83 Fed. Reg. 6218 (Feb. 13, 2018)

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

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Federal RegisterFeb 13, 2018

83 Fed. Reg. 6218 (Feb. 13, 2018)

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 January 13, 2018, to January 29, 2018. The last biweekly notice was published on January 30, 2018.

DATES:

Comments must be filed by March 15, 2018. A request for a hearing must be filed by April 16, 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 website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0021. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-9127; email: Jennifer.Borges@nrc.gov. 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: TWFN-3-D1, 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: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2018-0021, 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 website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0021.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in 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-2018-0021, 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 website 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 website 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 hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public website 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 website 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 website at http://www.nrc.gov/site-help/e-submittals.html,, by email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC 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.

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

Date of amendment request: November 15, 2017. A publicly available version is in ADAMS under Accession No. ML17331A484.

Description of amendment request: The amendments would revise fire protection license condition 2.B.(6) to allow, as a performance-based method, certain currently-installed thermal insulation materials to be retained and allow future use of these insulation materials in limited applications subject to appropriate engineering reviews and controls, as a deviation from the National Fire Protection Association Standard 805, Chapter 3, Section 3.3, Prevention.

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.

A fire hazards evaluation was performed for the areas of the plant where the identified insulation materials are installed. The fire hazards evaluation demonstrates that these materials do not contribute appreciably to the spread of fire, nor represent a secondary combustible beyond those currently analyzed in the Fire Probabilistic Risk Analysis (FPRA) due to the limited applications where these materials are installed. Therefore, it is concluded that this 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 identified installations of the insulation materials were evaluated against the fire scenarios supporting the FPRA. In all instances, the supporting analyses and existing fire scenarios were found to be bounding. Expanded zones of fire influence would not fail additional FPRA targets, or there were no FPRA credited targets in the area. Therefore, it is concluded that this 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 limited installations of the insulation materials do not compromise post-fire safe shutdown capability as previously designed, reviewed, and considered. Essential fire protection safety functions are maintained and are capable of being performed. Because the insulation materials do not compromise post-fire safe shutdown capability as previously designed, reviewed, and considered, it is concluded that this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Undine Shoop.

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Wake County, North Carolina

Date of amendment request: October 19, 2017, as supplemented by letter dated January 11, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML17292B648 and ML18011A911, respectively.

Description of amendment request: The amendment would revise the HNP Updated Final Safety Analysis Report (UFSAR) to incorporate the Tornado Missile Risk Evaluator (TMRE) Methodology contained in Nuclear Energy Institute (NEI) 17-02, Revision 1, “Tornado Missile Risk (TMRE) Industry Guidance Document,” September 2017 (ADAMS Accession No. ML17268A036). This methodology can only be applied to discovered conditions where tornado missile protection is not currently provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.

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

Response: No.

The proposed amendment does not involve an increase in the probability of an accident previously evaluated. The relevant accident previously evaluated is a Design Basis Tornado impacting the HNP site. The probability of a Design Basis Tornado is driven by external factors and is not affected by the proposed amendment. There are no changes required to any of the previously evaluated accidents in the UFSAR.

The proposed amendment does not involve a significant increase in the consequences of a Design Basis Tornado. [The methodology as proposed does not alter any input assumptions or results of the accident analyses. Instead, it reflects a methodology to more realistically evaluate the probability of unacceptable consequences of a Design Basis Tornado. As such, there is no significant increase in the consequence of an accident previously evaluated. A similar consideration would apply in the event additional non-conforming conditions are discovered in the future.]

Therefore, the proposed amendment, for both the conditions described herein and any future application of the methodology, does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment, including any future use of the methodology, will involve no physical changes to the existing plant, so no new malfunctions could create the possibility of a new or different kind of accident. The proposed amendment makes no changes to conditions external to the plant that could create the possibility of a new or different kind of accident. The proposed change will not create the possibility of a new or different kind of accident due to new accident precursors, failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. The existing UFSAR accident analysis will continue to meet requirements for the scope and type of accidents that require analysis.

Therefore, the proposed amendment, for both the conditions described herein and any future application of the methodology, does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed amendment does not exceed or alter any controlling numerical value for a parameter established in the UFSAR or elsewhere in the HNP licensing basis related to design basis or safety limits. The change does not impact any UFSAR Chapter 6 or 15 Safety Analyses, and those analyses remain valid. The change maintains diversity and redundancy as required by regulation or credited in the UFSAR. The change does not reduce defense-in-depth as described in the UFSAR.

Therefore, the proposed amendment, for both the conditions described herein and any future application of the methodology, does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: December 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17340B321.

Description of amendment request: The amendment would revise Technical Specification 3/4.3.2 Table 4.3-2, “Engineered Safety Features Actuation System [ESFAS] Instrumentation Surveillance Requirements.” The amendment would remove from Note 3 of the table the exemption from testing ESFAS relays K114, K305, and K313 at power.

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 remove the Technical Specification Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114 at power. The Technical Specification Table 4.3-2 Note 3 exemption allowed the K305, K313, and K114 to not be tested during power operation. The K305 and K313 relays are associated with the Main Steam Isolation Signal (MSIS). The K114 relays are associated with the Containment Spray Actuation Signal (CSAS). The removal of the exemption from testing during power operation means the impacted relays will be tested more frequently improving the ability to identify failed components.

The removal of the Technical Specification Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114 means these relays will be tested more frequently. This testing frequency will be consistent with the other Technical Specification Table 4.3-2 subgroup relays that do not have an exemption. The probability of an operator choosing the wrong subgroup relay during testing is no different for this change as it is for the existing Technical Specification Table 4.3-2 subgroup relays that are already tested on this same frequency. Thus, there will be no significant increase in the probability of an operator error causing an accident.

The change will also eliminate a potential single failure vulnerability associated with MSIS (relays K305 and K313) and CSAS (relay K114). The elimination of the single failure potential will lower the probability of an accident due to the spurious actuation of the MSIS or CSAS.

The change uses a parallel 2 out of 2 with second 2 out of 2 to ensure no single failure of one actuation path would prevent the other actuation path from completing its function. This ensures no additional failure mode would prevent required equipment from actuating and increasing accident consequences.

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 remove the Technical Specification Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114. The K305, K313, and K114 relays are part of the Engineered Safety Features Actuation System (ESFAS). The ESFAS is used for accident mitigation but an inadvertent actuation could cause an accident. The K305 and K313 relays are associated with the MSIS. The K114 relays are associated with the CSAS. The potential failures of the main steam isolation and containment spray systems have been evaluated in the Waterford 3 Updated Final Safety Analysis Report (UFSAR). The potential accidents are as follows:

Loss of External Load which could be caused by closure of the Main Steam Isolation Valves (MSIVs) (UFSAR Section 15.2, Decrease in Heat Removal by the Secondary System).
Loss of normal Feedwater Flow which could be caused by the closure of the Main Feedwater Isolation Valves (UFSAR Section 15.2, Decrease in Heat Removal by the Secondary System).
Asymmetric Steam Generator Transient which could be caused by the closure of one MSIV (UFSAR Section 15.9.1.1, Asymmetric Steam Generator Transient).
Loss of component cooling to Reactor Coolant Pumps (RCPs) which could be caused by the closure of the RCP Component Coolant Water valve. This could lead to RCP seal assembly damage and the possibility for a loss of coolant accident (UFSAR Section 15.6, Decrease In Reactor Coolant System Inventory).
Inadvertent containment spray which could be caused by actuation of one train of containment spray (UFSAR Section 6.2.1.1.3, Design Evaluation—Containment Pressure—Temperature Analysis).

The removal of the exemption from testing during power operation means the impacted relays will be tested more frequently thereby improving the ability to identify failed components; however, they will be tested at power. The ESFAS K305, K313, and K114 relay test logic is designed to test the relays at power and not actuate the end devices which could adversely impact the plant. Any failures that could actuate plant equipment would continue to be bounded by the existing UFSAR accidents; therefore, no new accident is being created.

The ESFAS is used for accident mitigation. The removal of the exemption from testing during power operation means the impacted relays will be tested more frequently thereby improving the ability to identify failed components. This lowers the possibility of the ESFAS equipment not being available when needed. This also means that with the ESFAS equipment available, this change does not create the possibility of a different kind of 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 proposed change will remove the Technical Specification Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114. The removal of the exemption from testing during power operation means the impacted relays will be tested more frequently thereby improving the ability to identify failed components. The more frequent testing will improve the margin of safety.

The change will also eliminate a potential single failure vulnerability associated with MSIS (relays K305 and K313) and CSAS (relay K114). The elimination of the single failure potential will improve the margin of safety by reducing the potential of an accident due to the spurious actuation of the MSIS or CSAS.

Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: December 13, 2017. A publicly-available version is in ADAMS under Accession No. ML17360A159.

Description of amendment request: The amendments would revise technical specifications (TSs) to adopt Technical Specification Task Force (TSTF)-542, Reactor Pressure Vessel Water Inventory Control (RPV WIC).

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

Response: No.

The proposed change replaces existing TS requirements related to OPDRVs [operations with a potential for draining the reactor vessel] with new requirements on RPV WIC water inventory control] 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 change reduces 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 change reduces 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 secondary containment and/or filtration would be available if needed.

The proposed change reduces or eliminates 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 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 replaces existing TS [technical specification] requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, 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 change is an unexpected draining event. The proposed change does 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 change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change replaces 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 change does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1, Oswego County, New York

Date of amendment request: December 15, 2017. A publicly available version is in ADAMS under Accession No. ML17349A027.

Description of amendment request: The amendment would revise the Nine Mile Point Nuclear Station, Unit 1, Technical Specifications (TSs) by replacing existing requirements related to “operations with a potential for draining the reactor vessel” (OPDRVs) with new requirements on reactor pressure vessel water (RPV) inventory control (WIC). The proposed changes are based on Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control” (ADAMS Accession No. ML16074A448).

1. Does the proposed amendment 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 ensure RPV water level remains above −10 inches indicator scale. Draining of RPV water inventory in the cold shutdown and refueling conditions is not an accident previously evaluated; 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 the cold shutdown or refueling condition 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 −10 inches indicator scale. 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 a Core Spray subsystem to be operable at all times in the cold shutdown and refueling conditions. The change in requirement from two Core Spray subsystems to one Core Spray subsystem in the cold shutdown or refueling conditions 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 a Core Spray subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in the cold shutdown or refueling condition 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 amendment 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 maintain RPV water level above −10 inches indicator scale. 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 change is 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 accident previously evaluated.

3. Does the proposed amendment 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 maintain RPV water level above −10 inches indicator scale. 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 −10 inches indicator scale 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 maintain RPV water level above −10 inches indicator scale 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.

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

NRC Branch Chief: James G. Danna.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: November 10, 2017. A publicly-available version is in ADAMS under Accession No. ML17314A024.

Description of amendment request: The amendment would make changes to the organization, staffing, and training requirements contained in Section 6.0, “Administrative Controls,” of the Three Mile Island Nuclear Station, Unit 1 (TMI-1), Technical Specifications (TSs) and define two new positions for Certified Fuel Handler and Non-Certified Operator in Section 1.0, “Definitions,” to reflect the permanently defueled condition.

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

Response: No.

The proposed changes would not take effect until TMI-1 has permanently ceased operation and certified a permanently defueled condition. The proposed changes would revise the TMI-1 TS by deleting or modifying certain portions of the TS administrative controls described in Section 6.0 of the TS that are no longer applicable to a permanently shutdown and defueled facility. Additionally, the “Certified Fuel Handler” and “Non-Certified Operator” would be added to Section 1.0 of the TS to define these positions that are applicable to permanently shutdown and defueled facility. These changes are administrative in nature.

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

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

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

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

Response: No.

The proposed changes to the TS definitions and administrative controls have no impact on facility plant Structures, Systems, and Components (SSCs) affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the actual handling and storage of spent irradiated fuel. The proposed changes do not result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and TMI-1 will no longer be authorized to operate the reactor.

The proposed changes do not affect systems credited in the accident analyses at TMI-1. The proposed changes will continue to require proper control and monitoring of safety significant parameters and activities.

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

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

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

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

Response: No.

The proposed changes involve TS administrative controls once the TMI-1 facility has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for TMI-1 will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). As a result, the occurrence of certain design basis postulated accidents are no longer considered credible when the reactor is permanently defueled.

The proposed changes are limited to those portions of the administrative TSs that are related to the safe storage and maintenance of spent irradiated fuel. The proposed TS changes do not affect plant design, hardware, system operation, or procedures for accident mitigation systems. There is no change in the established safety margins for these systems. The requirements that are proposed to be added, revised and/or deleted from the TMI-1 TS are not credited in the existing accident analysis for the applicable postulated accidents; therefore, they do not contribute to the margin of safety associated with the accident analysis. Certain postulated design basis accidents (DBAs) involving the reactor are no longer possible because the reactor will be permanently shutdown and defueled and TMI-1 will no longer be authorized to operate the reactor.

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

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

NRC Branch Chief: James G. Danna.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: December 20, 2017. A publicly-available version is in ADAMS under Accession No. ML17355A019.

Description of amendment request: The amendment would revise technical specification (TS) requirements related to direct current (DC) electrical systems, specifically limiting conditions for operation 3.8.4, 3.8.5, and 3.8.6. The proposed amendment would also add a new Battery and Monitoring Maintenance Program to TS Section 5.5, “Programs and Manuals.” The proposed changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.”

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

Response: No.

The proposed changes restructure the Technical Specifications (TS) for the direct current (DC) electrical power system and are consistent with TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.” The proposed changes modify TS Actions relating to battery and battery charger inoperability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Updated Safety Analysis Report (USAR). Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure TS and change surveillances for batteries and chargers to incorporate the updates included in TSTF-500, Revision 2, will maintain the same level of equipment performance required for mitigating accidents assumed in the USAR. Operation in accordance with the proposed TS would ensure that the DC electrical power system is capable of performing its specified safety function as described in the USAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis. The relocation of preventive maintenance surveillances, and certain operating limits and actions, to a licensee-controlled battery monitoring and maintenance program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance that are consistent with industry standards will continue to be performed. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC electrical power system.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the USAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes involve restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the USAR. Rather, the DC electrical power system supports equipment used to mitigate accidents. The proposed changes to restructure the TS and change surveillances for batteries and chargers to incorporate the updates included in TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360,” will maintain the same level of equipment performance required for mitigating accidents assumed in the USAR. Administrative and mechanical controls are in place to ensure the design and operation of the DC systems continues to meet the plant design basis described in the USAR. Therefore, operation of the facility in accordance with this proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment 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 equipment margins will be maintained in accordance with the plant-specific design bases as a result of the proposed changes. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery 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.

TS changes made in accordance with TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360,” maintain the same level of equipment performance stated in the USAR and the current TSs. Therefore, the proposed changes do not involve a significant reduction of safety.

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: David J. Wrona.

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin

Date of amendment request: August 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17243A201.

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 Plants.” The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced.

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.

Attorney for licensee: Steven Hamrick, Managing Attorney—Nuclear Florida Power & Light Company, LAW/WAS, 801 Pennsylvania Ave. NW #220, Washington, DC 20004.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: December 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17339A428.

Description of amendment request: The amendment would revise certain 18-month surveillance requirements previously performed while shut down to be performed during power operations. The amendment would also revise the administrative controls portion of the technical specifications (TSs) to replace plant-specific titles with generic titles and modify TSs 6.1.2, 6.2.2, 6.2.4, and Table 6.2-1 to be consistent with NUREG-1431, “Standard Technical Specifications, Westinghouse Plants.”

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

Response: No.

The technical specification (TS) surveillance requirements and administrative controls associated with the proposed changes to the TS are not initiators of any accidents previously evaluated, so the probability of accidents previously evaluated is unaffected by the proposed changes. The proposed change does not alter the design, function, or operation of any plant structure, system, or component (SSC). The capability of any operable TS-required SSC to perform its specified safety function is not impacted by the proposed change. As a result, the outcomes of accidents previously evaluated are unaffected. Therefore, the proposed changes do not result in a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

No physical changes are made to the plant, so no new causal mechanisms are introduced. Therefore, the proposed changes to the TS do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

The ability of any operable SSC to perform its designated safety function is unaffected by the proposed changes. The proposed changes do not alter any safety analyses assumptions, safety limits, limiting safety system settings, or method of operating the plant. The changes do not adversely affect plant operating margins or the reliability of equipment credited in the safety analyses. With the proposed change, each DC electrical train remains fully capable of performing its safety function. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

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

NRC Branch Chief: James G. Danna.

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: July 28, 2017, as supplemented by January 23, 2108, letter. Publicly-available versions are in ADAMS under Accession No. ML17209A755, and ML18023A440, respectively.

Description of amendment request: The requested amendment proposes changes to combined license Appendix A, plant-specific Technical Specifications (TS) to make them consistent with the remainder of the design, licensing basis, and the TS. The U.S. Nuclear Regulatory Commission (NRC) staff previously noticed this amendment request in the Federal Register on December 5, 2017 (82 FR 57473). However, due to administrative errors that were inadvertently introduced, the NRC staff is noticing this amendment request again.

An evaluation to determine whether or not a significant hazards consideration is involved with the proposed amendment was completed by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below. However, to provide for ease of review, similar changes have been grouped into categories to facilitate the significant hazards evaluations required by 10 CFR 50.92. Generic significant hazards evaluations are provided for the More Restrictive Changes and a specific significant hazards evaluation for each Clarification or Less Restrictive change. In regards to obvious editorial or administrative changes (e.g., formatting, page rolls, punctuation, etc.), an explicit discussion was not always provided, but is considered to be addressed by the applicable generic significant hazards evaluation.

Valuation for More Restrictive Changes

This generic category include changes that impose additional requirements, decrease allowed outage times, increase the Frequency of Surveillances, impose additional Surveillances, increase the scope of Specifications to include additional plant equipment, broaden the Applicability of Specifications, or provide additional actions. These changes have been evaluated to not be detrimental to plant safety.

More restrictive changes are proposed only when such changes are consistent with the current Vogtle Electric Generating Plant, Units 3 and 4 (VEGP) licensing basis; the applicable VEGP safety analyses; and good engineering practice such that the availability and reliability of the affected equipment is not reduced.

Changes to the Technical Specifications (TS) requirements categorized as More Restrictive are annotated with an “MR” in Section 2 Discussion of Change (DOC). This affects TS changes L05 and L08.

Southern Nuclear Operating Company (SNC) proposes to amend the VEGP TS. SNC has evaluated each of the proposed TS changes identified as More Restrictive in accordance with the criteria set forth in 10 CFR 50.92, “Issuance of amendment,” and has determined that the proposed changes do not involve a significant hazards consideration. This significant hazards consideration is applicable to each More Restrictive change identified in Section 2.

The basis for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92(c). The criteria and conclusions of the evaluation are 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 provide more stringent TS requirements. These more stringent requirements impose greater operational control and conservatism, and as a result, do not result in operations that significantly increase the probability of initiating an analyzed event, and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. 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 do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed changes do impose different Technical Specification requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The imposition of more restrictive requirements either has no effect on or increases a margin of plant safety. As provided in the discussion of change, each change in this category is, by definition, providing additional restrictions to enhance plant safety. The changes maintain requirements within the safety analyses and licensing basis. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Evaluation for Clarification Changes

This category consists of technical changes which revise existing requirements such that the design and operation of a system correctly reflects how the LCO is applied and how the Action or Surveillance Requirement (SR) is carried out. This adds detail and clarity to the Technical Specifications (TS) in operating the applicable portions of the as designed and licensed plant.

Technical changes to the TS requirements categorized as “Clarification” are identified with an “CL” and an individual number in Section 2 Discussion of Change (DOC).

Southern Nuclear Operating Company (SNC) proposes to amend the Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical Specifications. SNC has evaluated each of the proposed technical changes identified as “Clarification” individually in accordance with the criteria set forth in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

L09 SNC proposes to amend TS 3.3.19 Diverse Actuation System Manual Controls, Note (c) in Table 3.3.19-1 to “With upper internals in place.”

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed 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 does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The change applies to a Diverse Actuation System (DAS) Manual Controls Mode 6 note for operability of the Automatic Depressurization System (ADS) Stage 4 valves that involves revising the note from reactor internals in place to upper internals in place. In accordance with Limiting Condition for Operation (LCO) 3.4.13 ADS—Shutdown, Reactor Coolant System (RCS) Open Applicability and TS 3.3.9, Engineered Safeguards Actuation System Instrumentation, Function 7, the ADS Stage 4 valves are not required to be operable in MODE 6 with the upper internals removed. However, the reactor internals would still be present. The change involves clarification of the note (with no change in required system or device function), such that the appropriate configuration in Mode 6 would be in place and would not conflict with TS 3.4.13 or TS 3.3.9. The revised note is not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised note and associated requirements and actions are no different than the consequences of the same accident during the existing ones. As a result, the consequences of an accident previously evaluated are not affected by this change.

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 clarifies TS requirements for the DAS manual control ADS Stage 4 valves such that they would be in agreement with the requirements set forth for the ADS in RCS Shutdown Mode 6. However, the proposed change does not involve a physical alteration of the plant as described in the [Updated Final Safety Analysis Report (UFSAR)]. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the condition for the manual control of ADS Stage 4 actuation switches in Mode 6 has changed, no action is made less restrictive than currently approved for any associated actuated device inoperability. As such, there is no significant reduction in a margin of safety.

L10 SNC proposes to amend current TS 3.5.4, “Passive Residual Heat Removal Heat Exchanger PRHR HX—Operating,” Surveillance Requirement (SR) 3.5.4.6 to: Verify both PRHR HX air operated outlet valves stroke open and both IRWST gutter isolation valves stroke closed.

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The change involves correcting an existing surveillance requirement (with no change in required system or device function), such that the surveillance requirement complies with the In-Containment Refueling Water Storage Tank (IRWST) Gutter Isolation valve design and the Passive Residual Heat Removal (PRHR) Heat Exchanger (HX) outlet isolation valve design. Revised surveillance requirement presentation and compliance with TS actions are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised surveillance requirement are no different than the consequences of the same accident during the existing one. As a result, the consequences of an accident previously evaluated are not affected by this change.

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 clarifies the surveillance requirement such that it agrees with the IRWST and PRHR HX isolation valve design. However, the proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the surveillance requirement has changed for the IRWST and PRHR HX isolation valves, no action is made less restrictive than currently approved for any associated actuated device inoperability. As such, there is no significant reduction in a margin of safety.

10 CFR 50.92 Evaluations for Less Restrictive Changes

This category consists of technical changes which revise existing requirements such that more restoration time is provided, fewer compensatory measures are needed, unnecessary Surveillance Requirements (SR) are deleted, or less restrictive surveillance requirements are required. This would also include unnecessary requirements which are deleted from the Technical Specifications (TS) and other technical changes that do not fit a generic category. These changes are evaluated individually.

Technical changes to the TS requirements categorized as “Less Restrictive” are identified with an “LR” and an individual number in Section 2 Discussion of Change (DOC).

Southern Nuclear Operating Company (SNC) proposes to amend the Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical Specifications. SNC has evaluated each of the proposed technical changes identified as “Less Restrictive” individually in accordance with the criteria set forth in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

L01 SNC proposes to amend TS 1.1 Definitions—Shutdown Margin by:

Changing Shutdown Margin (SDM) definition c. “In MODE 2 with keff<1.0 and MODES 3, 4, and 5, the worth of fully inserted Gray Rod Cluster Assemblies (GRCAs) will be included in the SDM calculation.” to “In MODE 2 with keff<1.0 and in MODES 3, 4, and 5, the worth of the verified fully inserted Gray Rod Cluster Assemblies (GRCAs) which have passed the acceptance criteria for GRCA bank worth measurements performed during startup physics testing may be included in the SDM calculation.”

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The change proposed involves re-defining whether the worth of the Gray Rod Cluster Assemblies (GRCAs) should be included in MODE 2 with keff<1.0 and Modes 3, 4, and 5 when calculating the appropriate Shutdown Margin (SDM). The worth of the GRCAs for MODE 2 with keff<1.0 and Modes 3, 4, and 5 is not credited in the safety analyses as stated in the NRC Safety Evaluation Report (SER) “Westinghouse Electric Company's Final Topical Report Safety Evaluation For WCAP-16943, “Enhanced Gray Rod Cluster Assembly Rodlet Design,” Section 3.0 for ensuring adequate SDM exists.

The change involves revising the existing SDM definition (with no change in required system or device function), such that a more appropriate, albeit less restrictive, definition would be applied when calculating SDM. The revised SDM definition is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised definition requirements are no different than the consequences of the same accident during the existing one. As a result, the consequences of an accident previously evaluated are not affected by this change.

This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 removes the requirement to include the worth of the GRCAs when calculating the SDM because they are not credited for SDM in MODE 2 with keff<1.0 and in MODES 3, 4, and 5. The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the SDM calculation defined is made less restrictive by eliminating the worth of the GRCAs in MODE 2 with keff<1.0 and in MODES 3, 4, and 5, no credit is taken in the safety analyses for including their worth as discussed in the NRC Safety Evaluation Report (SER) “Westinghouse Electric Company's Final Topical Report Safety Evaluation For WCAP-16943, “Enhanced Gray Rod Cluster Assembly Rodlet Design,” Section 3.0. As such, there is no significant reduction in a margin of safety.

L02 SNC proposes to amend TS 3.1.4 Rod Group Alignment Limits by:

L02A. Change Limiting Condition of Operation (LCO) from “All shutdown and control rods shall be OPERABLE.” to “Each rod cluster control assembly (RCCA) shall be OPERABLE.”

L02B. Change LCO AND statement from “Individual indicated rod positions shall be within 12 steps of their group step counter demand position.” to “Individual indicated rod positions of each RCCA and Gray Rod Cluster Assembly shall be within their 12 steps of their group step counter demand position.”

L02C. Delete LCO 3.1.4 note.

L02D. Change Action Condition A from “one or more rod(s) inoperable.” to where it now applies to “One or more RCCA(s) inoperable.”

L02E. Acronym defined in change to Required Action B.1 Completion Time from “1 hour with the OPDMS not monitoring parameters” to “1 hour with the On-Line Power Distribution Monitoring System not monitoring parameters.”

L02F. Add Required Action B.2.3.1 where the Required Action will be to “Perform SR 3.2.5.1” with a Completion Time of “Once per 12 hours,” OR perform B.2.3, which is renumbered as B.2.3.2.1.

L02G. Delete Required Action B.2.4 Note, and renumber the Required Action to B.2.3.2.2.

L02H. Delete Required Action B.2.5 Note, and renumber the Required Action to B.2.3.2.3.

L02I. Renumber Required Action B.2.6 to B.2.4.

L02J. Change SR 3.1.4.2 Note from “Not applicable to GRCAs” to “Not applicable to Axial Offset (AO) Control Bank RCCAs.”

L02K. Change SR 3.1.4.2 from “Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core ≥10 steps in either direction.” to “Verify rod freedom of movement (trippability) by moving each RCCA not fully inserted in the core ≥10 steps in either direction.”

L02L. Delete the Note to SR 3.1.4.3

L02M. Change SR 3.1.4.3 from “Verify rod drop time of each rod . . .” to “Verify rod drop time of each RCCA . . .”.

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The proposed changes involve revising the existing LCO 3.1.4 operability to be applicable to RCCAs with accompanying changes in actions and surveillance requirements (with no change in required system or device function), such that more appropriate, albeit less restrictive, actions would be applied. The proposed changes involve excluding the Gray Rod Cluster Assemblies (GRCAs) in the LCO 3.1.4 Rod Group Alignments LCO since their trip reactivity worth is not credited in the shutdown margin assessments in MODES 1 and 2, nor required by the design basis to be operable. Only the rod cluster control assemblies (RCCAs) are required to be operable. The maximum rod misalignment is an initial assumption in the safety analyses that directly affects core power distributions and assumption of available shutdown margin (SDM). Since the GRCAs do not have a function to maintain the reactor sub-critical unless they are fully inserted, and the reactor is shut down, operability does not apply to GRCAs like it does for RCCAs in MODES 1 and 2. The design basis function of the GRCAs when the reactor is critical does not include a provision of trip reactivity.

The revised LCO, associated actions and surveillance requirements are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised LCO requirements, associated actions, and surveillance requirements are no different than the consequences of the same accident during the existing ones. As a result, the consequences of an accident previously evaluated are not affected by this change.

The proposed change does not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change.

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 revising the existing LCO 3.1.4 operability to be applicable to RCCAs with accompanying changes in actions and surveillance requirements (with no change in required system or device function), such that more appropriate, albeit less restrictive, actions would be applied. The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the LCO 3.1.4 for Rod Group Alignment Limits is made less restrictive by eliminating the worth of the GRCAs in MODES 1 and 2 with keff ≥1, no credit is taken in the current design basis for including their trip reactivity worth. As such, there is no significant reduction in a margin of safety.

L03 SNC proposes to amend TS 3.1.6 Control Bank Insertion Limits by changing Note 2. from “This LCO is not applicable to Gray Rod Cluster Assembly (GRCA) banks during GRCA bank sequence exchange with On-Line Power Distribution Monitoring System monitoring parameters” to “This LCO is not applicable to Gray Rod Cluster Assembly (GRCA) banks for up to one hour during GRCA bank sequence exchange.”

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The proposed change to TS 3.1.6 Control Bank Insertion Limits Note 2 is to not require On Line Power Distribution System (OPDMS) during GRCA bank sequence exchange and limit the LCO applicability exception for one hour after the insertion or sequence or overlap limits are violated due to the short duration of the sequence exchange. The final mechanical shim (MSHIM) design established that the GRCA bank sequence exchange will best be accomplished by moving both banks at the same time. The entire exchange sequence will only take a few minutes from the time banks begin moving. During this short duration, OPDMS is not suited for real time monitoring relative to the time constant for the vanadium fixed incore detector system. The exchange transient may be completed before the OPDMS detects a significant change in the core radial power distribution. In addition, it is unlikely there would be significant time to take corrective action in response to an OPDMS alarm if one occurred during the exchange.

The revised LCO note exception is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised LCO note exception is no different than the consequences of the same accident during the existing one. As a result, the consequences of an accident previously evaluated are not affected by this change.

The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the proposed change to TS 3.1.6, Note 2 would not require OPDMS be functional during GRCA bank sequence exchange for up to one hour, OPDMS operability is still required by TS 3.2.5 On-Line Power Distribution Monitoring System (OPDMS)—Monitored Parameters. As such, there is no significant reduction in a margin of safety.

L04 SNC proposes to amend TS 3.1.7 Rod Position Indication by deleting Required Action B.2 and renumbering the remaining Condition B Required Actions.

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The proposed change is to remove Required Action B.2 for monitoring and recording Reactor Coolant System (RCS) Tavg (with no change in required system or device function), such that more appropriate, albeit less restrictive, actions would be applied. There are no safety benefits, no acceptance criteria or no actions associated with any trends for recording Tavg. Monitoring Tavg provides no power distribution information for unmonitored rods that isn't already provided by complying with the existing requirements of Condition A, and average coolant temperature provides no indication of changes in shutdown margin.

The revised actions are not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised LCO requirements and actions are no different than the consequences of the same accident during the existing ones. As a result, the consequences of an accident previously evaluated are not affected by this change.

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

Response: No.

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

Response: No.

The proposed change will not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the required actions of LCO 3.1.7 for Rod Position Indication are made less restrictive by deletion of Action B.2 for monitoring Tavg, monitoring Tavg provides no power distribution information for unmonitored rods that aren't already provided by complying with the existing requirements of Condition A. As such, there is no significant reduction in a margin of safety.

L06 SNC proposes to amend TS 3.3.1 “Reactor Trip System Instrumentation,” Table 3.3.1-1 FUNCTION 12, (page 2 of 2), Passive Residual Heat Removal Actuation by deleting SR 3.3.1.9.

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

Response: No.

The proposed change is to delete the Surveillance Requirement (SR) 3.3.1.9 Channel Calibration for the passive residual heat removal (PRHR) reactor trip system actuation. The PRHR reactor trip actuation initiates a reactor trip in the event either of the parallel PRHR discharge valves is not fully closed. The proper adjustment of the valve position indication contact inputs to the breaker position are verified by performance of SR 3.3.1.10 Trip Actuating Device Operational Test (TADOT). The revised surveillance requirements are not an initiator to any accident previously evaluated. The reactor trip from PRHR actuation has not changed, and the proper adjustment of the valve position indication contact inputs continues to be addressed by current SR 3.3.1.10. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised surveillance requirements are no different than the consequences of the same accident during the existing ones. As a result, the consequences of an accident previously evaluated are not affected by this change.

The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated.

Therefore, this 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.

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

Response: No.

The proposed change will not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the surveillance requirements have been made less restrictive, the intent of the deleted surveillance requirement remains covered by an existing surveillance requirement. As such, there is no significant reduction in a margin of safety.

L07 SNC proposes to amend TS, Section 3.3.5, “Reactor Trip System Manual Actuation,” Table 3.3.5-1 “Reactor Trip System Manual Actuation,” Functions 1. Manual Reactor Trip, 2. Safeguards Actuation Input from Engineered Safety Feature Actuation System—Manual and 4. Core Makeup Tank Actuation Input from Engineered Safety Feature Actuation System—Manual for Required Channels to 2 switches.

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

Response: No.

The proposed changes define the required channels operable for manual reactor trip based upon the existing design. Required channels operable are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident with defined number of switches operable for manual reactor trip are no different than the consequences of the same accident using the existing required channels operable. As a result, the consequences of an accident previously evaluated are not affected by this change.

The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated.

Further, the proposed change does 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 change is consistent with the safety analysis assumptions and resultant consequences.

Therefore, this 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.

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

Response: No.

The proposed change to define the required channels operable consistent with the plant design does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. Therefore, there is no significant reduction in a margin of safety.

L11 SNC proposes to amend current TS 3.8.3, “Inverters—Operating,” by changing:

1. Action Condition A. from “One inverter inoperable.” to “One or two inverter(s) within one division inoperable.”

2. Second Note in Required Action A.1 from “Restore inverter to OPERABLE status.” to “Restore inverter(s) to OPERABLE status.”

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The proposed changes to action conditions to explicitly define an inverter division that contains two inoperable inverters is not an accident initiator nor do they impact mitigation of the consequences of any accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a physical alteration of the plant as described in the UFSAR and does not alter the method of operation or control of equipment as described in the UFSAR. The current assumptions in the safety analysis regarding accident initiators and mitigation of accidents are unaffected by this change. Plant equipment remains capable of performing mitigative functions assumed by the accident analysis. No additional failure modes or mechanisms are being introduced and the likelihood of previously analyzed failures remains unchanged.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by this change. Therefore, the consequences of previously analyzed accidents will not increase because of this change. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes to action conditions to explicitly define an inverter division that contains two inoperable inverters does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, that are affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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.

Margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change will not reduce a margin of safety because it has no such effect on any assumption of the safety analyses.

Operation in accordance with the proposed TS operability ensures that the plant response to analyzed events continues to provide the margins of safety assumed by the analysis. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. Therefore, there is no significant reduction in a margin of safety.

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: November 17, 2017. A publicly-available version is in ADAMS under Accession No. ML17321B080.

Description of amendment request: The amendment request proposes changes to combined license (COL) License Condition and changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* and associated Tier 2 information. Specifically, this amendment request involves a change to COL License Condition requirements regarding the Natural Circulation (first plant test) using the steam generators and the Passive Residual Heat Removal Heat Exchanger (first plant test). A COL License Condition is proposed to be revised to include an exception that would allow the requirements of a Technical Specification to be suspended during performance of the Natural Circulation (first plant test) using the steam generators. In addition, a revised Passive Residual Heat Removal Heat Exchanger (first plant test) is proposed to be performed as part of the Power Ascension Testing requirements instead of as part of the Initial Criticality and Low-Power Testing requirements as currently specified in a COL License Condition.

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 systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events. The proposed changes do not adversely affect the ability of the steam generators, applicable reactor trip functions, and the passive residual heat removal heat exchanger to perform the required safety function to remove core decay heat during forced and natural circulation when necessary to prevent exceeding the reactor core and the reactor coolant system design limits, and do not adversely affect the probability of inadvertent operation or failure of the passive residual heat removal heat exchanger. The proposed changes do not result in any increase in probability of an analyzed accident occurring, and maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that the reactor core and the reactor coolant system design limits are not exceeded for events requiring emergency core decay heat removal.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes do not adversely affect the ability of the steam generators, applicable reactor trip functions, and the passive residual heat removal heat exchanger to perform the required safety function to remove core decay heat during forced and natural circulation when necessary to prevent exceeding the reactor core and the reactor coolant system design limits, and do not adversely affect the probability of inadvertent operation or failure of the passive residual heat removal heat exchanger. The proposed changes do not result in the possibility of an accident occurring, and maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that the reactor core and the reactor coolant system design limits are not exceeded for events requiring emergency core decay heat removal.

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 results in significant fuel cladding failures.

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

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

Response: No.

The proposed changes maintain existing safety margins through continued application of the existing requirements of the UFSAR. The proposed changes maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that the reactor core and the reactor coolant system design limits are not exceeded for events requiring emergency core decay heat removal. Therefore, the proposed changes satisfy the same safety functions in accordance with the same requirements as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the requested amendment does not involve a significant reduction in a margin of safety.

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: December 21, 2017. A publicly-available version is in ADAMS under Accession No. ML17355A416.

Description of amendment request: The requested amendment proposes changes to combined license License Condition 2.D by adding a new condition to address the Tier 2* change process. The proposal also requests exemptions from the requirements of 10 CFR part 52, Appendix D, Paragraphs II.F, VIII.B.6.b, and VIII.B.6.c.

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

Response: No.

The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety. Changing the criteria by which departures from Tier 2* information are evaluated to determine if NRC approval is required does not affect the plant itself. Changing these criteria 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 Updated Final Safety Analysis Report (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 proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety. 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 proposed change is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The only impact of this activity is the application of the current Tier 2 departure evaluation process to Tier 2* departures.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: December 21, 2017. A publicly-available version is in ADAMS under Accession No. ML17355A177.

Description of amendment request: The proposed amendment establishes Conditions, Required Actions, and Completion Times in the Technical Specification (TS) 3.75 for the Condition where one steam supply to the turbine driven Auxiliary Feedwater (AFW) pump is inoperable concurrent with an inoperable motor driven AFW train. In addition, this amendment establishes changes to the TS, that establish specific Actions: (1) For when two motor driven AFW trains are inoperable at the same time and; (2) for when the turbine driven AFW train is inoperable either (a) due solely to one inoperable steam supply, or (b) due to reasons other than one inoperable steam supply. The licensee stated that the change is consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler, TSTF-412, Revision 3, “Provide Actions for One Steam Supply to Turbine Driven AFW/EFW Pump Inoperable.” (ADAMS Accession No. ML070100363).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 10.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, by referencing the environmental evaluation included in the model safety evaluation published in the Federal Register on July 17, 2007 (72 FR 39089), which is presented below:

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

Response: No.

The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an initiator of any design basis accident or event, and therefore the proposed changes do not increase the probability of any accident previously evaluated. The proposed changes to address the condition of one or two motor driven AFW/EFW trains inoperable and the turbine driven AFW/EFW train inoperable due to one steam supply inoperable do not change the response of the plant to any accidents.

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

Therefore, the changes do not involve a significant increase in the probability or consequences of any 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 result in a change in the manner in which the AFW/EFW System provides plant protection. The AFW/EFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes. The changes to the Conditions and Required Actions do not change any existing accident scenarios, nor create any new or different accident scenarios.

The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The proposed changes 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. The proposed changes will not result in plant operation in a configuration outside the design basis.

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

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

NRC Branch Chief: Michael T. Markley.

U.S. Department of Transportation, Maritime Administration, Docket No. 50-238, Nuclear Ship Savannah, Baltimore, Maryland

Date of amendment request: October 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17307A036.

Description of amendment request: The amendment would revise the license to remove a condition that prevents dismantling and disposing of the facility without prior approval of the Commission.

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

Response: No.

The proposed changes are administrative and do not involve modification of any plant equipment or affect basic plant operation.

The NSS's reactor is not operational and the level of radioactivity in the NSS has significantly decreased from the levels that existed when the 1976 Possession-only License was issued. No aspect of any of the proposed changes is an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased.

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

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

Response: No.

Both of the proposed changes are administrative and do not involve physical alteration of plant equipment that was not previously allowed by Technical Specifications. These proposed changes do not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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.

Both of the proposed changes are administrative in nature. No margins of safety exist that are relevant to the ship's defueled and partially dismantled reactor. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed changes. The proposed changes involve revising the language of the license to clearly state previously approved changes, and to delete archaic requirements.

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

Advisor for licensee: Erhard W. Koehler, U.S. Department of Transportation, Maritime Administration, 1200 New Jersey Ave. SE, Washington, DC 20590.

NRC Branch Chief: Bruce A. Watson, CHP.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: November 7, 2017. A publicly-available version is in ADAMS under Accession No. ML17317A464.

Description of amendment request: The amendments would revise the Surry Power Station (Surry), Units 1 and 2, Facility Operating License Numbers DPR-32 and DPR-37, respectively, in the form of new License Conditions, and Technical Specification (TS) 3.16, “Emergency Power System,” to allow a one-time extension of the Allowed Outage Time (AOT) in TS 3.16 Action B.2 from 7 days to 21 days. The requested temporary 21-day AOT is needed to replace Reserve Station Service Transformer C (RSST-C) and associated cabling during the Surry Unit 2 fall 2018 refueling outage. The existing RSST-C is original plant equipment and is reaching the end of its dependable service life.

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

Response: No.

The proposed change adds a footnote to TS 3.16, “Emergency Power System,” to allow a one-time extension of the AOT in TS 3.16 Action B.2 from 7 days to 21 days to facilitate the replacement of RSST-C and associated cabling.

During the temporary 21-day AOT, the station emergency buses will continue to be fed from redundant, separate, reliable offsite sources that are capable of supporting the emergency loads under worst-case conditions considering a single failure.

There are two (2) emergency buses for each unit: Buses 1H and 1J (Unit 1), and Buses 2H and 2J (Unit 2). While RSST-C is being replaced during the temporary 21-day AOT, Buses 1J and 2H will continue to be energized from a designated primary offsite source, System (Switchyard) Reserve Transformer (SRT) 4. Buses 1H and 2J will be energized from Main Step-up Transformer 2, which is the Unit 2 designated dependable alternate source.

In both configurations Transfer Bus F is fed through two, in series, transformers.

The normal configuration feeds Transfer Bus F from the 230 kV switchyard via two (2) transformers (SRT-2 and RSST-C) and two (2) breakers. The 230 kV switchyard is connected to ten (10) offsite circuits.
The temporary 21-day AOT configuration feeds Transfer Bus F from the 500 kV switch yard via two (2) transformers (Main Step-up Transformer 2 and Station Service Transformer 2C) and three (3) breakers. The 500 kV switchyard is connected to 3 offsite circuits.

A risk assessment has been performed for the temporary 21-day AOT configuration. The assessment concluded that the probability of a loss of offsite power for the proposed configuration is very low. Thus, the proposed change does not significantly increase the probability of an accident previously evaluated because: (a) The emergency buses continue to be feed from redundant, separate, reliable offsite sources and (b) the effect of the proposed configuration on the probability of a loss of offsite power is very low.

There is no increase in the consequences of an accident because the emergency buses continue to be fed from redundant, separate, reliable offsite circuits and the onsite power sources (i.e., the Emergency Diesel Generators) are unaffected.

The consequences of both a Loss of Offsite Power (LOOP) and a Station Blackout (SBO) have been evaluated in the UFSAR. There is no change in the station responses to a LOOP or an SBO as a result of the extended AOT because RSST-C is not included in designated equipment used in the LOOP and SBO coping strategies.

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

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

Response: No.

The proposed configuration does not result in a change in the manner in which the electrical distribution subsystems downstream of RSST-C provide plant protection. During the temporary AOT (21 days total), the only change is to substitute the reliable Unit 2 designated dependable alternate source for a primary offsite power source for Emergency Buses 1H and 2J. Other sources of offsite and onsite power are unaffected, and other aspects of the offsite and onsite power supplies are unchanged and unaffected.

There are no changes to the other RSSTs or to the supporting systems operating characteristics or conditions.

There is no change in the station responses to a LOOP or an SBO because RSST-C is not included in the designated equipment used in the LOOP and SSO coping strategies.

Therefore, the proposed change does 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 TS change does not affect the acceptance criteria for any analyzed event, nor is there a change to any safety limit. The proposed TS change does not affect any structures, systems or components or their capability to perform their intended functions. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are affected by this change. The proposed change will not result in plant operation in a configuration outside the current design basis as the design basis includes use of the Unit 2 dependable alternate source. The proposed TS change allows use of the Unit 2 dependable alternate power source as the primary source for buses 1H and 2J for a period of up to 21 days. The margin of safety is maintained by maintaining the capability to supply Emergency Buses 1H and 2J with a redundant, separate, reliable offsite power source, and maintaining the onsite power sources in their design basis configuration. Therefore, the proposed change does not involve a significant reduction in margin of safety.

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.

NRC Branch Chief: Michael T. Markley.

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

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

Arizona Public Service Company, et al. (APS), Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, Maricopa County, Arizona

Date of amendment: July 1, 2016, as supplemented by letters dated June 2 and December 15, 2017.

Description of amendment request: The amendments revised the Technical Specifications for PVNGS, Units 1, 2, and 3, to support the implementation of next generation fuel (NGF). In addition to the license amendment request, APS requested an exemption from certain requirements of 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors,” and 10 CFR part 50, Appendix K, “ECCS Evaluation Models,” to allow the use of Optimized ZIRLO^TM as a fuel rod cladding material.

The proposed change would allow for the implementation of NGF including the use of Optimized ZIRLO^TM fuel rod cladding material. The NGF assemblies contain advanced features to enhance fuel reliability, thermal performance, and fuel cycle economics.

Date of issuance: January 23, 2018.

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

Amendment Nos.: 205 (Unit 1), 205 (Unit 2), and 205 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17319A107; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: October 4, 2016 (81 FR 68469). The supplemental letters dated June 2 and December 15, 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 amendments is contained in a Safety Evaluation dated January 23, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: May 4, 2017.

Brief description of amendment: The amendment deletes a Surveillance Requirement Note associated with TS 3.5.1, “ECCS [Emergency Core Cooling System]—Operating,” TS 3.5.2, “ECCS—Shutdown,” and TS 3.6.1.7, “Residual Heat Removal (RHR) Containment Spray System,” to more appropriately reflect the RHR system design, and ensure the RHR system operation is consistent with the technical specification (TS) Limiting Condition for Operation (LCO) requirements. The amendment also adds a Note in the LCO for TS 3.5.1, TS 3.5.2, TS 3.6.1.7, TS 3.6.1.9, “Feedwater Leakage Control System,” and TS 3.6.2.3, “Residual Heat Removal (RHR) Suppression Pool Cooling,” to clarify that one of the required subsystems in each of the affected TS sections listed above may be inoperable during alignment and operation of the RHR system for Shutdown Cooling (i.e., decay heat removal) with the reactor steam dome pressure less than the RHR cut in permissive value.

Date of issuance: January 22, 2018.

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

Amendment No(s): 215. A publicly-available version is in ADAMS under Accession No. ML17324A354; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-62: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 5, 2017 (82 FR 31095).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 22, 2018.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, Pennsylvania

Date of amendment request: December 23, 2013, as supplemented by letters dated February 14, 2017; April 27, May 27, June 26, November 6, and December 21, 2015; February 24 and May 12, 2016; and January 30, April 21, June 23, August 22, October 25, and November 29, 2017.

Brief description of amendments: The amendments revised the Beaver Valley, Unit Nos. 1 and 2, Renewed Facility Operating Licenses (RFOLs) to establish and maintain a risk-informed, performance-based fire protection program in accordance with the requirements of 10 CFR 50.48(c).

Date of issuance: January 22, 2018.

Effective date: As of the date of issuance and shall be implemented consistent with paragraph 2.C.(5) for Unit No. 1, and paragraph 2.F for Unit No. 2, of the RFOLs.

Amendment Nos.: 301 (Unit No. 1) and 190 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17291A081; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.

RFOL Nos. DPR-66 and NPF-73: Amendments revised the RFOLs.

Date of initial notice in Federal Register: September 9, 2014 (79 FR 53458). The supplemental letters dated April 27, May 27, June 26, November 6, and December 21, 2015; February 24 and May 12, 2016; and January 30, April 21, June 23, August 22, October 25, and November 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 January 22, 2018.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: June 20, 2017.

Brief description of amendment: The amendment revised technical specifications (TSs) to delete the list of diesel generator critical trips from TS Surveillance Requirement (SR) 3.8.1.13 and clarify that the purpose of the SR is to verify that the non-critical automatic trips are bypassed.

Date of issuance: January 18, 2018.

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

Amendment No.: 179. A publicly-available version is in ADAMS under Accession No. ML17325B690; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-58: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 15, 2017 (82 FR 38718).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 18, 2018.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1 (FCS), Washington County, Nebraska

Date of amendment request: June 9, 2017, as supplemented by letter dated September 21, 2017.

Brief description of amendment: The amendment deleted Technical Specification (TS) 2.8.3(6), “Spent Fuel Cask Loading,” and associated Figure 2-11, “Limiting Burnup Criteria for Acceptable Storage in Spent Fuel Cask”; TS 3.2, Table 3-5, item 24, “Spent Fuel Cask Loading”; TS 4.3.1.3, Design Features associated with spent fuel casks; and portions of TS 3.2, Table 3-4, item 5, footnote (4) on boron concentration associated with cask loading.

Date of issuance: January 19, 2018.

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

Amendment No.: 296. A publicly-available version is in ADAMS under Accession No. ML17338A172; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-40: The amendment revised the renewed facility operating license and TSs.

Date of initial notice in Federal Register: August 15, 2017 (82 FR 38718).

The supplemental letter dated September 21, 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 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 January 19, 2018.

No significant hazards consideration comments received: No.

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: March 6, 2017, as supplemented by letters dated May 4, 2017, and September 14, 2017.

Brief description of amendments: The amendments revised Technical Specification 3.6.2.3, “Containment Cooling System,” to extend the containment fan coil unit allowed outage time from 7 days to 14 days for one or two inoperable containment fan coil units.

Date of issuance: January 18, 2018.

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

Amendment Nos.: 321 (Unit 1) and 302 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17349A108; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: June 6, 2017 (82 FR 26136). The supplemental letter dated September 14, 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 amendments is contained in a Safety Evaluation dated January 18, 2018.

No significant hazards consideration comment received: No.

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: April 7, 2017.

Brief description of amendments: The amendment revises the requirements of Technical Specification (TS) 3.6.4.1, “Secondary Containment,” associated with Surveillance Requirement (SR) 3.6.4.1.2. Specifically, SR 3.6.4.1.2 verifies that one secondary containment access door in each access opening is closed. The amendments would allow for brief, inadvertent, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions.

Date of issuance: January 22, 2018.

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

Amendment Nos.: Unit 1-289, Unit 2-234. A publicly-available version is in ADAMS under Accession No. ML17355A440; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: August 29, 2017 (82 FR 41070).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 22, 2018.

No significant hazards consideration comments received: No.

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: May 31, 2017, and supplemented by letter dated November 16, 2017.

Description of amendment: The amendment authorizes changes to the VEGP Units 3 and 4 Updated Final Safety Analysis Report in the form of departures from the plant-specific Design Control Document Tier 2 information and involves changes to the administrative controls for unborated water flow paths to the reactor coolant system to support chemical additions during periods when the reactor coolant pumps are not in operation. These proposed changes are reflected in Appendix A, Technical Specifications.

Date of issuance: January 9, 2018.

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

Amendment Nos.: 105 (Unit 3) and 104 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML17297A349; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined License.

Date of initial notice in Federal Register: September 12, 2017 (82 FR 42853). The supplemental letter dated November 16, 2017, provided additional information that clarified the application, did not expand the scope of the application request 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 the Safety Evaluation dated January 9, 2018.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 1, 2, and 3, San Diego County, California

Date of amendment request: December 15, 2016.

Brief description of amendments: The amendments replace the SONGS, Units 1, 2, and 3 Permanently Defueled Technical Specifications (TS) with Independent Spent Fuel Storage Installation (ISFSI) Only TS. These changes reflect the removal of all spent nuclear fuel from the SONGS, Units 2 and 3, spent fuel pools and its transfer to dry cask storage within the onsite ISFSI. The changes also make conforming revisions to the SONGS, Unit 1, TS and combine them with the SONGS, Units 2 and 3, TS. These changes will more fully reflect the permanently shutdown status of the decommissioning facility, as well as the reduced scope of structures, systems, and components necessary to ensure plant safety once all spent fuel has been permanently moved to the SONGS ISFSI, an activity which is currently scheduled for completion in 2019.

Date of issuance: January 9, 2017.

Effective date: As of the date Southern California Edison submits a written notification to the NRC that all spent nuclear fuel assemblies have been transferred out of the SONGS spent fuel pools and placed in storage within the onsite independent spent fuel storage installation, and shall be implemented within 60 days.

Amendment Nos.: Unit 1-169, Unit 2-237, and Unit 3-230: A publicly-available version is in ADAMS under Accession No. ML17345A657; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The amendments revise the Facility Operating Licenses.

Date of initial notice in Federal Register: February 14, 2017 (82 FR 10600).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 9, 2017.

No significant hazards consideration comments received: No.

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: January 25, 2017, as supplemented by letters dated March 21, 2017; August 4, 2017; and December 4, 2017.

Brief description of amendments: The amendments revised certain surveillance requirements in Technical Specification 3.8.1, “AC [Alternating Current] Sources—Operating.” The changes are in the use of steady-state voltage and frequency acceptance criteria for onsite standby power source of the diesel generators, allowing for the use of new and more conservative design analysis.

Date of issuance: January 22, 2018.

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

Amendment Nos.: 269 (Unit 1) and 251 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17352A711; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: June 6, 2017 (82 FR 26139). The supplemental letters dated August 4, 2017, and December 4, 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 amendments is contained in a Safety Evaluation dated January 22, 2018.

No significant hazards consideration comments received: No.

IV. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

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 for the amendment 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.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual notice of consideration of issuance of amendment, proposed no significant hazards consideration determination, and opportunity for a hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 51.12(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 application for amendment, (2) the amendment to Facility Operating License or Combined License, as applicable, 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.

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

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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 website 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.

B. Electronic Submissions (E-Filing)

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.

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

Date of amendment request: December 28, 2017.

Description of amendment: The amendment revised a note to Technical Specification Surveillance Requirement (SR) 4.1.3.1.2, such that Control Element Assembly (CEA) 4 may be excluded from the remaining quarterly performances of the SR in Cycle 26. The amendment allows the licensee to delay exercising CEA 4 until after repairs can be made during the next outage.

Date of issuance: January 18, 2018.

Effective date: As of the date of issuance and shall be implemented as soon as practicable and prior to the time in which SR 4.1.3.1.2 must be completed.

Amendment No.: 308. A publicly-available version is in ADAMS under Accession No. ML18011A064; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-6: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Public notice of the proposed amendment was published in the Arkansas Democrat-Gazette, located in Little Rock, Arkansas, from January 6 through January 7, 2018. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments were received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated January 18, 2018.

Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Robert J. Pascarelli.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant (WBN), Unit 2, Rhea County, Tennessee

Date of amendment request: January 10, 2018, as supplemented by letter dated January 17, 2018.

Description of amendment: The amendment revised Technical Specification (TS) 3.3.4, “Remote Shutdown Instrumentation,” to make a one-time change to TS Table 3.3.4-1, Function 4a, “RCS Hot Leg Temperature Indication,” to permit the temperature indicator for the Reactor Coolant System Loop 3 hot leg to be inoperable for the remainder of WBN Unit 2 Operating Cycle 2, the refueling outage for which is scheduled to start in spring 2019. The amendment also added a condition to the operating license to require implementation of compensatory measures described in the application that will remain in effect until the temperature indicator is returned to an operable condition.

Date of issuance: January 25, 2018.

Effective date: As of date of issuance.

Amendment No.: 19. A publicly-available version is in ADAMS under Accession No. ML18022B106; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-96: Amendment revised the technical specifications and operating license.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. The Rhea County Herald-News and The Advocate & Democrat on January 21, 2018, and The Daily Post-Athenian on January 22 and January 23, 2018. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. The supplemental letter dated January 17, 2018, 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 notice.

No comments have been received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated January 25, 2018.

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

NRC Branch Chief: Undine Shoop.

Dated at Rockville, Maryland, this 6th day of February 2018.

For the Nuclear Regulatory Commission.

Greg A. Casto,

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

[FR Doc. 2018-02636 Filed 2-12-18; 8:45 am]

BILLING CODE 7590-01-P