ML19225B632

Biweekly Federal Register Notice - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: 08/27/2019

ML19225B632
Person / Time
Issue date:	08/23/2019
From:	Gregory Suber Division of Operating Reactor Licensing
To:
Goldstein C, 415-1506
References
NRC-2019-0168
Download: ML19225B632 (59)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2019-0168]

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 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 July 30, 2019 to August 12, 2019. The last biweekly notice was published on August 13, 2019.

DATES: Comments must be filed by September 30, 2019. A request for a hearing must be filed by October 28, 2019.

ADDRESSES: You may submit comments by any of the following methods:

• Federal Rulemaking Web Site: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0168. Address questions about NRC docket IDs in

Regulations.gov to Jennifer Borges; telephone: 301-287-9127; e-mail:

Jennifer.Borges@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.

• Mail comments to: Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.

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.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2019-0168, 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 https://www.regulations.gov/ and search for Docket ID NRC-2019-0168.

• NRCs Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select Begin Web-based ADAMS Search. For problems with ADAMS, please 2

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-2019-0168, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

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

II. Background Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, 3

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.

III. 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 section 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination.

Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment 4

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 petition for leave 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. The NRCs regulations are accessible electronically from the NRC Library on the NRCs Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRCs Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (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 5

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.

In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in 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 must 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 the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must 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 the requirements at 10 CFR 2.309(f) 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. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that partys admitted contentions, including the opportunity to present evidence, consistent with the NRCs regulations, policies, and procedures.

Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 6

10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the Electronic Submissions (E-Filing) section of this document.

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 establish 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 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 the 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 no later than 60 days from the date of publication of this notice. 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.

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Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at 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 his or her 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 and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate 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. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

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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 (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (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. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must 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 document is 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 document on 9

those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.

A person filing electronically using the NRCs adjudicatory E-Filing system may seek assistance by contacting the NRCs Electronic Filing Help Desk through the Contact Us link located on the NRCs public Web site at https://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 6 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, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents 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 10

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 https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click Cancel when the link requests certificates and you will be automatically directed to the NRCs electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

For further details with respect to these license amendment application(s), 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.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois 11

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York Date of amendment request: June 25, 2019. A publicly-available version is in ADAMS under Accession No. ML19176A498.

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Description of amendment request: The amendments would revise instrument testing and calibration definitions in the technical specifications (TS) for each facility to incorporate the surveillance frequency control program. The proposed amendments are based on Technical Specification Task Force (TSTF) traveler TSTF- 563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program (ADAMS Accession No. ML17130A819).

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 revises the TS definitions of Channel Calibration, Channel Functional Test, Channel Operational Test, and Trip Actuating Device Operational Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program, as applicable. All components in the channel continue to be calibrated. The frequency at which a channel calibration is performed is not an initiator of any accident previously evaluated, so the probability of an accident is not affected by the proposed change. The channels surveilled in accordance with the affected definitions continue to be required to be operable and the acceptance criteria of the surveillances are unchanged. As a result, any mitigating functions assumed in the accident analysis will continue to be performed.

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

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

Response: No.

The proposed change revises the TS definitions of Channel Calibration, Channel Functional Test, Channel Operational Test, 13

and Trip Actuating Device Operational Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program, as applicable. The design function or operation of the components involved are not affected and there is no physical alteration of the plant (i.e., no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.

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

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

Response: No.

The proposed change revises the TS definitions of Channel Calibration, Channel Functional Test, Channel Operational Test, and Trip Actuating Device Operational Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program, as applicable. The Surveillance Frequency Control Program assures sufficient safety margins are maintained, and that design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants licensing basis. The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by method of determining surveillance test intervals under an NRC-approved licensee-controlled program.

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 requested amendments involve no significant hazards consideration.

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

NRC Acting Branch Chief: Lisa M. Regner Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No.

1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station Unit No. 1, Oswego County, New York Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania 15

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York Date of amendment request: June 27, 2019. A publicly-available version is in ADAMS under Accession No. ML19178A291.

Description of amendment request: The amendments would revise the requirements in the technical specifications for each facility related to the unavailability of barriers. The proposed amendments are based on Technical Specification Task Force (TSTF) traveler TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System OPERABILITY (ADAMS Accession No. ML061240055).

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

Criterion 1-The Proposed Change Does Not Involve a Significant Increase in the Probability of Consequences of an Accident Previously Evaluated The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable hazard barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed [Limiting Condition for Operation] LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9.

Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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Criterion 2-The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable hazard barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3-The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable hazard barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in [Regulatory Guide] RG 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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

NRC Acting Branch Chief: Lisa M. Regner 17

Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Unit Nos. 1 and 2, Calvert County, Maryland Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York 18

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania Date of amendment request: June 26, 2019. A publicly-available version is in ADAMS under Accession No. ML19178A304.

Description of amendment request: Except for Calvert Cliffs, the proposed amendments would revise the technical specifications (TS) for high radiation area administrative controls. The proposed amendments for Calvert Cliffs would add TS requirements for high radiation area administrative controls.

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. Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes are administrative in nature and only related to the control of access to high radiation areas for controlling dose to plant personnel. The proposed changes do not impact any accident initiators and do not require any plant modifications which affect the performance capability of the structures, systems and components relied upon to mitigate the consequences of postulated accidents; therefore, there is no impact to the probability or consequences of an accident previously evaluated.

Based on the above, EGC concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendments involve changes to radiological program controls for access to high radiation areas, which are 19

administrative in nature and do not impact physical plant systems.

These proposed changes do not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed changes do not require any plant modifications which affect the performance capability of the structures, systems and components relied upon to mitigate the consequences of postulated accidents.

Based on the above discussion, EGC concludes that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes are administrative in nature and only related to the control of access to high radiation areas to minimize dose to plant personnel. The proposed changes are intended to provide clarity and/or flexibility with respect to the administration and programmatic controls while retaining adequate margin of safety for minimizing dose to site personnel consistent with the requirements of 10 CFR 20, Standards for Protection Against Radiation, and the guidance of [Regulatory Guide] RG 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, published in May 2006. Since there are no associated physical plant changes, the ability of the plant to respond to and mitigate accidents is unchanged by the proposed changes.

Based on the above, EGC concludes that the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Acting Branch Chief: Lisa M. Regner 20

Exelon Generation Company (EGC), LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No.1, DeWitt County, Illinois and Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois Date of amendment request: June 18, 2019. A publicly-available version is in ADAMS under Accession No. ML19169A146.

Description of amendment request: The proposed amendments would revise the CPS, Unit No. 1, and DNPS, Units 2 and 3, technical specifications (TSs) associated with TS 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control (WIC)," and TS 3.8.2, "AC Sources - Shutdown," surveillance requirements considered no longer necessary following NRC-approved licensing activity at these sites. For each site, a change to TS 3.3.5.2, Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation, is proposed to support instrumentation functions. Additionally, edits are proposed to RPVWIC-related TSs to add consistency and clarity. For DNPS, Units 2 and 3 only, a change to TS 3.6.1.3, Primary Containment Isolation Valves, is proposed to support Mode 4 and 5 operations.

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 modifies existing TS requirements related to the maintenance of RPV inventory in Modes 4 and 5. Draining of RPV water inventory in Modes 4 and 5 is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a modified set of controls has no effect on any accident previously evaluated. RPV 21

water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing and the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes do not affect the probability of an unexpected draining event (which is not a previously evaluated accident) or the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the TAF [top of active fuel]. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes do not affect the consequences of an unexpected draining event (which is not a previously evaluated accident) or the current requirement to maintain an operable ECCS [emergency core cooling system] subsystem at all times in Modes 4 and 5. The proposed changes do not significantly affect the consequences of an unexpected draining event because the proposed Actions continue to ensure equipment is available within the limiting DRAIN TIME, and are equivalent to the current requirements.

The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as the automatic starting of EDGs [emergency diesel generators] on ECCS initiation signals. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements proposed for elimination are not needed to adequately respond to a draining event.

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 previously evaluated?

Response: No.

The proposed changes replace existing TS requirements related to RPV WIC with modified requirements that will continue to protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved.

The event of concern under the current requirements and the proposed changes is an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause an RPV or 22

refueling cavity draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

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

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

Response: No.

The proposed changes modify certain existing TS requirements related to RPV WIC. The safety basis for the current RPV WIC requirements is to protect Safety Limit 2.1.1.3. The new TS requirements continue to meet this safety basis in all respects.

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

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

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

NRC Acting Branch Chief: Lisa M. Regner.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: March 5, 2019, as supplemented by letters dated May 23 and July 22, 2019. Publicly-available versions are in ADAMS under Accession Nos.

ML19064B368, ML19143A347, and ML19203A176, respectively.

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Description of amendment request: The proposed amendment would: revise the combined main steam isolation valve (MSIV) leakage rate limit for all four steam lines in Technical Specification (TS) TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement (SR) 3.6.1.3; revise the leakage rate through each MSIV leakage path; add a new TS 3.6.2.6, Residual Heat Removal (RHR) Drywell Spray; and revise TS 3.6.4.1, Secondary Containment, to address short-duration conditions during which the secondary containment pressure may not meet the SR pressure requirement, in accordance with Technical Specifications Task Force Traveler (TSTF) 551, Revise Secondary Containment Surveillance Requirements, Revision 3.

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

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

Response: No.

The increase in the total MSIV leakage rate limit has been evaluated in a revision to the radiological consequence analysis of the Loss of Coolant Accident (LOCA). Based on the results of the analysis, it has been demonstrated that, with the requested change, the dose consequences of this limiting Design Basis Accident (DBA) are within the acceptance criteria provided by the NRC for use with the Alternative Source Term (AST) methodology in 10 CFR 50.67 and 10 CFR 50, appendix A, GDC [General Design Criteria] 19. Additional guidance is provided in Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" and Standard Review Plan (SRP) Section 15.0.1.

The proposed change to the MSIV leakage limit does not involve physical change to any plant structure, system, or component. As a result, no new failure modes of the MSIVs have been introduced.

24

The proposed change does not affect the normal design or operation of the facility before the accident; rather, it affects leakage limit assumptions that constitute inputs to the evaluation of the consequences. The radiological consequences of the analyzed LOCA have been evaluated using the plant licensing basis for this accident. The resulting doses are slightly higher than the previously approved AST doses; with exception of the Control Room dose that is slightly lower. However, adequate margin to the regulatory limits specified in 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room operator doses is still available. Thus, the results conclude that the control room and offsite doses remain within applicable regulatory limits. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

In addition, the proposed change to SR 3.6.4.1.1 addresses short-duration conditions during which the secondary containment vacuum requirement is not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four-hour Completion Time (i.e., allowed outage time) for an inoperable secondary containment. In addition, the proposed change provides an alternative means to ensure the secondary containment safety function is met. As a result, the consequences of an accident previously evaluated are not significantly increased.

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 previously evaluated?

Response: No.

The change in the MSIV leakage rate limits does not affect the design, functional performance, or normal operation of the facility.

Similarly, it does not affect the design or operation of any component in the facility such that new equipment failure modes are created. This is supported by operating experience at other EGC sites that have increased their MSIV leakage limits. As such the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

In addition, the proposed change to SR 3.6.4.1.1 does not alter the protection system design, create new failure modes, or 25

change any modes of operation. The proposed change does 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 previously evaluated.

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

Response: No.

This proposed license amendment involves changes in the MSIV leakage rate limits. The revised leakage rate limits are used in the reanalysis of the LOCA radiological consequences.

The analysis has been performed using conservative methodologies. Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed LOCA event has been carefully selected and margin has been retained to ensure that the analysis adequately bounds postulated event scenario. The dose consequences of this limiting event are within the acceptance criteria presented in 10 CFR 50.67 for offsite doses and 10 CFR 50, appendix A, GDC 19 for control room operator doses. The margin of safety is that provided by meeting the applicable regulatory limits.

In addition, the proposed change to SR 3.6.4.1.1 addresses short-duration conditions during which the secondary containment vacuum requirement is not met. Conditions in which the secondary containment vacuum is less than the required vacuum are acceptable provided the conditions do not affect the ability of the SGT [standby gas treatment] System to establish the required secondary containment vacuum under post-accident conditions within the time assumed in the accident analysis. This condition is incorporated in the proposed change by requiring an analysis of actual environmental and secondary containment pressure conditions to confirm the capability of the SGT System is maintained within the assumptions of the accident analysis.

Therefore, the safety function of the secondary containment is not affected.

Therefore, the proposed changes do 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 26

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

Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Acting Branch Chief: Lisa M. Regner.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa Date of amendment request: April 9, 2019. A publicly-available version is in ADAMS under Accession No. ML19101A280.

Description of amendment request: The amendment would revise the DAEC Emergency Plan on-shift and augmented Emergency Response Organization (ERO) staffing to support the planned permanent cessation of operations and permanent defueling of the DAEC reactor. Specifically, the proposed changes would eliminate the on-shift positions not needed for the safe storage of spent fuel in the spent fuel pool during the initial decommissioning period and eliminate the ERO positions not necessary to effectively respond to credible accidents. The proposed changes in staffing are commensurate with the reduced spectrum of credible accidents for a permanently shut down and defueled power reactor facility.

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.

27

The proposed changes to the DAEC Emergency Plan do not impact the function of plant Structures, Systems, or Components (SSCs). The proposed changes do not involve the modification of any plant equipment or affect plant operation. The proposed changes do not affect accident initiators or precursors, nor do the proposed changes alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and ERO to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently defueled condition. The proposed changes only remove positions that will no longer be needed or credited in the Emergency Plan in the permanently defueled condition.

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

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

Response: No.

The proposed changes reduce the number of on-shift and ERO positions commensurate with the hazards associated with a permanently shut down and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Additionally, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new accident initiators are created.

Therefore, the proposed changes do 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.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes.

The proposed changes are associated with the Emergency Plan 28

and staffing and do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes.

Safety analysis acceptance criteria are not affected by the proposed changes and margins of safety are maintained. The revised Emergency Plan will continue to provide the necessary response staff with the proposed changes.

Therefore, the proposed changes have no impact to 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: Steven Hamrick, Managing Attorney - Nuclear, Florida Power Light Company, P. O. Box 14000, Juno Beach, FL 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

NextEra Energy Duane Arnold (NEDA), LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa Date of amendment request: June 20, 2019. A publicly-available version is in ADAMS under Accession No. ML19176A356.

Description of amendment request: NEDA requests an amendment to the DAEC operating license (OL) and technical specifications (TSs). The proposed changes will revise the OL and TSs consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. The revised OL and TSs will be identified as the DAEC post defueled technical specifications (PDTSs). By letter dated January 18, 2019 (ADAMS Accession No. ML19023A196), NEDA provided formal notification to the NRC 29

pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR 50.4(b)(8) of the intention to permanently cease power operations at the DAEC in the fourth quarter of 2020. After the certifications of permanent cessation of power operation and of permanent removal of fuel from the DAEC reactor vessel are docketed, in accordance with 10 CFR 50.82(a)(1)(i) and (ii) respectively, and pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 license will no longer authorize reactor operation or emplacement or retention of fuel in the reactor vessel. As a result, certain license conditions and TSs may be revised or removed to reflect the permanently defueled condition. In general, the changes propose the elimination of items applicable in operating conditions where fuel is placed in the reactor 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes would not take effect until DAEC has certified to the NRC that it has permanently ceased operation and entered a permanently defueled condition. Because the 10 CFR part 50 license for DAEC will no longer authorize operation of the reactor, or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. DAECs accident analyses are contained in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR).

In a permanently defueled condition, the only credible UFSAR described accident that remains is the Fuel Handling Accident (FHA). Other Chapter 15 accidents will no longer be applicable to a permanently defueled reactor.

The UFSAR-described FHA analyses for DAEC shows that, following the required decay time after reactor shutdown and provided the SFP [spent fuel pool] water level requirement of TS LCO [limiting condition for operation] 3.7.8 is met, the dose 30

consequences are acceptable without relying on secondary containment or the Standby Gas Treatment System. The control building envelop is credited for reduction of operator dose.

Consequently, the TS requirements for the Standby Filter Unit and Control Building Chillers are retained.

The probability of occurrence of previously evaluated accidents is not increased, since safe storage and handling of fuel will be the only operations performed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible in the permanently defueled condition. This significantly reduces the scope of applicable accidents. The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility SSCs

[structures, system, and components] or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled DAEC has no impact on the remaining applicable DBA [design-basis accident].

The removal of LCOs or SRs [surveillance requirements] that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the permanently defueled condition.

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 changes to delete or modify certain DAEC Operating License, TS, and current licensing bases (CLB) 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 the spent irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor, or only to the prevention, diagnosis, or mitigation of reactor related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled.

31

The proposed modification or deletion of requirements of the DAEC Operating License, TS, and CLB do not affect systems credited in the accident analysis for the remaining credible DBA at DAEC. The proposed Operating License and PDTS will continue to require proper control and monitoring of safety significant parameters and activities. The TS regarding SFP water level and spent fuel storage is retained to preserve the current requirements for safe storage of irradiated fuel. The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding, spent fuel racks, SFP integrity, and SFP water level).

Since extended operation in a defueled condition and safe fuel handling 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.

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 changes are to delete or modify certain Operating License, TS and CLB once the DAEC facility has been permanently shut down and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for DAEC 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 UFSAR described accident is a FHA.

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 Operating License, TS, and CLB that are not related to the safe storage of irradiated fuel. The requirements proposed to be revised or deleted from the Operating License, TS, and CLB are not credited in the existing accident analysis for the remaining postulated accident (i.e., FHA); and, as such, do not contribute to the margin of safety associated with the accident analysis.

Certain postulated DBAs involving the reactor are no longer possible because the reactor will be permanently shut down and defueled and DAEC will no longer be authorized to operate the reactor.

32

Therefore, the proposed changes have no impact to the margin of safety.

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

Attorney for licensee: Steven Hamrick, Managing Attorney - Nuclear, Florida Power Light Company, P. O. Box 14000, Juno Beach, FL 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1 (Seabrook), Rockingham County, New Hampshire Date of amendment request: June 4, 2019. A publicly-available version is in ADAMS under Accession No. ML19157A057.

Description of amendment request: The amendment would revise the Seabrook Technical Specifications (TSs) associated with the emergency core cooling system (ECCS) accumulators. Specifically, the proposed amendment would modify the TS actions for an inoperable accumulator, relocate the actions for inoperable accumulator instrumentation, and delete an unnecessary surveillance requirement. The proposed change would also delete a duplicate surveillance requirement associated with the accumulator isolation valves.

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:

33

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

Response: No.

Operability of the ECCS accumulators ensure that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the reactor coolant system (RCS) pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The proposed change does not change the limiting condition for operation (LCO) for the accumulators.

The proposed change deletes a surveillance requirement that verifies the accumulator isolation valves automatically open on an actuation signal because the technical specifications require maintaining the motor-operated valves open and de-energized. In addition, the completion times for an inoperable accumulator are revised to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for inoperability due to reasons other than boron concentration outside limits and to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for boron not within limits. The consequences of an accident that might occur during the revised completion times are no different from those that might occur during the current completion times. The change to eliminate a duplicate surveillance requirement makes no technical changes and is administrative in nature.

The proposed change does not alter the design, function, or operation of any plant structure, system, or component (SSC).

The capability of any operable TS-required SSC to perform its specified safety function is not impacted by the proposed change.

As a result, the outcomes of accidents previously evaluated are unaffected. Therefore, the proposed changes do 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 change does not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant system or component. No physical changes are made to the plant, so no new causal mechanisms are introduced. Therefore, the proposed changes to the TS do not create the possibility of a new or different kind of accident from any accident previously evaluated.

34

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

Response: No.

The ability of any operable ECCS equipment to perform its designated safety function is unaffected by the proposed changes.

The proposed changes do not alter any safety analyses assumptions, safety limits, limiting safety system settings, or method of operating the plant. The changes do not adversely affect plant operating margins or the reliability of equipment credited in the safety analyses. With the proposed change, the ECCS remains capable of performing its safety function.

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: Debbie Hendell, Managing Attorney - Nuclear, Florida Power &

Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: James G. Danna.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), Unit No. 1, Fairfield County, South Carolina Date of amendment request: July 30, 2019. A publicly-available version is in ADAMS under Accession No. ML19214A046.

35

Description of amendment request: The proposed amendment would replace South Carolina Electric & Gas Company with Dominion Energy South Carolina, Inc. or DESC where appropriate in the Renewed Facility Operating License NPF-12.

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 amendment is administrative in nature. SCE&G, which has been renamed Dominion Energy South Carolina, Inc.,

will remain the licensee authorized to operate and possess VCSNS Unit 1, and its functions, powers, resources and management as described in the license will not change. The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

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

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

Response: No.

The proposed amendment is purely administrative in nature. The functions of the licensee will not change. These changes do not involve any physical alteration of the plant (i.e., no new or different type of equipment will be installed), and installed equipment is not being operated in a new or different manner. Thus, no new failure modes are introduced. 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 36

margin of safety?

Response: No.

The proposed amendment is administrative in nature. SCE&G, which has been renamed Dominion Energy South Carolina, Inc.,

will remain the licensee authorized to operate and possess the units, and its functions as described in the license will not change.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There are no changes to setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected. Therefore, the proposed changes do 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., Docket Nos. 50-348 and 50-364, Joseph M.

Farley Nuclear Plant (FNP), Units 1 and 2, Houston County, Alabama; Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant (HNP), Unit Nos. 1 and 2, Appling County, Georgia; and Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Burke County, Georgia 37

Date of amendment request: July 15, 2019. A publicly-available version is in ADAMS under Accession No. ML19196A222.

Description of amendment request: The amendments would adopt Technical Specification Task Force (TSTF)-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program. TSTF-563 revises the Technical Specification (TS) definitions of Channel Calibration and Channel Functional Test in the HNP TS, and the definitions of Channel Calibration, Channel Operational Test (COT), and Trip Actuating Device Operational Test (TADOT) in the FNP and VEGP TS. The HNP, FNP, and VEGP Channel Calibration definition and the HNP Channel Functional Test definition currently permit performance by means of any series of sequential, overlapping, or total channel steps. The FNP and VEGP definitions of COT and TADOT are revised to explicitly permit performance by means of any series of sequential, overlapping, or total channel steps. The Channel Calibration, Channel Functional Test, COT, and TADOT definitions are revised to allow the required frequency for testing the components or devices in each step to be determined in accordance with the Surveillance Frequency Control 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test in the HNP TS, and the definitions of Channel Calibration, COT, and TADOT in the FNP and VEGP TS to allow the frequency for testing the components or devices in each step to be determined in accordance with the 38

Surveillance Frequency Control Program. The proposed change also explicitly permits the FNP and VEGP COT and TADOT to be performed by any series of sequential, overlapping, or total channel steps. All components in the channel continue to be tested. The frequency at which a channel test is performed is not an initiator of any accident previously evaluated, so the probability of an accident is not affected by the proposed change. The channels surveilled in accordance with the affected definitions continue to be required to be operable and the acceptance criteria of the surveillances are unchanged. As a result, any mitigating functions assumed in the accident analysis will continue to be performed.

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

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

Response: No.

The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test in the HNP TS, and the definitions of Channel Calibration, COT, and TADOT in the FNP and VEGP TS to allow the frequency for testing the components or devices in each step to be determined in accordance with the Surveillance Frequency Control Program. The proposed change also explicitly permits the FNP and VEGP COT and TADOT to be performed by any series of sequential, overlapping, or total channel steps. The design function or operation of the components involved are not affected and there is no physical alteration of the plant (i.e., no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test in the HNP TS, and the 39

definitions of Channel Calibration, COT, and TADOT in the FNP and VEGP TS to allow the frequency for testing the components or devices in each step to be determined in accordance with the Surveillance Frequency Control Program. The proposed change also explicitly permits the FNP and VEGP COT and TADOT to be performed by any series of sequential, overlapping, or total channel steps. The Surveillance Frequency Control Program assures sufficient safety margins are maintained, and that design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants' licensing basis. The proposed change does not adversely affect existing plant safety margins, or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by method of determining surveillance test intervals under an NRC-approved licensee-controlled program.

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

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

Attorney for licensee: Millicent Ronnlund, Vice President and General Counsel, Southern Nuclear Operating Co., Inc., P. O. Box 1295, Birmingham, AL 35201-1295.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: July 9, 2019. A publicly-available version is in ADAMS under Accession No. ML19190A309.

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Description of amendment request: The amendments would revise the actions of Technical Specification (TS) 3.7.7, Component Cooling Water (CCW) System, TS 3.7.8, Nuclear Service Cooling Water (NSCW) System, TS 3.8.1, AC Sources -

Operating, TS 3.8.4, DC Sources - Operating, TS 3.8.7, Inverters - Operating, and TS 3.8.9, Distribution Systems - Operating. The proposed license amendments modify action end states for the subject TS in conditions where more than one safety-related train is inoperable or the electrical power system is significantly degraded. Specifically, if the related required action statements are not met, instead of requiring the plant to achieve hot shutdown (i.e., Mode 4), the end state of cold shutdown (i.e., Mode 5) is required.

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 requires the plant to be placed in cold shutdown instead of hot shutdown when more than one safety-related train of the cooling water or electrical distribution systems are inoperable or when the electrical power system is significantly degraded (e.g., three or more required AC [alternating current]

sources inoperable). Transitioning the plant from hot shutdown to cold shutdown is not an initiator of any accident previously evaluated but is assumed in the mitigation of accidents previously evaluated. Therefore, the probability of an accident previously evaluated is not adversely impacted by the proposed change.

Component cooling water (CCW) and nuclear service cooling water (NSCW) systems and the safety-related electrical power and distribution systems are assumed in accident mitigation. SNC concludes the proposed change to require the plant be placed in cold shutdown instead of hot shutdown is acceptable because placing the unit in cold shutdown is considered a safe condition, since most design basis accidents and transients either cannot 41

physically occur during cold shutdown, or would have significantly reduced plant impact and occur much less frequently due to the reduced temperatures and pressures in the plant. Therefore, the consequences of any accident that assumes the cooling water systems or electrical power and distribution systems are not significantly affected by this change.

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

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

Response: No.

The proposed change does not change the design function or operation of the cooling water systems or the electrical power and distribution systems. No plant modifications or changes to the plant configuration or method of operation are involved. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not affect any of the controlling values of parameters used to avoid exceeding regulatory or licensing limits. The proposed change does not exceed or alter the design basis or safety limits, or any limiting safety system settings. The requirement for the CCW and NSCW systems to perform their designated support functions is unaffected. The requirement for the safety-related electrical power and distribution systems to perform their designated support functions is unaffected. The proposed change to require the plant be placed in cold shutdown instead of hot shutdown is acceptable because placing the unit in cold shutdown is considered a safe condition, since most design basis accidents and transients either cannot physically occur during cold shutdown, or would have significantly reduced plant impact and occur much less frequently due to the reduced temperatures and pressures in the plant.

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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: Millicent Ronnlund, Vice President and General Counsel, Southern Nuclear Operating Co., Inc., P. O. Box 1295, Birmingham, AL 35201-1295.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: June 28, 2019. A publicly-available version is in ADAMS under Accession No. ML19179A209.

Description of amendment request: The amendment proposes changes to credit previously completed first plant only startup testing described in the Updated Final Safety Analysis Report (UFSAR), and related changes to the Combined License (COL)

Nos. NPF-91 and NPF-92 for VEGP Units 3 and 4. Specifically, the proposed changes would revise the COL to delete conditions requiring the following tests: Natural Circulation (Steam Generator) Test, Rod Cluster Control Assembly (RCCA) Out of Bank Measurements, Load follow Demonstration, and the Passive Residual Heat Exchanger Test.

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

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

Response: No.

The proposed change does not affect the operation of any systems or equipment that initiates an analyzed accident or alter any structures, systems, or components (SSC) accident initiator or initiating sequence of events. The proposed change involves removing the requirement to perform first plant only startup tests including the Natural Circulation (Steam Generator) Test, the RCCA Out of Bank Measurements, the Load Follow Demonstration, and the Passive Residual Heat Exchanger Test.

The request is based on the successful completion of these tests at the lead AP1000 unit. The change does not adversely affect any methodology which would increase the probability or consequences of a previously evaluated accident.

The change does not impact the support, design, or operation of mechanical or fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions.

There is no change to predicted radioactive releases due to normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, 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 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 change does 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 change credits previously completed first plant only startup tests including the Natural Circulation (Steam Generator)

Test, the RCCA Out of Bank Measurements, the Load Follow Demonstration, and the Passive Residual Heat Exchanger Test.

The request is based on the successful completion of the tests at the lead AP1000 unit. The proposed changes do not adversely affect any design function of any SSC design functions or methods of operation in a manner that results in a new failure 44

mode, malfunction, or sequence of events that affect safety-related or non-safety-related equipment. This activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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 maintains existing safety margin and provides adequate protection through continued application of the existing requirements in the UFSAR. The proposed change satisfies the same design functions in accordance with the same codes and standards as stated in the UFSAR. This change does not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change. Since no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by this change, no significant 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 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Jennifer L. Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia 45

Date of amendment request: July 8, 2019. A publicly-available version is in ADAMS under Accession No. ML19189A180.

Description of amendment request: The amendment request proposes changes to the Combined License (COL) Numbers NPF-91 and NPF-92 for VEGP Units 3 and 4. The requested amendment proposes changes to Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in COL Appendix C, with corresponding changes to the associated plant-specific Tier 1 information. 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 (DCD) Tier 1 material departures. Specifically, the requested amendment proposes changes to COL Appendix C (and plant-specific Tier 1) to remove a number of functional arrangement ITAAC, whose design commitments may be completed via other ITAAC or otherwise verified by other means.

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 non-technical change to COL Appendix C will remove a number of functional arrangement ITAAC to improve efficiency of the ITAAC completion and closure process. No structure, system, or component (SSC) design or function is affected. No design or safety analysis is affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not affected.

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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 COL Appendix C do not affect the design or function of any SSC but will remove a number of functional arrangement ITAAC to improve efficiency of the ITAAC completion and closure process. The proposed changes would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release.

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 to COL Appendix C will remove a number of functional arrangement ITAAC to improve efficiency of the ITAAC completion and closure process, and would not affect any design parameter, function or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit or criterion is involved.

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

NRC Branch Chief: Jennifer L. Dixon-Herrity.

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4. 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 48

accessed as described in the Obtaining Information and Submitting Comments section of this document.

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 5, 2019.

Brief description of amendments: The amendments correct an editorial error in Section 3.0, SR [Surveillance Requirement] APPLICABILITY, specifically, SR 3.0.5. The amendments also modified Technical Specifications (TS) 3.5.2, ECCS [Emergency Core Cooling System] - Operating, TS 3.6.6, Containment Spray System, TS 3.7.5, Auxiliary Feedwater (AFW) System, TS 3.7.6, Component Cooling Water (CCW)

System, TS 3.7.7, Nuclear Service Water System (NSWS), TS 3.7.9, Control Room Area Ventilation System (CRAVS), TS 3.7.11, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), TS 3.8.1, AC [Alternating Current] Sources - Operating, and TS 3.8.4, DC [Direct Current] Sources - Operating to remove expired TS footnotes.

Date of issuance: August 8, 2019.

Effective date: These amendments are effective as of the date of issuance and shall be implemented within 120 days of issuance.

Amendment Nos.: 316 (Unit 1) and 295 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML19184A585; 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: The amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: April 23, 2019 (84 FR 16893).

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The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated August 8, 2019.

No significant hazards consideration comments received: Yes. One comment from a member of the public was received, however it was not related to the no significant hazards consideration determination or the license amendment request.

Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1, Oswego County, New York Date of amendment request: June 26, 2018, as supplemented by letters dated February 25, 2019, May 17, 2019, and July 30, 2019. Publicly-available versions are in ADAMS under Accession Nos. ML18177A044, ML19056A387, ML19137A070, and ML19211C702, respectively.

Brief description of amendment: The amendment revised Technical Specification 3.3.1, Oxygen Concentration, to require inerting the primary containment to less than four percent by volume oxygen concentration within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of entering power operating condition. Also, the amendment added a new requirement to identify required actions, if the primary containment oxygen concentration increases to greater than or equal to four volume percent while in the power operating condition.

Date of issuance: July 30, 2019.

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

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Amendment No.: 237. A publicly-available version is in ADAMS under Accession No. ML19176A086; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

On December 18, 2018, the Nuclear Regulatory Commission (NRC or the Commission) staff published a proposed no significant hazards consideration (NSHC) determination in the Federal Register (83 FR 64894) for the proposed amendment. Subsequently, by letters dated February 28, 2019, and May 17, 2019, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC published a second proposed NSHC determination in the Federal Register on June 18, 2019 (84 FR 28346), which superseded the original notice in its entirety. The supplemental letter dated July 30, 2019, provided additional information that clarified the application, did not expand the scope of the application as noticed, and did not change the staff's second proposed no significant hazards consideration determination as published in the Federal Register.

Date of initial notice in Federal Register: December 18, 2018 (83 FR 64894).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated July 30, 2019.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant (CNP), Unit 1, Berrien County, Michigan 51

Date of amendment request: March 7, 2018.

Brief description of amendment: The amendment approves the use of a leak-before-break methodology on designated reactor coolant system (RCS) piping segments associated with the CNP, Unit 1, accumulator, residual heat removal (RHR), and safety injection (SI) systems. The approved methodology provides the CNP, Unit 1, with additional design margin for future RCS piping analysis on these systems. The amendment also modifies technical specification 3.4.13, RCS Operational LEAKAGE, including adding requirements to meet the RCS operational leakage limits as specified in the technical specifications limiting conditions for operations 3.4.13.

Date of issuance: August 1, 2019.

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

Amendment No.: 346. A publicly-available version is in ADAMS under Accession No. ML19170A362; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-58: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: May 18, 2018 (83 FR 20862).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 1, 2019.

No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1 (Seabrook), Rockingham County, New Hampshire 52

Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2 (St. Lucie), 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 (Turkey Point), Miami-Dade County, Florida Date of amendment request: May 29, 2018, as supplemented by letter dated March 26, 2019.

Brief description of amendments: The amendments revised the Technical Specifications to include the provisions of Limited Condition for Operation (LCO) 3.0.6 in the Standard Technical Specifications. In support of this change, the licensee also added a new Safety Function Determination Program to the administrative section of the Technical Specification; added new notes and actions that direct entering the actions for the appropriate supported systems; made changes to LCO 3.0.2 for Seabrook, St. Lucie, and Turkey Point; and made changes to LCO 3.0.1 for Seabrook and Turkey Point.

Date of issuance: July 31, 2019.

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

Amendment Nos: 161 (Seabrook, Unit No. 1); 249 and 200 (St. Lucie, Unit Nos. 1 and 2); and 287 and 281 (Turkey Point, Unit Nos. 3 and 4). A publicly-available version is in ADAMS under Accession No. ML19148A744; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-86, DPR-67, NPF-16, DPR-31, and DPR-41: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: September 11, 2018 (83 FR 45985). The supplement dated March 26, 2019, provided additional information that clarified the 53

application, did not expand the scope of the application as originally noticed, and did not change the NRC 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 July 31, 2019.

No significant hazards consideration comments received: No.

Northern States Power Company - Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: November 12, 2018, as supplemented by letter dated April 18, 2019.

Brief description of amendment: The amendment revised the technical specifications to delete the note associated with limiting condition for operation 3.5.1. The deleted note permitted low pressure coolant injection subsystems to be consider operable in certain plant conditions.

Date of issuance: July 30, 2019.

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

Amendment No.: 202. A publicly-available version is in ADAMS under Accession No. ML19162A093; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

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Date of initial notice in Federal Register: January 2, 2019 (84 FR 24). The supplemental letter dated April 18, 2019 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 July 30, 2019.

No significant hazards consideration comments received: No.

Northern States Power Company - Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant PINGP), Units 1 and 2, Goodhue County, Minnesota Date of amendment request: May 18, 2018, as supplemented by letters dated July 10, 2018, December 8, 2018, and April 8, 2019.

Brief description of amendment: The amendments revised the approved fire protection program (FPP). Specifically, the amendments deleted several modifications which are required as part of PINGPs implementation of its risk-informed, performance-based FPP in accordance with 10 CFR paragraph 50.48(c), National Fire Protection Association Standard 805.

Date of issuance: July 30, 2019.

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

Amendment Nos.: 228-Unit 1; 216-Unit 2. A publicly-available version is in ADAMS under Accession No. ML19140A447; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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Renewed Facility Operating License Nos. DPR-42 and DPR-60: The amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: August 14, 2018 (83 FR 40350). The supplemental letters dated July 10, 2018, December 8, 2018, and April 8, 2019, 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 July 30, 2019.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station (Hope Creek),

Salem County, New Jersey Date of amendment request: October 30, 2018.

Brief description of amendment: The amendment revised Hope Creek Technical Specification 3.3.7.4, Remote Shutdown System Instrumentation and Controls, to make the requirements consistent with Standard Technical Specification 3.3.3.2, Remote Shutdown System, in NUREG-1433, Volume 1, Revision 4. The amendment increases the allowed outage time for inoperable remote shutdown system components from 7 days to 30 days. The amendment also deletes Tables 3.3.7.4-1, 3.3.7.4-2, and 4.3.7.4-1, and relocates these tables to the Technical Requirements Manual, where they will be directly controlled by the licensee.

Date of issuance: August 6, 2019.

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

Amendment No.: 217. A publicly-available version is in ADAMS under Accession No. ML19186A205; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-57: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: December 18, 2018 (83 FR 64897).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated August 6, 2019.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, Joseph M.

Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request: December 14, 2018.

Brief description of amendments: The amendments revise a license condition associated with its approved fire protection program under 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805. Specifically, the plant operating licenses have been revised to allow, as a performance-based method, use of thermal insulation materials in limited applications subject to appropriate engineering reviews and controls, as a deviation from NFPA 805 Chapter 3, Section 3.3, Prevention.

Date of issuance: July 30, 2019.

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

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Amendment Nos.: 224 (Unit 1) and 221 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML19156A262; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: February 12, 2019 (84 FR 3510).

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated July 30, 2019.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit No. 1, Callaway County, Missouri Date of amendment request: September 4, 2018, as supplemented by letter dated February 20, 2019.

Brief description of amendment: The amendment revised Emergency Action Levels CA6.1, Cold Shutdown/Refueling System Malfunction - Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert, and SA9.1, System Malfunction - Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert. In addition, the amendment added a new definition for the term Loss of Safety Function (LOSF) and re-definition of the term Visible Damage and deleted Initiating Condition HG1 and associated EAL HG1.1, Hazard - HOSTILE ACTION resulting in loss of physical control of the facility: General Emergency, within the Callaway Plant, Unit No. 1 Radiological Emergency Response Plan.

Date of issuance: July 30, 2019.

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

Amendment No.: 220. A publicly-available version is in ADAMS under Accession No. ML19158A290; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-30: The amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: December 4, 2018 (83 FR 62621). The supplement dated February 20, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated July 30, 2019.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 23rd day of August, 2019.

For the Nuclear Regulatory Commission.

/RA/

Gregory F. Suber, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

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