2017-25063. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from October 24, 2017 to November 6, 2017. The last biweekly notice was published on November 7, 2017.

DATES:

Comments must be filed by December 21, 2017. A request for a hearing must be filed by January 22, 2018.

ADDRESSES:

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

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0220. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: May Ma, Office of Administration, Mail Stop: OWFN-2-A13, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: [phone], email: [e-mail].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0220, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0220.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. 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.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2017-0220, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

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

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

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

Within 60 days after the date of publication of this notice, any 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 Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. Alternatively, a copy of the regulations is available at the NRC's 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 petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's 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 party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's 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 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 petitioner's 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. 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 NRC's 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.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [e-mail], 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 NRC's 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 NRC's 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 NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on 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 NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email to [e-mail], 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 a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at 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 NRC's 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 applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

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

Date of amendment request: August 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17248A389.

Description of amendment request: The proposed amendment would revise the PNP Site Emergency Plan (SEP) for the permanently shut down and defueled condition. The proposed PNP SEP changes would revise the shift staffing and Emergency Response Organization (ERO) staffing.

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

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

Response: No.

The proposed changes to the PNP SEP do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not affect accident initiators or precursors, nor does it alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and augmented ERO to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently shut down and defueled condition. The proposed changes only remove positions that will no longer be credited in the PNP SEP.

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 reduce the number of on-shift and augmented 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. Also, 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 amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Margin of safety is associated with 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 are associated with the PNP SEP and do not impact operation of the plant or its response to transients or accidents. The change does 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. The revised PNP SEP will continue to provide the necessary response staff with the proposed changes.

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: William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of amendment request: September 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17244A093.

Description of amendment request: The amendments would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR, Section 50.69, “Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.”

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 will permit the use of a risk-informed categorization process to modify the scope of SSCs [structures, systems, and components] subject to NRC [Nuclear Regulatory Commission] special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change.

The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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 Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: August 30, 2017, as supplemented by letter dated October 24, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17243A014 and ML17297B521, respectively.

Description of amendment request: The amendments would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR 50.69, “Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.”

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of [structures, systems, and components] SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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 Rd., Warrenville, IL 60555.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: September 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A069.

Description of amendment request: The amendments would revise Technical Specification (TS) requirements related to the direct current (DC) electrical power system. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.”

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

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

Response: No.

The proposed change restructures the TS for the direct current (DC) electrical power system. The proposed changes add actions to specifically address battery charger inoperability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Updated Final Safety Analysis Report (UFSAR). Operation in accordance with the proposed TS ensures that the DC electrical power system is capable of performing its function as described in the UFSAR. Therefore, the mitigative functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis, and the probability of previously analyzed accidents will not increase by implementing these changes.

The relocation of preventive maintenance surveillances, and certain operating limits and actions, to a newly created licensee-controlled Battery Monitoring and Maintenance Program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC electrical power system.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes.

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

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

Response: No.

The proposed change involves restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the UFSAR. Rather, the DC electrical power system is used to supply equipment used to mitigate an accident.

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

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery maintenance and monitoring program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions.

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 Rd., Warrenville, IL 60555.

NRC Branch Chief: James G. Danna.

Exelon Generation Company, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: October 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A520.

Description of amendment request: The amendment would revise the James A. FitzPatrick Nuclear Power Plant Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control” (ADAMS Accession No. ML16074A448). Specifically, the licensee proposed changes to replace TS requirements related to operations with a potential for draining the reactor vessel (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3.

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated, and therefore replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

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 automatic initiation of an ECCS subsystem and control room ventilation. 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 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 change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

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.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

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

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

Attorney for licensee: Donald P. Ferraro, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305, Kennett Square, PA 19348.

NRC Branch Chief: James G. Danna.

Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3, and 4, Miami-Dade County, Florida

Date of amendment request: August 23, 2017, as supplemented by letter dated October 19, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17235B008 and ML17292A789, respectively.

Description of amendment request: The amendments would modify the Technical Specifications (TSs) to relocate the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, and Gas Decay Tanks System requirements to licensee-controlled documents and establish a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed amendments also relocate the Standby Feedwater System requirements to licensee-controlled documents and modify related Auxiliary Feedwater (AFW) System requirements.

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

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

Response: No.

The proposed license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes are administrative in nature and do not alter any plant equipment or the manner in which plant equipment is operated and maintained. All equipment limitations, applicable methodologies and surveillances are maintained by the proposed changes. In addition, the proposed changes to the AFW System requirements enhance plant safety. As such, the proposed changes cannot affect the initiators, the likelihood or the expected outcomes of any analyzed accidents.

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

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

Response: No.

The proposed license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes neither install or remove plant equipment nor alter any plant equipment design, configuration, or method of operation. Hence, no new failure mechanisms are introduced as a result of the proposed changes.

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.

The proposed license amendments modify the Turkey Point TS by relocating the Explosive Gas Monitoring Instrumentation, Explosive Gas Mixture, Gas Decay Tanks and Standby Feedwater System requirements to licensee controlled documents, by relatedly modifying the AFW System requirements and by establishing a Gas Decay Tank Explosive Gas and Radioactivity Monitoring Program. The proposed changes neither involve changes to safety analyses assumptions, safety limits, or limiting safety system settings nor adversely impact plant operating margins or the reliability of equipment credited in safety analyses.

Therefore, operation of the facility in accordance with the proposed changes will not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Undine Shoop.

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

Date of amendment request: September 5, 2017. A publicly-available version is in ADAMS under Accession No. ML17248A284.

Description of amendment request: The proposed amendment would revise DAEC Technical Specifications 3.5.1, “ECCS [emergency core cooling system]-Operating.” The proposed change would decrease the nitrogen supply requirement for the Automatic Depressurization System (ADS) in Surveillance Requirement (SR) 3.5.1.3 from 100 days to 30 days.

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

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

Response: No.

The proposed change modifies a SR for verification of the nitrogen supply for the ADS accumulators. Accidents are initiated by the malfunction of plant equipment, or the catastrophic failure of plant structures, systems or components. The performance of this surveillance is not a precursor to any accident previously evaluated and does not change the manner in which the ADS operates. Technical evaluation of the change concluded that a 30-day nitrogen supply is more than adequate to ensure that the reactor is depressurized, so the consequences of an accident remain unchanged.

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

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 involve physical alterations to the plant. No new or different type of equipment will be installed, and there are no physical modifications required to existing installed equipment associated with the proposed change. The proposed change does not create any failure mechanism, malfunction or accident initiator not already considered in the design and licensing basis.

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

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

Response: No.

Although the proposed change will decrease the required supply of nitrogen for the ADS accumulators from 100 days to 30 days, the assessment above has shown that the reactor would be depressurized within 3 days following any postulated accident or event that would create a hostile environment in the drywell. Once initial depressurization is completed, long term core cooling can be assured without ADS.

Therefore, the proposed change will 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: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: August 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17243A469.

Description of amendment request: The proposed amendment would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR, part 50.69, “Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors.”

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

NextEra Energy, Point Beach Nuclear Plant (PBNP), LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: June 23, 2017, as supplemented by letter dated August 21, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17174A458, and ML17233A283, respectively.

Description of amendment request: The amendments would revise the Emergency Plan for PBNP to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” which has been endorsed by the NRC.

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

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 impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. No actual facility equipment or accident analyses are affected by the proposed changes.

The change revises the NextEra Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, “Methodology for Development of Emergency Action Levels,” but does not alter any of the requirements of the Operating License or the Technical Specifications.

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 does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed change does not create any new failure modes for existing equipment or any new limiting single failures. Additionally, the proposed change does not involve a change in the methods governing normal plant operation, and all safety functions will continue to perform as previously assumed in the accident analyses. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system.

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

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

Response: No.

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of safety-related systems and components. The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis. The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses.

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: William Blair, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: September 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17270A076.

Description of amendment request: The amendments would relocate the reactor coolant system pressure isolation valve (RCS PIV) table from the technical specifications (TSs) to the technical requirements manual (TRM). The request would also remove references to the table and move all notes and leakage acceptance criteria from the table to the TS surveillance requirements.

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

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

Response: No.

The proposed changes to the TS will not alter the way any structure, system, or component (SSC) functions, and will not alter the manner in which the plant is operated. The proposed changes do not alter the design of any SSC. The relocation of the RCS PIV valve lists from the TS to the TRM is an administrative change. Future revisions to the TRM are subject to 10 CFR 50.59. Therefore the probability of an accident previously evaluated is not significantly increased.

The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. Therefore, the consequences of an accident previously evaluated are not increased.

Therefore, these proposed changes do not represent 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 do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism. The proposed changes are administrative in nature.

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

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

Response: No.

The proposed changes to the RCS PIV TS are administrative in nature. The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. The proposed changes will not result in changes to system design or setpoints that are intended to ensure timely identification of plant conditions that could be precursors to accidents or potential degradation of accident mitigation systems.

The proposed amendment will not result in a design basis or safety limit being exceeded or altered. Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: October 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17279A715.

Description of amendment request: The proposed amendment would increase the Integrated Leak Rate Test (ILRT) Peak Calculated Containment Internal Pressure, P_a, listed in Technical Specification (TS) 6.8.4.g, “Containment Leakage Rate Testing Program,” to remove the reference to Regulatory Guide (RG) 1.163, “Performance-Based Containment Leak Test Program,” dated September 1995 and ANSI/ANS (American National Standards Institute/American Nuclear Society)—56.8-2002, “Containment System Leakage Testing Requirements,” and to replace the reference of Nuclear Energy Institute (NEI) 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based option of 10 CFR part 50, Appendix J,” with NEI 94-01, Revision 2-A.

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

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

Response: No.

The proposed changes involve removal of RG 1.163 and ANSl/ANS-56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the P_a [Peak Calculated Containment Internal Pressure] value for containment leakage testing. The activity does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the reactor containment itself and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

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

The integrity of the reactor containment is subject to two types of failure mechanisms which can be categorized as (1) activity based and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The updated P_a value reflects the updated mass and energy release and containment response calculations, ensuring a sound technical basis for the local and integrated leakage tests.

To mitigate time-based mechanisms, the design and construction requirements of the containment itself combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers], Section XI and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by a Type A test. The change to the P_a value is less than 1 psid [per square inch differential]. Radiological consequences will continue to be evaluated at the Technical Specification allowed leakage, L_a [allowed leakage] of 0.20 percent by weight of air, which will not be increased despite the increase in P_a. As described in Section 3.5, past leakage testing yielded values well under L_a. Based on the above, neither the reference changes nor the P_a change involves a significant increase in the 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 involve removal of RG 1.163 and ANSl/ANS-56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the P_a value for containment leakage testing. The reactor containment and the testing requirements invoked to periodically demonstrate the integrity of the reactor containment exist to ensure the plant's ability to mitigate the consequences of an accident. There are not any accident initiators or precursors affected by the revision. The proposed TS change does not involve a physical change to the plant or the manner in which the plant is operated or controlled.

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

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

Response: No.

The proposed changes involve removal of RG 1.163 and ANSl/ANS-56.8-2002 references, replacement of NEI 94-01, Revision 3-A with NEI 94-01, Revision 2-A, and an increase in the P_a value for containment leakage testing. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. Using the same analysis methodology as described in WCAP-10325-P-A [Westinghouse LOCA [loss-of-accident coolant] Mass and Energy Release Model for Containment Design], the updated mass and energy release and containment response analyses corrected input errors identified in the NSALs [Westinghouse Nuclear Safety Advisory Letters] described previously. As shown in Figure 1 [October 6, 2017, submittal], the correction of these errors resulted in a slightly higher predicted peak pressure than that of the current licensing basis but does not pose a significant challenge to the design limit.

The specific requirements and conditions of the Primary Containment Leak Rate Testing Program, as defined in the Technical Specifications, exist to ensure that the degree of reactor containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by the Technical Specification is maintained. The containment inspections performed in accordance with ASME, Section XI and the Maintenance Rule serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety that is in plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards will continue to be met.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: July 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17209A759.

Description of amendment request: The amendment request proposes to revise Technical Specification Section 1.1 (TS), Definition of Actuation Logic Test, by adding a new TS Section 1.1 Definition of Actuation Logic Output Test (ALOT), revising existing Surveillance Requirements 3.3.15.1 and 3.3.16.1 and adding new Surveillance Requirements 3.3.15.2 and 3.3.16.2 to implement the new ALOT. This submittal requests approval of the license amendment that is necessary to implement these changes.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(A), licensee has provided its analysis of the issue on no significant hazards consideration determination, 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.

There are no design changes associated with the proposed amendment. All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable.

The [Processor Module Self-Diagnostic (PMS)] will continue to function in a manner consistent with the plant design basis. There will be no changes to the PMS operating limits. The existing ACTUATION LOGIC TEST Surveillance Requirements are revised such that different portions of the PMS logic circuitry are tested on appropriate surveillance test frequencies.

The proposed change will not adversely affect accident initiators or precursors or adversely alter the design assumptions, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors.

The proposed changes will not alter the ability of structures, systems, and components (SSCs) to perform their specified safety functions to mitigate the consequences of an initiating event within the assumed acceptance limits.

Accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR).

The applicable radiological dose acceptance criteria will continue to be met.

The proposed change revises the frequency of testing certain portions of the PMS logic circuitry, but does not physically alter any safety-related systems.

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 different kind of accident from any accident previously evaluated?

Response: No.

With respect to any new or different kind of accident, there are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC performs its specified safety function. The proposed change will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed change will not alter any assumptions made in the safety analyses.

The proposed change revises the frequency of testing certain portions of the PMS logic circuitry. The proposed change does not involve a physical modification of the plant.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment.

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 existing ACTUATION LOGIC TEST Surveillance Requirements are revised such that different portions of the PMS logic circuitry are tested on appropriate surveillance test frequencies. The reliability of the PMS is such that not testing the Component Interface Module (CIM) logic and driver output circuits when the reactor is at power will have a net positive impact on Engineered Safety Feature Actuation System (ESFAS) availability. There will be a reduction in the potential for challenges to the safety systems, coupled with less time that the safety systems are unavailable.

There will be no effect on those plant systems necessary to effect the accomplishment of protection functions.

No instrument setpoints or system response times are affected. None of the acceptance criteria for any accident analysis will be changed.

The proposed change will have no impact on the radiological consequences of a design basis accident.

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

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 Dixon-Herrity.

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

Date of amendment request: August 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17230A365.

Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information (text) and involved Tier 2* information (as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information).

This amendment request proposes increasing the design pressure of the main steam (MS) isolation valve (MSIV) compartments from 6.0 to 6.5 psi and proposes other changes to the licensing basis regarding descriptions of the MSIV compartments.

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

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

Response: No.

The proposed changes do not adversely affect the operation of any structures, systems, and components inside or outside the auxiliary building that could initiate or mitigate abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods, tornado missiles, and turbine missiles, or their safety or design analyses, evaluated in the UFSAR. The changes do not adversely affect any design function of the auxiliary building or the structures, systems, and components contained therein. The ability of the affected auxiliary building main steam isolation valve compartments and adjacent rooms, including the main control room, to withstand the pressurization effects from the postulated pipe ruptures is not adversely affected by the increase in design pressure, since the structures, systems, and components therein remain qualified for this service.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that might initiate a new or different kind of accident, or alter any [structure, system, and component (SSC)] such that a new accident initiator or initiating sequence of events is created. The proposed changes do not adversely affect the physical design and operation of the [in-containment refueling water storage tank (IRWST)] injection, drain, containment recirculation, and fourth-stage [automatic depressurization system (ADS)] valves, including as-installed inspections, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves is not adversely affected. These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, 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 margin of safety for the design of the auxiliary building is maintained through continued use of approved codes and standards as stated in the UFSAR, and adherence to the assumptions used in the analyses of this structure and the events associated with this structure. The auxiliary building continues to be a seismic Category I building with all current structural safety margins maintained. The 3-hour fire rating requirements for the impacted auxiliary building walls are maintained. The equipment housed in the main steam isolation valve compartments continue to be environmentally qualified for their intended service in accordance with the approved codes and standards stated within the UFSAR. Thus, the requested changes will not adversely affect any safety-related equipment, 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 requested change, thus, no margin of safety is reduced. 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: October 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17279A084.

Description of amendment request: The amendment request proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document (DCD) Tier 2 information) and involves related changes to plant-specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C 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 DCD Tier 1 material departures. Specifically, the requested amendment proposes to depart from Tier 2 information in UFSAR Subsection 8.3.2.4 describing raceway and cable routing criteria and hazard protection, and involves related changes to plant-specific Tier 1 Table 3.3-6, inspections, tests, analyses, and acceptance criteria information, with corresponding changes to the associated COL Appendix C information.

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

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

Response: No.

Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between raceways. Change 2 reduces the required separation distances between raceways from those documented in [Institute of Electrical and Electronics Engineers (IEEE)] 384-1981. These reduced separation distances are based on specific tests performed on the specified raceway configurations, and the recommendations from those tests contained in the associated report. The NRC staff previously reviewed the descriptions of the ten tests documented in this report, including the ones applicable to the existing UFSAR exceptions, and concluded that they were acceptable, as documented in NUREG-1793, “Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design,” (Initial Report) Subsection 8.3.2.2.

The reduced separation does not adversely impact the ability to safely shutdown the plant, and maintain it shutdown. The referenced test report has shown a failure of a faulted cable will not propagate to a nearby target cable in way that adversely impacts its function.

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.

Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between circuits. Change 2 reduces the required separation distances between circuits from those documented in IEEE 384-1981. This change does not result in a new accident initiator or impact a current accident initiator.

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.

Changes 1, 3 and 4 are clarifications only and do not represent a change to the minimum required separation distance between circuits. Change 2 reduces the required separation distances between circuits from those documented in IEEE 384-1981. These reduced separation distances are based on specific tests performed on the specified raceway configurations, and the recommendations from those tests contained in the associated report. The NRC staff previously reviewed the descriptions of the ten tests documented in this report, including the ones applicable to the existing UFSAR exceptions, and concluded that they were acceptable, as documented in NUREG-1793, “Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design,” (Initial Report) Subsection 8.3.2.2.

The reduced separation does not adversely impact the ability to safely shutdown the plant, and maintain it shutdown. The referenced test report has shown a failure of a faulted cable will not propagate to a nearby target cable in a way that adversely impacts its function.

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 Dixon-Herrity.

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

Date of amendment request: July 10, 2017. A publicly-available version is in ADAMS under Accession No. ML17191B163.

Description of amendment request: The amendments would revise the technical specifications (TSs) by: (1) Adding a Note to the surveillance requirements (SRs) of TS 3.7.7, “Main Turbine Bypass System,” to clarify that the SRs are not required to be met when the limiting condition for operation (LCO) does not require the Main Turbine Bypass System to be operable, (2) clarifying that LCO 3.2.3, “LINEAR HEAT GENERATION RATE (LHGR),” also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deleting an outdated footnote for LCO 3.2.3.

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

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

Response: No.

The proposed change (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, “LINEAR HEAT GENERATION RATE” also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. The proposed change does not affect the requirement to meet the LCO, nor does it affect the requirements to perform the SRs when the Main Turbine Bypass System is being used to meet the LCO. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The current safety analysis evaluation is unaffected by this proposed change. The change regarding the outdated footnote has no effect on the actual TS requirements.

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

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

Response: No.

The proposed change (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, “LINEAR HEAT GENERATION RATE” also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The change regarding the outdated footnote has no effect on the actual TS requirements. The current safety analysis evaluation is unaffected by these proposed changes.

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

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

Response: No.

The proposed change (1) adds a Note to the Surveillance Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are not required to be met when the LCO does not require the Main Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, “LINEAR HEAT GENERATION RATE” also has limits for an inoperable Main Turbine Bypass System that are made applicable as specified in the Core Operating Limits Report, and (3) deletes an outdated footnote for LCO 3.2.3. This change simply clarifies the existing allowance to apply the Main Turbine Bypass System inoperable limits to minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) in lieu of the requirement for the Main Turbine Bypass System to be Operable. The applicable safety analyses for TS 3.7.7 is unaffected by this clarification. The change regarding the outdated footnote has no effect on the actual TS requirements.

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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: September 13, 2017. A publicly-available version is in ADAMS under Accession No. ML17256A626.

Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information, and from Technical Specifications as incorporated in Appendix A of the Combined License (COL). Specifically, the proposed changes revise COL Appendix A Technical Specification 3.6.8 to identify the trisodium phosphate (TSP) mass value required in the pH adjustment baskets. The TSP mass value adjusts the pH of the containment water to >7.0 following a postulated accident.

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 activity revises the mass of trisodium phosphate (TSP), which raises the pH of post-accident containment water to 7.0 or greater following a postulated accident. The change to the TSP mass value does not adversely impact the ability to support radionuclide retention with high radioactivity in containment and helps prevent corrosion of containment equipment during long-term floodup conditions. The proposed changes do not adversely impact previously evaluated accidents, because pH control capability is provided to mitigate already postulated accidents. As described in Updated Final Safety Analysis Report (UFSAR) Subsection 15.6.5.3.1.3, the passive core cooling system (PXS) is assumed to provide sufficient TSP to the post-loss-of-coolant accident (LOCA) cooling solution to maintain the pH at greater than or equal to 7.0 following a LOCA. The pH adjustment baskets provide for long-term pH control. Long-term pH control is not adversely impacted as the pH adjustment baskets contain the required amount of TSP to support pH control requirements following a design basis accident (DBA).

No safety-related structure, system, component (SSC) or function is adversely affected by this change. The change does not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. The proposed changes do not involve a change to the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The proposed activity revises the mass of TSP, which raises the pH of containment to 7.0 or greater following a postulated accident. The proposed activity does not create the possibility of a new or different kind of accident as pH adjustment is used to support proper containment chemistry requirements following an accident. The proposed activity does not adversely affect any safety related equipment, and does not add any new interfaces to safety-related SSCs that adversely affect safety functions. No system or design function or equipment qualification is adversely affected by these changes as the changes do not modify any SSCs that prevent safety functions from being performed. The capability to maintain a maximum containment pH below 9.5 is not adversely impacted by these changes. The changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety related equipment.

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

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

Response: No.

The proposed activity revises the mass of TSP, which raises the pH of containment to 7.0 or greater following a postulated accident. The proposed activity does not affect any other safety-related equipment or fission product barriers. Containment water pH adjustment is not adversely impacted. The requested changes will not adversely affect compliance with 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 requested changes as previously evaluated accidents are not impacted.

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 hazard consideration.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: September 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17272A957.

Description of amendment request: The requested amendment proposes to depart from Tier 2* and associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific DCD Tier 2 information). The requested amendment proposes to depart from UFSAR Tier 2* information regarding resolution of human engineering deficiencies (HEDs) contained in Westinghouse Electric Company's report APP-OCS-GEH-320, “AP1000 Human Factors Engineering Integrated Systems Validation Plan,” which is incorporated by reference into the VEGP Units 3 and 4 UFSAR.

The proposed changes would revise the licensing basis of the combined licenses regarding the process for addressing and re-testing of HEDs identified during the integrated system validation (ISV) as described in Tier 2* document, APPOCS-GEH-320 “AP1000 Human Factors Engineering Integrated System Validation Plan.” APPOCS-GEH-320 references APP-OCS-GEH-420, “Human Factors Engineering Discrepancy Resolution Process,” which defines the process for tracking, resolution, and closure of HEDs. The proposed changes to APP-OCS-GEH-320 do not impact APP-OCS-GEH-420.

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 Integrated System Validation (ISV) provides a comprehensive human performance-based assessment of the design of the AP1000 Human-System Interface (HSI) resources, based on their realistic operation within a simulator driven Main Control Room (MCR). The ISV is part of the overall AP1000 Human Factors Engineering (HFE) program. The changes to APP-OCS-GEH-320, which is incorporated by reference into the UFSAR, clarify the resources and methodology used during re-testing performed to verify the effectiveness of Human Engineering Deficiency (HED) resolution. The ISV Plan does not affect the plant itself. Changing APP-OCS-GEH-320 and the UFSAR does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, component (SSC) or function is adversely affected. The changes neither involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The changes to APP-OCS-GEH-320 and the VEGP 3 and 4 UFSAR affect only the testing and validation of the MCR design and HSI using a plant simulator. Therefore, the changes do not affect the safety-related equipment itself, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected. No system or design function or equipment qualification is adversely affected by the changes. 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 would result in significant fuel cladding failures. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety related equipment.

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

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

Response: No.

The changes to APP-OCS-GEH-320 and the UFSAR affect the testing and validation of the MCR design and HSI using a plant simulator. Therefore, the changes do not affect the assessments or the plant itself. These changes do not affect safety-related equipment or equipment whose failure could initiate an accident, nor does it adversely interface with safety-related equipment or fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change.

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 hazard consideration.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: September 20, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17265A434.

Description of amendment request: The amendments would revise technical specification (TS) requirements related to “operations with a potential for draining the reactor vessel” (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires RPV water level to be greater than the top of active irradiated fuel. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control,” dated December 20, 2016.

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 replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

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 automatic initiation of an ECCS subsystem and the Control Room Emergency Outside Air Supply (CREOAS) system. 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 are not needed to adequately respond to a draining event.

The administrative update to delete expired completion time notes is purely administrative in nature.

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

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

The administrative update to delete expired completion time notes is purely administrative in nature.

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.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

The administrative update to delete expired completion time notes is purely administrative in nature.

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

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

Attorney for licensee: Damon D. Obie, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101.

NRC Branch Chief: James G. Danna.

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County, Alabama

Date of amendment request: August 15, 2017. A publicly-available version is in ADAMS under Accession No. ML17228A490.

Description of amendment request: The amendments would revise the BFN, Units 1, 2, and 3 Technical Specification (TS) 5.5.12, “Primary Containment Leakage Rate Testing Program,” by adopting Nuclear Energy Institute (NEI) 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” as the implementation document for the performance-based Option B of 10 CFR part 50, Appendix J. The proposed changes permanently extend the Type A containment integrated leak rate testing (ILRT) interval from 10 years to 15 years and the Type C local leakage rate testing (LLRT) intervals from 60 months to 75 months.

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

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

Response: No.

The proposed revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current Type A test interval of 10 years would be extended to 15 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident because research documented in NUREG-1493, “Performance-Based Containment System Leakage Testing Requirements” [“Performance-Based Containment Leak-Test Program”], September 1995, has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. NUREG-1493 concluded that reducing the Type A testing frequency to one per 20 years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3) shows leakage to be below acceptance criteria, indicating a very leak tight containment. Inspections required by the ASME Code [American Society of Mechanical Engineers Boiler and Press Vessel Code] Section Xl (Subsection IWE) and Maintenance Rule monitoring (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants”) are performed in order to identify indications of containment degradation that could affect that leak tightness. Types B and C testing required by TSs will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test interval extension will not represent a significant increase in the consequences of an accident.

The proposed amendment involves changes to the BFN, Units 1, 2, and 3, 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the units are operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, performance-based leakage testing program. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses. The potential consequences of extending the ILRT interval from 10 years to 15 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem [roentgen equivalent man] per year resulting from design basis accidents was estimated to be very small, and the increase in the LERF [large early release frequency] resulting from the proposed change was determined to be within the guidelines published in NRC RG [Regulatory Guide] 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. TVA has determined that the increase in CCFP [conditional containment failure probability] due to the proposed change would be very small.

Based on the above discussions, the proposed changes do not involve an 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 revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3). The proposed extension to Type A and Type C test intervals does not create the possibility of a new or different type of accident because there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident.

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

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

Response: No.

The proposed revision to TS 5.5.12 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR part 50, Appendix J, Option B, ILRT) and a 75-month interval for Type C testing (10 CFR part 50, Appendix J, Option B, LLRT). The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (performed November 19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, Unit 3). The proposed extension to Type A testing will not significantly reduce the margin of safety. NUREG-1493, “Performance-Based Containment System Leakage Testing Requirements” [“Performance-Based Containment Leak-Test Program”], September 1995, generic study of the effects of extending containment leakage testing, found that a 20 year extension to Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1% to the individual risk and that the decrease in Type A testing frequency would have a minimal effect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the ASME Code Section Xl (Subsection IWE) and maintenance rule monitoring (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants”) will further reduce the risk of a containment leakage path going undetected.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, performance-based leakage testing program, and establishes a 15-year interval for the performance of the primary containment ILRT and a 75-month interval for Type C testing. The amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the 10 CFR part 50, Appendix J Testing Program Plan, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant safety analyses is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, B, and C containment leakage tests will continue to be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A.

Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by an ILRT. This ensures that evidence of containment structural degradation is identified in a timely manner. Furthermore, a risk assessment using the current BFN, Units 1, 2, and 3, PRA [probabilistic risk assessment] model concluded that extending the ILRT test interval from 10 years to 15 years results in a very small change to the BFN, Units 1, 2, and 3, risk profile.

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

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.

NRC Branch Chief: Undine Shoop.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2 (SQN), Hamilton County, Tennessee

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee

Date of amendment request: August 7, 2017. A publicly-available version is in ADAMS under Accession No. ML17219A505.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” and TS 3.3.1, “Reactor Trip System (RTS) Instrumentation,” to avoid confusion as to when an incore power distribution measurement for QPTR is required. The amendment would also revise the WBN TSs for consistency with the existing SQN TSs and Westinghouse Standard TSs in NUREG-1431, Revision 4.

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

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

Response: No.

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

Therefore, the changes do not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not result in a change in the manner in which the reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. There are no design changes associated with the change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as providing primary or secondary protection, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis.

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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.

NRC Branch Chief: Undine Shoop.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment requests: December 15, 2016.

Brief description of amendments: The amendments modified Technical Specification (TS) 3.4.10, “Pressurizer Safety Valves,” TS 3.7.4, “Steam Generator Power Operated Relief Valves (SG PORVs),” and TS 3.7.6, “Condensate Storage System,” to revise the Completion Times for Limiting Condition for Operation (LCO) of TS LCO 3.4.10 Required Action B.2, TS LCO 3.7.4 Required Action C.2, and TS LCO 3.7.6 Required Action B.2 from 12 to 24 hours. The proposed changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A, Revision 1, “Provide Consistent Completion Time to Reach MODE 4.”

Date of issuance: October 23, 2017.

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

Amendment Nos.: 294 (Unit 1) and 290 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17254A144; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: April 25, 2017 (82 FR 19099).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 23, 2017.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment requests: December 15, 2016.

Brief description of amendments: The amendments modified technical specification (TS) limiting condition for operation (LCO) 3.7.5, “Auxiliary Feedwater (AFW) System,” Condition A and Required Action A.1. Condition A was revised to include the situation when one turbine-driven AFW pump is inoperable in MODE 3, immediately following a refueling outage, only applicable if MODE 2 has not been entered following the refueling outage. Required Action A.1 was revised to include the turbine-driven AFW addition to Condition A. The amendments are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-340-A, Revision 3, “Allow 7 day Completion Time for a turbine-driven AFW pump inoperable.”

Date of issuance: October 23, 2017.

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

Amendment Nos.: 295 (Unit 1) and 291 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17257A297; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: April 25, 2017 (82 FR 19100).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 23, 2017.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment requests: December 15, 2016.

Brief description of amendments: The amendments revised Technical Specification 3.1.2, “Core Reactivity,” to revise the Completion Times of Required Actions A.1 and A.2 from 72 hours to 7 days. This proposed change is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-142-A, Revision 0, “Increase the Completion Time when the Core Reactivity Balance is Not Within Limit.”

Date of issuance: October 23, 2017.

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

Amendment Nos.: 296 (Unit 1) and 292 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17261B290; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: April 11, 2017 (82 FR 17457).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 23, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment requests: January 11, 2017.

Brief description of amendments: The amendments modified Technical Specification (TS) 3.8.1, “AC Sources—Operating,” to allow greater flexibility in performing Surveillance Requirements (SRs) by modifying Mode restriction notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. This proposed change was consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-283-A, Revision 3, “Modify Section 3.8 Mode Restriction Notes.”

Date of issuance: October 25, 2017.

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

Amendment Nos.: 300 (Unit 1) and 279 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A055; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications.

Date of initial notice in Federal Register: May 23, 2017 (82 FR 23620).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 25, 2017.

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 nor the license amendment request.

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 requests: January 11, 2017.

Brief description of amendments: The amendments modified Technical Specification (TS) 3.1.8, “PHYSICS TESTS Exceptions,” to allow the numbers of channels required by the Limiting Condition for Operation (LCO) section of TS 3.3.1, “Reactor Trip System (RTS) Instrumentation,” to be reduced from “4” to “3” to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. This proposed change is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-315-A, Revision 0, “Reduce plant trips due to spurious signals to the NIS [Nuclear Instrumentation System] during physics testing.”

Date of issuance: October 25, 2017.

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

Amendment Nos.: 301 (Unit 1) and 280 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17261B218; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications.

Date of initial notice in Federal Register: May 23, 2017 (82 FR 23621).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 25, 2017.

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

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 requests: January 11, 2017.

Brief description of amendments: The amendments modify the limiting condition for operation (LCO) Required Action B.2 for Technical Specification (TS) 3.4.10, “Pressurizer Safety Valves,” LCO Required Action C.2 for TS 3.7.4, “Steam Generator Power Operated Relief Valves (SG PORVs),” and LCO Required Action G.1 for TS 3.4.12, “Low Temperature Overpressure Protection (LTOP) System.” Specifically, the Completion Times are revised from 12 hours to 24 hours for TS LCO 3.4.10, Required Action B.2, and TS LCO 3.7.4, Required Action C.2; and from 8 hours to 12 hours for TS LCO 3.4.12, Required Action G.1. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A, Revision 1, “Provide Consistent Completion Time to Reach MODE 4.”

Date of issuance: October 31, 2017.

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

Amendment Nos.: 302 (Unit 1) and 281 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A198; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: May 23, 2017 (82 FR 23622).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017.

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

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 requests: January 11, 2017.

Brief description of amendments: The amendments modify Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System,” Limiting Condition for Operation (LCO) Condition A and Required Action A.1. The proposed changes modify Condition A to expand the condition to include when one turbine driven AFW pump is inoperable in MODE 3. This expanded condition is applicable immediately following a refueling outage and only if MODE 2 has not been entered. Required Action A.1 is revised to state “affected equipment” as opposed to “steam supply” as a result of the addition of the turbine driven AFW pump to Condition A. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-340-A, Revision 3, “Allow 7 day Completion Time for a turbine-driven AFW pump inoperable.”

Date of issuance: October 31, 2017.

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

Amendment Nos.: 304 (Unit 1) and 283 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17277A313; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the renewed facility operating licenses and technical specifications.

Date of initial notice in Federal Register: May 23, 2017 (82 FR 23621).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017.

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

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

Date of application for amendments: January 11, 2017.

Brief description of amendments: The amendments modify Technical Specification (TS) Limiting Condition for Operation (LCO) 3.9.6, “Residual Heat Removal (RHR) and Coolant Circulation—Low Water Level,” to add a note which allows all RHR pumps to be secured for less than or equal to 15 minutes to support the switching of the shutdown cooling loops from one train to another. The changes are consistent with Technical Specifications Task Force (TSTF) Travelers TSTF-349-A, Revision 1, “Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation,” TSTF-361-A, Revision 2, “Allow standby [Shutdown Cooling] SDC/RHR/[Decay Heat Removal] DHR loop to [be] inoperable to support testing,” and TSTF-438-A, Revision 0, “Clarify Exception Notes to be Consistent with the Requirement Being Excepted.”

Date of issuance: October 31, 2017.

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

Amendment Nos.: Unit 1—303; Unit 2—282. A publicly-available version is in ADAMS under Accession No. ML17271A034; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: May 23, 2017 (82 FR 23623).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 31, 2017.

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

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: December 2, 2016, as supplemented by letters dated April 25, May 22, and October 2, 2017.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to (1) relocate cycle-specific parameters to the Core Operating Limits Report (COLR) consistent with Technical Specification Task Force (TSTF)-339, “Relocate TS Parameters to COLR;” (2) delete duplicate reporting requirements in the Administrative Section of TSs consistent with TSTF-5, “Delete Safety Limit Violation Notification Requirements,” Revision 1; and (3) delete reference to plant procedure PLP-6, “Technical Specification Equipment List Program and Core Operating Limits Report,” in TSs as it pertains to the COLR.

Date of issuance: November 6, 2017.

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

Amendment No.: 161. A publicly-available version is in ADAMS under Accession No. ML17250A202; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: February 14, 2017 (82 FR 10595). The supplemental letters dated April 25, May 22, and October 2, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 6, 2017.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of amendment request: November 8, 2016, as supplemented by letter dated July 11, 2017.

Brief description of amendment: The amendment would, on a one-time basis, extend the completion time from 7 days to 14 days for the Residual Heat Removal Train A subsystem to operable status associated with Technical Specification (TS) 3.5.1, “ECCS [Emergency Core Cooling System]—Operating”; TS 3.6.1.5, “Residual Heat Removal (RHR) Drywell Spray”; and TS 3.6.2.3, “Residual Heat Removal (RHR) Suppression Pool Cooling.” This amendment will be used to support preventive maintenance, which replaces the RHR Train A subsystem's pump and motor.

Date of issuance: October 30, 2017.

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

Amendment No.: 245. A publicly-available version is in ADAMS under Accession No. ML17290A127; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-21: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 14, 2017 (82 FR 10596). The supplemental letter dated July 11, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 30, 2017.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 2, York and Lancaster Counties, Pennsylvania

Date of amendment request: May 19, 2017, as supplemented by letter dated August 29, 2017.

Brief description of amendment: The amendment revised the Technical Specifications to decrease the number of safety relief valves and safety valves required to be operable when operating at a power level less than or equal to 3,358 megawatts thermal. This change is applicable only to the current Cycle 22 that is scheduled to end in October 2018.

Date of issuance: October 25, 2017.

Effective date: As of the date of issuance and shall be implemented within 5 days.

Amendment No.: 315. A publicly-available version is in ADAMS under Accession No. ML17249A151; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-44: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 5, 2017 (82 FR 31094). The supplemental letter dated August 29, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 25, 2017.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: October 27, 2016, as supplemented by the letters dated July 28, 2017, August 30, 2017, and October 19, 2017.

Brief description of amendments: The amendments revised the suppression pool swell design analysis. The new analysis utilizes a different computer code and incorporates different analysis assumptions than the current analysis. The changes are necessary because the current design analysis determining the suppression pool swell response to a loss-of-coolant accident was determined to be non-conservative.

These changes to the suppression pool swell design analysis do not require any changes to the LSCS Technical Specifications. Changes to the LSCS updated final safety analysis report related to changes to the suppression pool swell design analysis shall be made in accordance with 10 CFR 50.71(e) based on the NRC approval of these changes.

Date of issuance: October 30, 2017.

Effective date: These license amendments are effective as of the date of its issuance and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: 225 for NPF-11 and 211 for NPF-18. A publicly-available version is in ADAMS under Accession No. ML17257A304; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments approved to revise the LSCS updated final safety analysis report related to changes to the suppression pool swell design analysis and the Licenses.

Date of initial notice in Federal Register: March 8, 2017 (82 FR 13022). The supplements dated July 28, 2017, August 30, 2017, and October 19, 2017, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 30, 2017.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station (Nine Mile Point), Unit 2, Oswego County, New York

Date of amendment request: December 13, 2016, as supplemented by letter dated February 17, 2017.

Brief description of amendment: The amendment revised the Nine Mile Point, Unit 2, Technical Specification (TS) safety limit (SL) to increase the low pressure isolation setpoint allowable value, which will result in earlier main steam line isolation. The revised main steam line low pressure isolation capability and the revised SL are intended to ensure that Nine Mile Point, Unit 2, remains within the TS SLs in the event of a pressure regulator failure maximum demand transient.

Date of issuance: October 31, 2017.

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

Amendment No.: 164. A publicly-available version is in ADAMS under Accession No. ML17268A263; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: March 28, 2017 (82 FR 15381). The supplemental letter dated February 17, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 31, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 23, 2017, as supplemented by letter dated July 3, 2017.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) by limiting the MODE of applicability for the Reactor Protection System, Startup, and Operating Rate of Change of Power—High, functional unit trip. Additionally, the amendments added new Limiting Condition for Operation (LCO) 3.0.5 and relatedly modified LCO 3.0.1 and LCO 3.0.2, to provide for placing inoperable equipment under administrative control for the purpose of conducting testing required to demonstrate OPERABILITY.

Date of issuance: November 2, 2017.

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

Amendment Nos.: 243 and 194. A publicly-available version is in ADAMS under Accession No. ML17257A015; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: March 28, 2017 (82 FR 15383). The supplemental letter dated July 3, 2017, provided additional information that expanded the scope of the application as originally noticed and changed the NRC staff's original proposed no significant hazards consideration (NSHC) determination as published in the Federal Register. Accordingly, the NRC published a second proposed no significant hazards consideration determination in the Federal Register on September 12, 2017 (82 FR 42849). This notice superseded the original notice in its entirety. It also provided an opportunity to request a hearing by November 13, 2017, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendments.

The Commission's related evaluation of the amendments and final NSHC are contained in a Safety Evaluation dated November 2, 2017.

No significant hazards consideration comments received: No.

Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

Date of amendment request: December 21, 2016.

Brief description of amendments: The amendments modify the Technical Specifications by deleting high-range noble gas effluent monitors' requirements and relocating the requirements to the Turkey Point Offsite Dose Calculation Manual.

Date of issuance: October 26, 2017.

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

Amendment Nos: 277 and 272. A publicly-available version is in ADAMS under Accession No. ML17228A563. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: March 14, 2017 (82 FR 13666).

The Commission's related evaluation of the amendments is contained in a safety evaluation dated October 26, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 11, 2017.

Brief description of amendments: The amendments request an extension to the time to achieve full compliance with 10 CFR 50.48(c), National Fire Protection Association (NPFA) 805, from November 6, 2017, to the conclusion of the FNP, Unit 1, Spring 2018 Refueling Outage (1R28). The amendments update Attachment S, “Modification and Implementation Items”; of the previously approved NFPA-805 amendment.

Date of issuance: November 1, 2017.

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

Amendment Nos.: 215 (Unit 1) and 212 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17269A166; 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: August 29, 2017 (82 FR 41059).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 1, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 14th day of November 2017.

For the Nuclear Regulatory Commission.

Kathryn M. Brock,

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