3.1.2 ensures that the criticality margins of safety for fuel storage are maintained, by excluding the new fuel storage vault as an approved fuel storage location. The change restricts the storage of new fuel to the spent fuel pool racks, which are fully analyzed from a criticality standpoint. The change does not physically alter the fuel storage systems, or modify fuel storage requirements in such a way as to degrade the margins of criticality 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: Sherry A. Quirk, General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.

NRC Acting Branch Chief: Jeanne A. Dion.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia Date of amendment request: May 10, 2016. A publicly-available version is in ADAMS under Accession No. ML16134A069.

Description of amendment request: The amendments would extend the Surry Power Station, Unit Nos. 1 and 2, Technical Specification 3.2, Chemical and Volume Control System, paragraph E requirements for primary grade water (PG) lockout from being applicable in Refueling Shutdown and Cold Shutdown to being applicable in Refueling Shutdown, Cold Shutdown, Intermediate Shutdown, and Hot Shutdown (except during the approach to critical and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following reactor shutdown from reactor critical or power operation).

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 conservatively imposes additional operational controls on the highest capacity flow path of PG to the Reactor Coolant System (RCS). These controls are currently credited in the boron dilution analysis in Refueling Shutdown and Cold Shutdown modes. The proposed change extends these controls into Intermediate and Hot Shutdown modes. As such, the change will provide defense against rapid reactivity insertions due to boron dilution events and reduce the probability of boron dilution events. The proposed change will have no impact on normal operating plant releases and will not increase the predicted radiological consequences of accidents postulated in the UFSAR [Updated Final Safety Analysis Report]. The proposed change makes no physical modifications and does not change plant design.

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

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

Response: No.

The proposed change is an extension of existing operational controls on PG flow to the RCS to include additional operating modes. The change precludes high flow rate boron dilutions in Intermediate and Hot Shutdown modes similar to the current TS requirement in Refueling and Cold Shutdown modes. It does not affect the operation of the emergency boration function of the Chemical and Volume Control System (CVCS).

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

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

Response: No.

The proposed change provides defense against rapid reactivity insertions to potential boron dilution events in shutdown operating modes and reduces the probability of boron dilution events. As such, it increases the margin of safety for the boron dilution event.

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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc.,

120 Tredegar St., RS-2, Richmond, VA 23219.

NRC Branch Chief: Michael T. Markley.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the 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.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: January 18, 2016, as supplemented by letter dated June 20, 2016.

Brief description of amendments: The amendments revised Technical Specification (TS) 5.5.2, Containment Leakage Rate Testing Program, to allow (1) an increase in the existing Type A Integrated Leakage Rate Testing Program test interval from 10 years to 15 years, in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A; (2) adoption of an extension of the containment isolation valve leakage testing (Type C) frequency from the 60 months currently permitted by 10 CFR part 50, appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A; (3) adoption of the use of American National Standards Institute/American Nuclear Society (ANSI/ANS)-56.8-2002, Containment System Leakage Testing Requirements; and (4) adoption of a more conservative grace interval of 9 months for Type A, Type B, and Type C leakage tests, in accordance with NEI 94-01, Revision 3-A.

The amendments also made the following administrative changes: (1) deletion of the information regarding the performance of containment visual inspections as required by Regulatory Position C.3, as the containment inspections are addressed in TS Surveillance Requirement 3.6.1.1, and (2) deletion of the information regarding the performance of the next Catawba Nuclear Station, Unit 1, Type A test no later than November 13, 2015, and the next Catawba Nuclear Station, Unit 2, Type A test no later than February 6, 2008, as both Type A tests have already occurred.

Date of issuance: September 12, 2016.

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

Amendment Nos.: 286 (Unit 1) and 282 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16229A113; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: March 15, 2016 (81 FR 13839). The supplemental letter dated June 20, 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 amendments is contained in a Safety Evaluation dated September 12, 2016.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: February 18, 2016, as supplemented by letter dated June 30, 2016.

Brief description of amendment: The amendments modified Technical Specification (TS) 5.5.2, Containment Leakage Rate Testing Program, for a one-time extension to the 10-year frequency of the integrated leakage rate test (ILRT) or Type A test. This revision extends the period from 10 years to 10.5 years between successive tests, changing the performance of the next ILRT from fall 2017 to spring 2019 for Unit 1 and from spring 2017 to fall 2018 for Unit 2.

Date of issuance: September 26, 2016.

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

Amendment Nos.: 290 (Unit 1) and 269 (Unit 2). A publicly available version is in ADAMS under Accession No. ML16236A053; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: May 10, 2016 (81 FR 28894). The supplemental letter dated June 30, 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 amendments is contained in a Safety Evaluation dated September 26, 2016.

No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina Duke Energy Progress, Inc., Docket No. 50-261; H. B. Robinson Steam Electric Plant Unit No. 2 (RNP), Darlington County, South Carolina Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, (HNP), Wake and Chatham Counties, North Carolina Date of amendment request: February 1, 2016.

Description of amendment request: The amendments revised the licensees name from Duke Energy Progress, Inc. to Duke Energy Progress, LLC.

Date of issuance: September 13, 2016.

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

Amendment Nos.: 271 and 299 (BSEP); 152 (HNP); 246 (RNP). A publicly-available version is in ADAMS under Accession No. ML16217A118; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-71, DPR-62 (BSEP), NPF-63 (HNP), and NFP-23 (RNP): Amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: April 12, 2016 (81 FR 21596).

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated September 13, 2016.

No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant (HNP),

Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: October 29, 2015, as supplemented by letters dated, February 16, 2016, August 8 and 26, 2016, and September 8 and 16, 2016.

Brief description of amendment: The amendment revised Technical Specifications to allow the A Emergency Service Water (ESW) pump to be inoperable for 14 days to allow for the replacement of the A Train ESW pump. The amendment is applicable on a one-time basis.

Date of issuance: September 16, 2016.

Effective date: As of the date of issuance and shall be implemented by October 29, 2016.

Amendment No.: 153. A publicly-available version is in ADAMS under Accession No. ML16253A059; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.

Renewed Facility Operating License No. NPF-63: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 5, 2016 (81 FR 260). The supplemental letters dated February 16, 2016, August 8 and 26, 2016, and September 8 and 16, 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 an SE dated September 16, 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 application for amendment: March 18, 2016.

Brief description of amendment: The amendments revised the technical specifications (TSs) on a change to the method of calculating core reactivity for the purpose of performing the Reactivity Anomalies surveillance.

Date of issuance: September 15, 2016.

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

Amendment Nos.: Unit 1 - 224 and Unit 2 - 158. A publicly-available version is in ADAMS under Accession No. ML16188A029; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License No. DPR-63 and NPF-69: The amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: May 10, 2016 (81 FR 28897).

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated September 15, 2016.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant (PNPP), Unit No. 1, Lake County, Ohio Date of amendment request: October 29, 2015, as supplemented by letter dated April 22, 2016.

Brief description of amendment: The amendment revised the PNPP emergency action level (EAL) scheme to one based on the Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors.

Date of issuance: September 14, 2016.

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

Amendment No.: 173. A publicly-available version is in ADAMS under Accession No. ML16158A331; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-58: The amendment revised the Facility Operating License to authorize revision to the PNPP emergency plan.

Date of initial notice in Federal Register: December 22, 2015 (80 FR 79620). The supplemental letter dated 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 staff's 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 September 14, 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: January 19, 2016, as supplemented by a letter dated May 6, 2016.

Brief description of amendments: The amendments revised the Operating Licenses licensing basis to allow elimination of the end-of-cycle moderator temperature coefficient (MTC) surveillance test as supported by NRC-Approved Topical Report CE NPSD-91 1-A and Amendment 1-A, Analysis of Moderator Temperature Coefficients in Support of a Change in the Technical Specification End of Cycle Negative MTC Limit, and St. Lucie specific supporting information. The amendments also add NRC-approved Westinghouse PARAGON Topical Report WCAP-16045-P-A, Revision 0, Qualification of the Two-Dimensional Transport Code PARAGON, to the Technical Specification list of Core Operating Limits Report methodologies.

Date of issuance: September 19, 2016.

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

Amendment Nos.: 235 and 185. A publicly-available version is in ADAMS under Accession No. ML16183A138; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments.

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

Date of initial notice in Federal Register: March 29, 2016 (81 FR 17506). The supplemental letter dated May 6, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in an SE dated September 19, 2016.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 28th day of September 2016.

For the Nuclear Regulatory Commission.

/RA/

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