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

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Start Preamble

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 April 24, 2018, to May 7, 2018. The last biweekly notice was published on May 8, 2018.

DATES:

Comments must be filed by June 21, 2018. A request for a hearing must be filed by July 23, 2018.

ADDRESSES:

You may submit comments by any of the following methods:

  • Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0096. Address questions about NRC dockets to Jennifer Borges; 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-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, Start Printed Page 23729see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

Start Further Info

FOR FURTHER INFORMATION CONTACT:

Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC 2018-0096, 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-2018-0096.
  • 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-2018-0096, 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 Start Printed Page 23730of 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 Start Printed Page 23731on 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, Unit Nos. 1 and 2, Brunswick County, North Carolina

Date of amendment request: January 10, 2018. A publicly-available version is in ADAMS under Accession No. ML18010A344.

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

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Response: No.

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

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

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

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

NRC Acting Branch Chief: Brian W. Tindell.

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

Date of amendment request: February 1, 2018. A publicly-available version is in ADAMS under Accession No. ML18033B768.Start Printed Page 23732

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

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

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

Response: No.

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

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

Response: No.

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

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

Response: No.

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

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

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

Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.

NRC Acting Branch Chief: Brian W. Tindell.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

Date of amendment request: January 29, 2018. A publicly-available version is in ADAMS under Accession No. ML18029A187.

Description of amendment request: The proposed change would modify the RBS Updated Safety Analysis Report (USAR) and Technical Requirements Manual to relocate the reactor core isolation cooling (RCIC) piping injection point from the reactor vessel head spray nozzle to the feedwater line using the residual heat removal (RHR) shutdown cooling return line.

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.

Basis: The relocation of the RCIC injection point from the reactor vessel head spray nozzle to the `A' Feedwater line via the `A' RHR shutdown cooling return line does not adversely affect the design function of an System, Structure, or Component (SSC) or a method of performing or controlling a design function of an SSC as described in the USAR so there is no change to the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the USAR. There is no impact to the likelihood of occurrence of a malfunction of a structure, system, or component because there are no structures systems or components changed or affected by the scope of this evaluation.

Inadvertent initiation of RCIC may be categorized as either a Decrease in Reactor Coolant Temperature event or an Increase in Reactor Coolant Inventory event. River Bend Transient Safety Analysis Design Report, 6224.302-000-035A, states that three systems were considered that could introduce a cold water perturbation (Decrease in Reactor Coolant Temperature Event) at operating pressures: RCIC, High Pressure Core Spray (HPCS), and the feedwater system. This report qualifies improper startup of HPCS or RCIC as events that would produce no significant power transients. The proposed change relocated the injection point of the RCIC flow from the reactor head (RPV [reactor pressure vessel]) to the feedwater line (FWS). This change will reduce the effects of steam quenching. However, the effect of steam quenching is not credited in any of the safety analysis. The only portion of the RCIC system operation that is credited is water injection at the required flow rate, and the design function as described in the USAR of the RCIC system is to maintain or supplement the reactor vessel water inventory. The source of water for the Inadvertent RCIC injection remains the same. The destination of the water for the Inadvertent RCIC injection is still the RPV. The ability of the rerouted equipment to satisfy the RCIC design function is not reduced from the original design requirement to inject 600 gpm [gallons per minute] into the RPV. This is maintained by the RCIC flow controller. The entry location from the RPV head spray to the feedwater line has no impact to the consequences of an inadvertent initiation of RCIC. As the consequences of an inadvertent initiation of RCIC are unchanged, the consequences of this event remain quantitatively bounded by the Loss of Feedwater Heating event described in section 15.1.1 of the USAR for the Decrease in Reactor Coolant Temperature category and bounded by the Inadvertent HPCS Startup for the Increase in Reactor Coolant Inventory category.

Changing the injection point of RCIC does not increase the probability or consequences of an inadvertent RCIC injection. All affected piping, fittings, and valve pressure boundaries are qualified to the appropriate fluid transients and operational conditions in accordance with the design and licensing basis. No instrument setpoints were changed as a result of this modification. The RCIC system's modes of operation are not changed or affected by this modification. Therefore there is no change in the frequency of an inadvertent initiation of RCIC event. There is no change in the frequency of inadvertent initiation of RCIC by this modification, so there is no impact to the probability of any previously evaluated accident.Start Printed Page 23733

Therefore, it is concluded that this change does not significantly increase 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.

Basis: The spurious start of RCIC accident is evaluated in the USAR as Event 9 “Inadvertent HPCS Pump Start (Moderator Temperature Decrease) as shown in USAR Appendix 15A. The Inadvertent HPCS Pump Start event bounds the inadvertent operation of RCIC event and is quantitatively analyzed in accordance with Reg Guide 1.70 rev. 3. This event may be classified as either a Decrease in Core Coolant Temperature event or an Increase in Reactor Coolant Inventory Event, however was categorized as an Increase in Reactor Coolant Inventory Event in the RBS USAR as this is the initial effect of this event. No new accident is created by the scope of this modification because all aspects of the existing Decrease in Core Coolant Temperature and Increase in Reactor Coolant Inventory events and their relationship to the spurious start of RCIC remain applicable.

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.

Basis: The proposed change does not change any accident analyses. The proposed change does not exceed or alter a design basis or safety limit; therefore it does not significantly reduce the margin of safety.

Therefore, 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: Anna Vinson Jones, Senior Counsel—Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington DC 20001.

NRC Branch Chief: Robert J. Pascarelli.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: February 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18067A115.

Description of amendment request: The amendment would modify the River Bend Station Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk-Informed TSTF] Initiative 5b.” Additionally, the change would add a new program, the Surveillance Frequency Control Program (SFCP), to TS Chapter 5.0, “Administrative Controls.”

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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.

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

Therefore, the 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 design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Entergy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

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

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

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

NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas Nuclear One (ANO), Unit Nos. 1 and 2, Pope County, Arkansas

Date of amendment request: March 29, 2018. A publicly-available version is in ADAMS under Accession No. ML18088B412.

Description of amendment request: The amendments would revise the ANO, Units 1 and 2, currently approved Emergency Plan Emergency Action Level (EAL) scheme, which is based on the Nuclear Energy Institute (NEI) guidance established in NEI 99-01, Revision 5, “Methodology for Development of Emergency Action Levels,” by adopting the EAL schemes based on the guidance provided in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors.”

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

1. Do the proposed changes involve a significant increase in the probability or Start Printed Page 23734consequences of an accident previously evaluated?

Response: No.

The proposed changes to the ANO EALs do not involve any physical changes to plant equipment or systems and do not alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors and do not alter design assumptions, plant configuration, or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems or components (SSCs) to perform intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.

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

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

Response: No.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do 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. Because EALs are not accident initiators and no physical changes are made to the plant, no new causal mechanisms are introduced.

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

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

Response: No.

Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not impact operation of the plant and no accident analyses are affected by the proposed changes. The changes do not affect the Technical Specifications or the method of operating the plant. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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

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

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

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: February 1, 2018. A publicly-available version is in ADAMS under Accession No. ML18036A227.

Description of amendment request: The proposed amendments would revise the Braidwood Station licensing basis for protection from tornado-generated missiles by identifying the TORMIS Computer Code as the methodology used for assessing tornado-generated missile protection of unprotected plant structures, systems, and components (SSCs).

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

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

Response: No.

The NRC TORMIS Safety Evaluation Report [ADAMS Accession No. ML080870291] states the following:

“The current Licensing criteria governing tornado missile protection are contained in [NUREG-0800] Standard Review Plan (SRP) Section 3.5.1.4, [Missiles Generated by Natural Phenomena] and 3.5.2 [Structures, Systems and Components [SSCs]] to be Protected from Externally Generated Missiles]. These criteria generally specify that safety-related systems be provided positive tornado missile protection (barriers) from the maximum credible tornado threat. However, SRP Section 3.5.1.4 includes acceptance criteria permitting relaxation of the above deterministic guidance, if it can be demonstrated that the probability of damage to unprotected essential safety-related features is sufficiently small.”

As permitted by these SRP sections, the combined probability will be maintained below an allowable level, i.e., an acceptance criterion threshold, which reflects an extremely low probability of occurrence. SRP Section 2.2.3, “Evaluation of Potential Accidents,” established this threshold as approximately 1.0E-06 per year if, “when combined with reasonable qualitative arguments, the realistic probability can be shown to be lower.” The Braidwood Station analysis approach assumes that if the sum of the individual probabilities calculated for tornado missiles striking and damaging portions of safety-significant SSCs is greater than or equal to 1.0E-06 per year per unit, then installation of tornado missile protection barriers would be required for certain components to lower the total cumulative damage probability below the acceptance criterion of 1.0E-06 per year per unit. Conversely, if the total cumulative damage probability remains below the acceptance criterion of 1.0E-06 per year per unit, no additional tornado missile protection barriers would be required for any of the unprotected safety-significant components.

With respect to the probability of occurrence or the consequences of an accident previously evaluated in the UFSAR [Updated Final Safety Analysis Report], the possibility of a tornado impacting the Braidwood Station site and causing damage to plant SSCs is a licensing basis event currently addressed in the UFSAR. The change being proposed (i.e., the use of the TORMIS methodology for assessing tornado-generated missile protection of unprotected plant SSCs), does not affect the probability of a tornado strike on the site; however, from a licensing basis perspective, the proposed change does affect the probability that missiles generated by a tornado will strike and damage certain safety-significant plant SSCs. There are a defined number of safety-significant components that could theoretically be struck and damaged by tornado-generated missiles. The probability of tornado-generated missile hits on these “important” systems and components is calculated using the TORMIS probabilistic methodology. The combined probability of damage for unprotected safety-significant equipment will be maintained below the acceptance criterion of 1.0E-06 per year per unit to ensure adequate equipment remains available to safely shutdown the reactors, and maintain overall plant safety, should a tornado strike occur. Consequently, the proposed change does not constitute a significant increase in the probability of occurrence or the consequences of an accident based on the extremely low probability of damage caused by tornado-generated missiles and the commensurate extremely low probability of a radiological release.

Finally, the use of the TORMIS methodology will have no impact on accident initiators or precursors; does not alter the accident analysis assumptions or the manner in which the plant is operated or maintained; and does not affect the probability of operator error.

Based on the above discussion, 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 impact of a tornado strike on the Braidwood Station site is a licensing basis event that is explicitly addressed in the Start Printed Page 23735UFSAR. The proposed change simply involves recognition of the acceptability of using an analysis tool (i.e., the TORMIS methodology) to perform probabilistic tornado missile damage calculations in accordance with approved regulatory guidance. The proposed change does not result in the creation of any new accident precursors; does not result in changes to any existing accident scenarios; and does not introduce any operational changes or mechanisms that would create the possibility of a new or different kind of accident.

Therefore, the proposed change will not create the possibility of a new or different kind of accident than those previously evaluated.

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

Response: No.

The existing Braidwood Station licensing basis regarding tornado missile protection of safety-significant SSCs assumes that missile protection barriers are provided for safety-significant SSCs; or the unprotected component is assumed to be unavailable post-tornado.

The results of the Braidwood Station TORMIS analysis have demonstrated that there is an extremely low probability, below an established regulatory acceptance limit, that these “important” SSCs could be struck and subsequently damaged by tornado-generated missiles. The change in licensing basis from protecting safety-significant SSCs from tornado missiles, to demonstrating that there is an extremely low probability that safety-significant SSCs will be struck and damaged by tornado-generated missiles, does not constitute a significant decrease in the margin of safety.

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

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

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

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

NRC Branch Chief: David J. Wrona.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County, Minnesota

Date of amendment request: March 15, 2018. A publicly-available version is in ADAMS under Accession No. ML18074A308.

Brief description of amendment request: The proposed amendments would revise Prairie Island Nuclear Generating Plant Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with implementation of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,” Revision 1. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—Risk Informed Technical Specifications Task Force (RITSTF) Initiative 5b,” Revision 3.

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new SFCP [surveillance frequency control program]. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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.

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

Therefore, the 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 design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, NSPM will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

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

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

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: March 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18087A323.

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

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

1. Does the proposed amendment involve a significant increase in the probability or Start Printed Page 23736consequences of an accident previously evaluated?

Response: No.

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

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: April 3, 2018. A publicly-available version is in ADAMS under Accession No. ML18094A189.

Description of amendment request: The proposed amendment would change Functional Units 17.A and 17.B of Technical Specification (TS) Table 4.3-1, “Reactor Trip System Instrumentation Surveillance Requirements.” The Trip Actuating Device Operational Test (TADOT) column of this table would be revised to delete the “S/U” frequency and replace it with a reference to Table Notation (8), which would state, “Prior to entering MODE 1 whenever the unit has been in MODE 3.” The licensee stated that the change would align the surveillance requirements and the mode requirement for the Turbine Trip TADOT with the TS 3/4.3.1, Table 3.3-1, “Reactor Trip System Instrumentation,” channels and interlocks mode requirement.

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

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

Response: No.

The proposed changes revise the surveillance frequency for reactor trip functions from a turbine trip event. These changes do not alter these functions physically, or how they are maintained. Changing the surveillance from “prior to Startup” to “prior to entering MODE 1” will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation and prior to applicability. Since these changes will not affect the ability of these trips to perform the initiation of reactor trips when appropriate, the offsite dose consequences for an accident will not be impacted. Equally, the potential to cause an accident is not affected because no plant system or component has been altered by the proposed changes.

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

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

Response: No.

The proposed changes only affect surveillance frequency requirements for the turbine trip functions. This does not affect any physical features of the plant, or the manner in which these functions are utilized. The proposed surveillance frequency will require the functions to be verified operable before the turbine trip functions are applicable and able to perform their trip functions. Changing the surveillance from “prior to Startup” to “prior to entering MODE 1” will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes do not alter any plant setpoints or functions that are assumed to actuate in the event of postulated accidents. The proposed changes do not alter any plant feature and only alters the MODE which the surveillance tests must be performed. The proposed changes ensure the functionality of the turbine trips when assumed in the analysis for accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: April 6, 2018. A publicly-available version is in ADAMS under Accession No. ML18096B463.

Description of amendment request: The requested amendments require changes to the Updated Final Safety Analysis Report (UFSAR) in the form of Start Printed Page 23737departures from the incorporated plant-specific Design Control Document (DCD) Tier 2 information and related changes to the Vogtle Electric Generating Plant, Unit Nos. 3 and 4, combined license (COL) and COL Appendix C (and corresponding plant-specific DCD Tier 1) information. Specifically, the requested amendments include changes to the equipment survivability assessment requirements associated with hydrogen burns during beyond design-basis accidents as described in the licensing basis documents, including COL Condition 2.D(12)(g)9 and plant-specific Tier 1 Sections 2.2.3 and 2.3.9.

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

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

Response: No.

The proposed changes and clarifications to the locations of Hydrogen Igniters 27, 30, 35, 36, 37, and 38 do not adversely affect any safety-related structure, system or component (SSC) or function. The hydrogen ignition subsystem is designed to mitigate beyond design basis hydrogen generation in the containment. With the proposed changes, the hydrogen ignition subsystem continues to maintain the designed and analyzed beyond design basis functions. The hydrogen ignition subsystem maintains its design function to maintain containment integrity. The proposed changes also reconcile the as-built equipment with the list of equipment on which the equipment survivability assessment is performed to provide additional assurance containment penetrations and combustible gas control components will perform their design functions after a hydrogen burn in containment. The changes are to the equipment assessed, not to the design functions of the equipment. The changes do not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the plant-specific UFSAR are not affected. The proposed changes do not involve a change to any mitigation sequence or the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected.

The maximum allowable containment vessel leakage rate specified in the Technical Specifications is unchanged, and radiological material release source terms are not affected; thus, the radiological releases in the accident analyses are not affected. The proposed changes do not affect the prevention and mitigation of other abnormal events (e.g. anticipated operational occurrences, earthquakes, floods and turbine missiles), or their safety or design analyses. Therefore, 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 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 reconcile the as-built equipment with the list of equipment on which the equipment survivability assessment is performed to provide additional assurance that containment penetrations and combustible gas control components will perform their design functions after a hydrogen burn in containment. The equipment survivability assessment changes are to the equipment assessed, not to the design functions of the equipment. The VLS Hydrogen Ignition subsystem does not interface with/affect safety-related equipment or a fission product barrier. The subsystem is provided to address the production of hydrogen following a beyond design basis accident in accordance with 10 CFR 50.44(c). The hydrogen ignition subsystem is a non-Class 1E subsystem and does not interface with any safety-related system; thus, no system or design function or equipment qualification is affected by the proposed changes. The changes to the hydrogen ignition subsystem do not result in a new failure mode, malfunction or sequence of events that could affect a radioactive material barrier or safety-related equipment. The proposed changes do not adversely affect any system or design function or equipment qualification as the changes do not modify any SSCs that prevent safety functions from being performed. The changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety-related equipment.

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

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

Response: No.

The proposed changes and clarifications to the locations of Hydrogen Igniters 27, 30, 35, 36, 37, and 38 maintain the beyond design basis function of the hydrogen ignition subsystem. The proposed changes also reconcile the as-built equipment with the list of equipment on which the equipment survivability assessment is performed to provide additional assurance containment penetrations and combustible gas control components will perform their design functions after a hydrogen burn in containment. The equipment survivability assessment changes are to the equipment assessed, not to the design functions of the equipment. The proposed changes would not affect any safety-related design code, function, design analysis, safety analysis input or result, or existing design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: April 13, 2018. A publicly-available version is in ADAMS under Accession No. ML18103A249.

Description of amendment request: The requested amendments require changes to combined license (COL) Appendix A, Technical Specifications and the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information. Specifically, the requested amendments include changes to the COL Appendix A, Technical Specifications related to the statuses of the remotely operated containment isolation valves.

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

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

Response: No.

This change clarifies that only Class 1E valves in the nonessential containment penetration flow paths that receive the containment isolation signal (T signal) are part of the [Post-Accident Monitoring (PAM)] Technical Specifications and adds additional valves to the PAM table in the UFSAR. The Normal Residual Heat Removal System (RNS), Chemical and Volume Control System (CVS), Component Cooling Water System (CCS), and Steam Generator System (SGS) have containment isolation valves that do not Start Printed Page 23738close on a T signal because they have an accident mitigation function to be open.

The status of the valves in the essential containment flow paths are summarized on one non-safety display screen and are separately indicated on the safety display screens within their respective systems. Keeping these indications separate from the “Remotely Operated Containment Isolation Valve Status” which is on the Category 1 display allows the operators to quickly verify that the nonessential containment flow paths are isolated and then focus on the availability of the essential flow paths for their defense-in-depth capabilities.

The valve position indications in the essential flow paths that penetrate containment are not Post-Accident Monitoring System (PAMS) B1 variables. These essential flow paths support accident mitigation functions of non-safety systems and may be intentionally opened for extended periods of time following an accident. As a result, excluding them from the PAMS B1 summary indication will increase the value of the summary indication during operation of the essential flow paths.

Furthermore, opening these essential flow paths pose low risk of becoming an unmonitored leak path through the containment vessel. The valves are isolated when required by separate Protection and Safety Monitoring System (PMS) signals that are associated with each system's post-accident functions, and the valve position indications are designated as PAMS D2 accordingly.

No structure, system, or component (SSC) or function is changed within this activity. Therefore, the proposed amendment does not involve a significant increase in the probability of an accident previously evaluated.

The proposed amendment does not affect the prevention and mitigation of abnormal events, e.g., accidents, anticipated operation occurrences, earthquakes, floods, turbine missiles, and fires or their safety or design analyses. This change does not involve containment of radioactive isotopes or any adverse effect on a fission product barrier. There is no impact on previously evaluated accidents.

Therefore, the proposed changes do not involve a significant increase in the 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 involve a new failure mechanism or malfunction, which affects an SSC accident initiator, or interface with any SSC accident initiator or initiating sequence of events considered in the design and licensing bases. There is no adverse effect on radioisotope barriers or the release of radioactive materials. The proposed amendment does not adversely affect any accident, including the possibility of creating a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

This activity clarifies that only Class 1E valves in the nonessential containment penetration flow paths that receive the containment isolation signal (T signal) are part of the PAM Technical Specifications and adds additional valves to the PAM table in the UFSAR.

The status of the valves in the essential containment flow paths are summarized on one non-safety display screen and are separately indicated on the safety display screens within their respective systems. Keeping these indications separate from the “Remotely Operated Containment Isolation Valve Status” which is on the Category 1 display allows the operators to quickly verify that the nonessential containment flow paths are isolated and then focus on the availability of the essential flow paths for their defense-in-depth capabilities.

The valve position indications in the essential flow paths that penetrate containment are not PAMS B1 variables. These essential flow paths support accident mitigation functions of non-safety systems and may be intentionally opened for extended periods of time following an accident. As a result, excluding them from the PAMS B1 summary indication will increase the value of the summary indication during operation of the essential flow paths.

Furthermore, opening these essential flow paths pose low risk of becoming an unmonitored leak path through the containment vessel. The valves are isolated when required by separate PMS signals that are associated with each system's post-accident functions and the valve position indications are designated as PAMS D2 accordingly.

No SSC or function is changed within this activity. The proposed changes would not affect any safety-related design code, function, design analysis, safety analysis input or result, or existing design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes.

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

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

Attorney for licensee: 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, Unit Nos. 3 and 4, Burke County, Georgia

Date of amendment request: April 20, 2018. A publicly-available version is in ADAMS under Accession No. ML18110A113.

Description of amendment request: The requested amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the plant-specific Design Control Document (DCD) Tier 2 information and involves changes to the plant-specific Tier 1 information (and associated Combined License (COL) Appendix C information). Specifically, the amendment proposes changes to plant-specific Tier 1 (and COL Appendix C) Table 2.5.2-3, “PMS Automatically Actuated Engineered Safety Features,” to revise the nomenclature for “Auxiliary Spray and Letdown Purification Line Isolation” and to include “Component Cooling System Containment Isolation Valve Closure.” Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific DCD Tier 1 material departures.

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

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

Response: No.

The proposed nomenclature changes reflect the current plant design. These changes provide consistency with the approved plant design. The changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components accident initiator or initiating sequence of events. The proposed changes do not result in any increase in probability of an analyzed accident occurring. 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 nomenclature changes reflect the current plant design. These changes provide consistency with the approved plant design. The proposed changes do not affect plant electrical systems, Start Printed Page 23739and do not affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is affected by the proposed changes. 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 nomenclature changes reflect the current plant design. These changes provide consistency with the approved plant design. No safety analysis or design basis acceptance limit/criterion is involved. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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.

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

Date of amendment request: August 30, 2016, as supplemented by letter dated November 20, 2017.

Brief description of amendment: The amendment revised the Columbia Generating Station Final Safety Analysis Report to reclassify reactor water cleanup piping, valves, pumps, and mechanical modules located outside of the primary and secondary containment in the radwaste building from Quality Group C to Quality Group D.

Date of issuance: April 17, 2018.

Effective date: As of its date of issuance and shall be implemented from the date of issuance until restart after Refueling Outage 24 (spring 2019).

Amendment No.: 248. A publicly-available version is in ADAMS under Accession No. ML18075A351; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. This Notice of Issuance is being reissued in its entirety to reflect a correction to the “Effective date” by letter dated April 27, 2018 (ADAMS Accession No. ML18109A215).

Renewed Facility Operating License No. NPF-21: The amendment revised the Final Safety Analysis Report.

Date of initial notice in Federal Register: December 6, 2016 (81 FR 87968). The supplemental letter dated November 20, 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 April 17, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 25, 2017.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.5.1, “ECCS [Emergency Core Cooling Systems]—Operating,” and deleted the Note associated with Surveillance Requirement 3.5.1.2 to reflect the residual heat removal system design and ensure the residual heat removal system's operation is consistent with the TS 3.5.1 limiting condition for operation requirements.

Date of issuance: May 2, 2018.

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

Amendment No.: 249. A publicly-available version is in ADAMS under Accession No. ML18100A199; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: November 7, 2017 (82 FR 51649).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 2, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 28, 2017, as supplemented by letter dated February 28, 2018.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.8.1.3, “Diesel Fuel Oil,” by relocating the current required stored diesel fuel oil numerical volumes from the TSs to the TS Bases and replacing them with comparable duration-based requirements. In addition, the amendment revised TS 3.8.1.1 and TS 3.8.1.2, “AC [Alternating Current] Sources Operating,” and “AC Sources Shutdown,” respectively, to relocate the specific numerical value for feed tank fuel oil volume to the TS Bases and replace it with the feed tank operating time requirement. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-501, Revision 1, “Relocate Fuel Oil and Lube Oil Volume Values to Licensee Control.”Start Printed Page 23740

Date of issuance: April 26, 2018.

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

Amendment No.: 251. A publicly-available version is in ADAMS under Accession No. ML18026B053; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-38: The amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: July 5, 2017 (82 FR 31093). The supplemental letter dated February 28, 2018, 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 April 26, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 30, 2017.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.7.11, “Control Room Ventilation (VC) Temperature Control System,” to modify the TS Actions for two inoperable VC temperature control system trains.

Date of issuance: April 30, 2018.

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

Amendment Nos.: 195/195; 201/201. A publicly-available version is in ADAMS under Accession No. ML18054B436; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revised the TSs and Licenses.

Date of initial notice in Federal Register: August 29, 2017 (82 FR 41068).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 30, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 18, 2017.

Brief description of amendment: The amendment revised the design value for the spent fuel storage pool in Technical Specification 4.3.2, “Drainage,” to an appropriate value, consistent with the original design basis.

Date of issuance: April 30, 2018.

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

Amendment No: 217. A publicly-available version is in ADAMS under Accession No. ML18072A050; 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: September 12, 2017 (82 FR 42848).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 30, 2018.

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County, Minnesota

Date of amendment request: August 4, 2017, as supplemented by letter dated November 6, 2017.

Brief description of amendments: The amendments revised the non-destructive examination inspection interval for special lifting devices from annually or prior to each use, typically at each refueling outage, to a 10-year interval.

Date of issuance: May 1, 2018.

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

Amendment Nos.: 225 (Unit 1) and 212 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18100A788; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-42 and DPR-60: The amendments revised the Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Updated Safety Analysis Report.

Date of initial notice in Federal Register: September 26, 2017 (82 FR 44855). The supplemental letter dated November 6, 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 May 1, 2018.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California

Date of amendment request: September 28, 2017.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.1.4, “Rod Group Alignment Limits”; TS 3.1.5, “Shutdown Bank Insertion Limits”; TS 3.1.6, “Control Bank Insertion Limits”; and TS 3.1.7, “Rod Position Indication,” to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1, “Clarification of Rod Position Requirements.” The NRC approved the TSTF and issued an associated model safety evaluation by letter dated March 4, 2016.

Date of issuance: April 30, 2018.

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

Amendment Nos.: 232 (Unit 1) and 234 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18096A054; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: November 7, 2017 (82 FR 51653).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 30, 2018.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 7, 2017, as supplemented by letters dated November 1, November 27, December 14, December 19 (four letters), and Start Printed Page 23741December 22, 2017, and January 22, 2018.

Brief description of amendment: The amendment revised the Renewed Facility Operating License and Technical Specifications to implement a measurement uncertainty recapture power uprate. Specifically, the amendment authorized an increase in the maximum licensed thermal power level from 3,840 megawatts thermal to 3,902 megawatts thermal, which is an increase of approximately 1.6 percent.

Date of issuance: April 24, 2018.

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

Amendment No.: 212. A publicly-available version is in ADAMS under Accession No. ML18096A542; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: October 3, 2017 (82 FR 46098). The supplemental letters dated November 1, November 27, December 14, December 19 (four letters), and December 22, 2017, and January 22, 2018, 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 April 24, 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), Unit Nos. 1, 2, and 3, San Diego County, California

Date of amendment request: December 19, 2016, as supplemented by letters dated April 25, 2017, and November 2, 2017.

Brief description of amendments: The amendments replaced the SONGS, Unit Nos. 1, 2, and 3, Physical Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan (the Security Plan) with an Independent Spent Fuel Storage Installation (ISFSI) Only Security Plan. The NRC staff determined that the proposed SONGS ISFSI-Only Security Plan continues to meet the standards in 10 CFR 72.212, “Conditions of general license issued under § 72.210,” paragraph (b)(9). As such, the SONGS ISFSI-Only Security Plan provides reasonable assurance that adequate protective measures can and will be taken in the event of a design-basis threat of radiological sabotage related to the spent fuel. These changes more fully reflect the status of the facility, as well as the reduced scope of potential physical security challenges at the site once all spent fuel has been moved to dry cask storage within the onsite ISFSI, an activity that is currently scheduled for completion in 2019.

Date of issuance: April 23, 2018.

Effective date: As of its date of issuance and shall be implemented within 60 days following Southern California Edison Company's submittal of a written certification to the NRC that all spent nuclear fuel assemblies have been transferred out of the spent fuel pools and placed in storage within the onsite ISFSI.

Amendment Nos.: 170 (Unit 1), 238 (Unit 2), and 231 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17311A364; the Safety Evaluation enclosed with the amendments includes safeguards information and is withheld from public disclosure.

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

Date of initial notice in Federal Register: April 4, 2017 (82 FR 16422).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 5, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 18, 2017, as supplemented by letter dated December 15, 2017.

Description of amendments: The amendments authorized the Southern Nuclear Operating Company to depart from the Vogtle Electric Generating Plaint Updated Final Safety Analysis Report (UFSAR) Tier 2* and Tier 2 information regarding changes necessary to reflect an increase in the design pressure of the main steam isolation valve (MSIV) compartments from 6.0 pounds per square inch (psi) to 6.5 psi and other changes regarding descriptions of the MSIV compartments. The Tier 2* changes affect Wall 11 information contained in UFSAR Subsections 3H.3.3, 3H.5.1, and 3H.5.1.3. This change provides additional design margin for the MSIV Compartments A and B at the Vogtle Electric Generating Plant.

Date of issuance: April 18, 2018.

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

Amendment Nos.: 122 (Unit 3) and 121 (Unit 4). Publicly-available versions are in ADAMS Package Accession No. ML18085A932, which includes the Safety Evaluation that references documents related to these amendments.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendments revised the Facility Combined Licenses.

Date of initial notice in Federal Register: November 21, 2017 (82 FR 55411). The supplemental letter dated 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.

The Commission's related evaluation of the amendments is contained in the Safety Evaluation dated April 18, 2018.

No significant hazards consideration comments received: No.

United States Maritime Administration (MARAD), Docket No. 50-238, Nuclear Ship SAVANNAH (NSS), Baltimore, Maryland

Date of amendment request: October 31, 2017.

Brief description of amendment: The amendment permits MARAD to begin dismantling and disposing of the NSS without prior approval of the NRC, consistent with existing decommissioning regulations.

Date of issuance: April 23, 2018.

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

Amendment No.: 15. A publicly-available version is in ADAMS under Accession No. ML18081A134.

Facility Operating License No. NS-1: This amendment revised the License.

Date of initial notice in Federal Register: February 13, 2018 (83 FR 6235).

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

No significant hazards consideration comments received: No.

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

Date of amendment request: May 23, 2017, as supplemented by letters dated January 16, 2018, and March 14, 2018.Start Printed Page 23742

Brief description of amendments: The amendments revised plant Technical Specifications Table 3.7-2 and associated Table Notations, Table 3.7-4 and Table 4.1-1, reflecting the installation of the Class 1E 4160V negative sequence voltage (open phase) protective circuitry at Surry Power Station, Unit Nos. 1 and 2, to address the potential for a consequential open phase condition that could exist on one or two phases of a primary offsite power source and that would not currently be detected and mitigated by the existing station electrical protection scheme.

Date of issuance: May 3, 2018.

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

Amendment Nos.: 292 (Unit No. 1) and 292 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML18106A007; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: October 10, 2017 (82 FR 47040). The supplemental letters dated January 16, 2018, and March 14, 2018, 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 May 3, 2018.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 14th day of May, 2018.

For the Nuclear Regulatory Commission.

Tara Inverso,

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

End Signature End Supplemental Information

[FR Doc. 2018-10565 Filed 5-21-18; 8:45 am]

BILLING CODE 7590-01-P