2019-12573. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from May 21, 2019 to June 3, 2019. The last biweekly notice was published on June 4, 2019.

DATES:

Comments must be filed by July 18, 2019. A request for a hearing must be filed by August 19, 2019.

ADDRESSES:

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

• Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0135. Address questions about NRC dockets IDs in Regulations.gov to Jennifer Borges; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.

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

FOR FURTHER INFORMATION CONTACT:

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2019-0135 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-2019-0135.
• 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 “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail] The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2019-0135 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. Background

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

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

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

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

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

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

Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's website at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest.

In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures.

Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [e-mail], or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public website at http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public website at http://www.nrc.gov/​site-help/​electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the “Contact Us” link located on the NRC's public website at http://www.nrc.gov/​site-help/​e-submittals.html,, by email to [e-mail], or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

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

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

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

Date of amendment request: March 27, 2019. A publicly available version is in ADAMS under Accession No. ML19086A315.

Description of amendment request: The proposed amendment would remove License Condition 2.C.(11), “Shield Wall Deferral (Section 12.3.2, SSER #4, License Amendment #7)” and its related Attachment 3, “List of Shield Walls” from Columbia's Renewed Facility Operating License, as these items are outdated and no longer applicable to Columbia's operation.

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

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

Response: No.

The proposed amendment involves the removal of an outdated license condition. The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment or any design changes to the facility. The proposed change does not affect any plant operations, design functions, or analyses that verify the capability of structures, systems, and components (SSCs) to perform a design function. The proposed change does not change any of the accidents previously evaluated in the Final Safety Analysis Report (FSAR). The proposed change does not affect SSCs, operating procedures, and administrative controls that have the function of preventing or mitigating any of these accidents.

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

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

Response: No.

The proposed amendment only involves the removal of an outdated license condition. No actual plant equipment or accident analyses will be affected by the proposed change. The proposed change will not change the design function or operation of any SSCs. The proposed change will not result in any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events.

Therefore, this proposed change does not create the possibility of an accident of a new or different kind than previously evaluated.

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

Response: No.

The proposed amendment only involves the removal of an outdated license condition. The proposed change does not involve any physical changes to the plant or alter the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

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

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

Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street NW, Washington, DC 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-286, Indian Point Nuclear Generating (Indian Point) Unit Nos. 1, 2, and 3, Westchester County, New York

Date of amendment request: April 15, 2019. A publicly available version is in ADAMS under Accession No. ML19105B278.

Description of amendment request: The amendments would revise the Indian Point Site Emergency Plan (SEP) for the permanently shutdown and defueled condition.

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Response: No.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the IPEC SEP and do not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised IPEC SEP will continue to provide the necessary response staff with the proposed changes.

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

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

Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: James G. Danna.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3 (Indian Point 2 and 3 or IP2 and IP3), Westchester County, New York

Date of amendment request: April 15, 2019. A publicly available version is in ADAMS under Accession No. ML19105B236.

Description of amendment request: The amendments propose changes to the staffing and training requirements for the Indian Point staff contained in Section 5.0, “Administrative Controls,” of the Indian Point 2 and Indian Point 3 Technical Specifications (TSs). Additional changes are also proposed to Section 1.1, “Definitions”; Section 4.0, “Design Features”; and Section 5.0, “Administrative Controls,” that are no longer applicable to a permanently defueled facility once Indian Point 2, and subsequently Indian Point 3, are permanently defueled.

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

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

Response: No.

The proposed amendment would not take effect until IP2 has permanently ceased operation and entered a permanently defueled condition and the Certified Fuel Handler Training and Retraining Program is approved by the NRC. The proposed amendment would modify the IP2 TS by deleting the portions of the TS that are no longer applicable to a permanently defueled facility, while modifying the other sections to correspond to the permanently defueled condition.

The deletion and modification of provisions of the administrative controls do not directly affect the design of structures, systems, and components (SSCs) necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the spent fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor. Thus, the consequences of an accident previously evaluated are not increased.

In a permanently defueled condition, the only credible accidents are the fuel handling accident (FHA) and those involving radioactive waste systems remaining in service. The probability of occurrence of previously evaluated accidents is not increased, because extended operation in a defueled condition will be the only operation allowed. This mode of operation is bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

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 have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The administrative removal or modifications of the TS that are related only to administration of the facility cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and IP2 will no longer be authorized to operate the reactor or retain or place fuel in the reactor vessel.

The proposed changes to the IP2 TS do not affect systems credited in the accident analysis for the FHA or radioactive waste system upsets at IP2. The proposed TS will continue to require proper control and monitoring of safety significant parameters and activities.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Extended operation in a defueled condition will be the only operation allowed, and it is bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

Since the 10 CFR part 50 license for IP2 will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel once the certifications required by 10 CFR 50.82(a)(1) are docketed, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible accidents are a FHA and those involving radioactive waste systems remaining in service. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact these analyzed conditions.

The proposed changes are limited to those portions of the TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the IP2 TS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated design basis accidents involving the reactor are no longer possible because the reactor will be permanently shutdown and defueled and IP2 will no longer be authorized to operate the reactor or retain or place fuel in the reactor vessel.

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

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

Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: James G. Danna.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2 (Indian Point 2 or IP2), Westchester County, New York

Date of amendment request: April 15, 2019. A publicly available version is in ADAMS under Accession No. ML19105B241.

Description of amendment request: The amendment would revise the Indian Point 2 Operating License (OL) and revise the Technical Specifications (TSs) in Appendix A to Permanently Defueled TSs, the Environmental TS Requirements in Appendix B of the OL, and the Inter-Unit Transfer TSs in Appendix C. The proposed changes would revise certain requirements contained within the Indian Point 2 OL and Appendices A through C TSs and remove the requirements that would no longer be applicable after Indian Point 2 is permanently shut down and defueled.

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

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

Response: No.

The proposed amendment would not take effect until IP2 has permanently ceased operation, entered a permanently defueled condition, met the decay requirements established in the analysis of the Fuel Handling Accident (FHA), implemented NRC approved License Amendments regarding fuel storage requirements and administrative controls for the permanently defueled condition, and received NRC approval of the Certified Fuel Handler Training and Retraining Program. The proposed amendment would modify the IP2 OL and TSs in Appendices A through C by deleting the portions of the OL and TSs that are no longer applicable to a permanently defueled facility, while modifying other portions to correspond to the permanently defueled condition. These proposed changes are consistent with the criteria set forth in 10 CFR 50.36 for the contents of TSs.

Section 14 of the IP2 Updated Final Safety Analysis Report (UFSAR) describes the DBA [design-basis accident] and transient scenarios applicable to IP2 during power operations. After the reactor is in a permanently defueled condition, the spent fuel pit (SFP) and its cooling systems will be dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents will be much smaller than for an operational plant. After the certifications are docketed for IP2 in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2), the majority of the accident scenarios previously postulated in the UFSAR will no longer be possible and will be removed from the UFSAR under the provisions of 10 CFR 50.59.

The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility structures, systems, and components (SSCs) or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled status of IP2 has no impact on the remaining applicable DBAs.

The removal of LCOs [limiting conditions for operation] or SRs [surveillance requirements] that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the permanently defueled condition. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system (RCS) inventory control, and containment integrity are no longer applicable at IP2 as a permanently shut down and defueled facility. The analyzed accidents involving damage to the RCS, main steam lines, reactor core, and the subsequent release of radioactive material will no longer be possible at IP2.

After IP2 permanently ceases operation, the future generation of fission products will cease and the remaining source term will decay. The radioactive decay of the irradiated fuel following shut down of the reactor will have reduced the consequences of the FHA below those previously analyzed.

The SFP water level, boron concentration, and fuel storage TSs are retained to preserve the current requirements for safe storage of irradiated fuel. SFP cooling and make-up related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since there will be sufficient time to effect repairs, establish alternate sources of make-up flow, or establish alternate sources of cooling in the event of a loss of cooling and make-up flow to the SFP.

The deletion and modification of provisions of the administrative controls of the Appendix A TSs and the non-radiological environmental protection requirements in Appendix B do not directly affect the design of SSCs necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the SFP. The changes do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shut down and defueled condition of the reactor.

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

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

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

Response: No.

The proposed changes to the IP2 OL and Appendices A through C TSs have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of TSs that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled and IP2 will no longer be authorized to operate the reactor.

The proposed deletion and modification of requirements of the IP2 OL and Appendices A through C TSs do not affect systems credited in the accidents that remain applicable at IP2 in the permanently defueled condition. The proposed OL and TSs will continue to require proper control and monitoring of safety significant parameters and activities.

The Appendix A TSs regarding SFP water level, boron concentration, and fuel storage are retained to preserve the current requirements for safe storage of irradiated fuel. The restriction on the SFP water level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

Response: No.

Because the 10 CFR part 50 license for IP2 will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2 as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible. The only remaining credible accidents are the FHA and the accidental release of waste liquids or waste gas. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the remaining DBAs.

The proposed amendment would modify the IP2 OL and TSs in Appendices A through C by deleting the portions of the OL and TSs that are no longer applicable to a permanently defueled facility, while modifying other portions to correspond to the permanently defueled condition. The requirements that are proposed to be deleted from the IP2 OL and Appendix A TSs are not credited in the existing accident analyses for the remaining DBAs; and as such, do not contribute to the margin of safety associated with the accident analyses. Postulated DBAs involving the reactors will no longer be possible because the reactor will be permanently shut down and defueled and IP2 will no longer be authorized to operate the reactor.

The Appendix A TSs regarding SFP water level, boron concentration, and fuel storage are retained to preserve the current requirements for safe storage of irradiated fuel.

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

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

Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: April 26, 2019. A publicly available version is in ADAMS under Accession No. ML19116A196.

Description of amendment request: The amendments would revise Peach Bottom Atomic Power Station, Units 2 and 3, Technical Specification (TS) 3.8.1, “AC [Alternating Current] Power—Operating,” Required Action A.3, to provide a temporary one-time extension of the completion time to allow sufficient time to perform physical modifications to replace 27 inaccessible electrical cables. These electrical cables are reaching the end of their dependable service life and are in need of replacement.

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 temporary one-time change to extend the Completion Time for TS 3.8.1, Required Action A.3, will not increase the probability of an accident, since the proposed Completion Time extension in the time duration that one qualified offsite circuit is out of service has no direct physical impact on the plant. The proposed inoperable offsite circuit limits the available redundancy of the offsite electrical system to a period not to exceed 21 days. Therefore, the proposed TS change does not have a direct impact on the plant that would make an accident more likely to occur due to extended Completion Time. Other sources of offsite and onsite power remain available.

During transients or events which require these systems/subsystems to be operating, there is sufficient capacity in the operable systems/subsystems to support plant operation or shutdown. Therefore, failures that are accident initiators will not occur more frequently than previously postulated as a result of the proposed temporary one-time TS change.

In addition, the consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR) will not be increased. With one offsite circuit inoperable, the consequences of any postulated accidents occurring on Unit 2 or Unit 3 during the proposed one-time Completion Time extension are bounded by the previous analyses as described in the UFSAR. The minimum equipment required to mitigate the consequences of an accident and/or safely shut down the plant will be operable or available during the extended Completion Time period of 21 days.

A risk evaluation has also been performed for the temporary one-time 21-day Completion Time extension. The evaluation concluded that the probability of a Loss of Offsite Power (LOOP) for the proposed configuration is very low. Therefore, the proposed change does not significantly increase the probability of an accident previously evaluated because: (a) The emergency buses continue to be fed from a reliable offsite source and; (b) the effect of the proposed configuration on the probability of a LOOP is very low.

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

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

Response: No.

The proposed temporary one-time change to extend the Completion Time for TS 3.8.1, Required Action A.3, will not create the possibility of a new or different type of accident since it will only extend the time period that one of the offsite circuits can be out of service; the extension of the time duration for one offsite circuit being inoperable has no direct physical impact on the plant and does not create any new accident initiators. Other sources of offsite and onsite power remain available. The systems involved are accident mitigation systems. The possible impacts that the inoperable equipment may have on supported systems was previously analyzed in the UFSAR. The impact of inoperable support systems was also previously assessed, and any accident initiators created by the inoperable systems were evaluated. Extending the duration of the Completion Time does not create any additional accident initiators for the plant.

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 existing TS Completion Time limit of seven (7) days for one offsite circuit inoperable was established to ensure that sufficient safety-related equipment is available for response to all accident conditions and that sufficient decay heat removal capability is available for a Loss of Coolant Accident (LOCA) coincident with a LOOP on one unit and simultaneous safe shutdown of the other unit. Although a very slight reduction in the margin of safety might be incurred during the proposed one-time extended Completion Time period, this slight reduction is judged to be minimal due to the low probability of an event occurring during the extended period. Other sources of offsite and onsite power remain available and operable during the 21-day extended period along with maintaining the availability of essential Emergency Core Cooling System (ECCS)/decay heat removal capability. The very slight reduction in the margin of safety resulting from extending the Completion Time from seven (7) days to 21 days when an offsite circuit is inoperable is not considered significant, since the remaining operable offsite circuit, the emergency Diesel Generators (DGs), the Station Blackout (SBO) line, and the FLEX DGs are available and provide an effective defense-in-depth plan to support the station electrical plant configurations during the extended 21-day Completion Time period.

The proposed TS change to extend the Completion Time does not affect the acceptance criteria for any analyzed event, nor is there a change to any safety limit. The proposed TS change does not affect any Structures, Systems or Components (SSC) or their capability to perform their intended functions. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are affected by this change. The proposed change will not result in plant operation in a configuration outside the current design basis. The margin of safety is maintained by maintaining the capability to supply emergency buses with a redundant, separate, reliable offsite power source, and maintaining the onsite power sources in their design basis configuration.

Operations personnel are fully qualified and trained to respond to, and mitigate, a Design Basis Accident (DBA), including actions needed to ensure decay heat removal systems are available while PBAPS [Peach Bottom Atomic Power Station], Units 2 and 3, are in the operational electrical configurations described within this submittal. Accordingly, existing procedures are in place that address safe plant shutdown and decay heat removal for situations applicable during the extended one-time Completion Time period.

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

NRC Branch Chief: James G. Danna.

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

Date of amendment request: June 26, 2018, as supplemented by letters dated February 25, 2019, and May 17, 2019 (ADAMS Accession Nos. ML18177A044, ML19056A387, and ML19137A070, respectively).

Description of amendment request: The license amendment request was originally noticed in the Federal Register on December 18, 2018 (83 FR 64894). This notice is being reissued in its entirety to include a revised description of the amendment request. The amendment would modify Technical Specification 3.3.1, “Oxygen Concentration,” to require inerting the primary containment to less than 4 percent by volume oxygen concentration within 72 hours of entering power operating condition. Also, the amendment would add a new requirement to identify required actions if the primary containment oxygen concentration increases to greater than or equal to four volume percent while in the power operating condition.

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

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

Response: No.

The proposed change modifies the Technical Specifications (TS) by adopting containment inerting and de-inerting requirements that are consistent with the guidance of NUREG-1433, “Standard Technical Specifications—General Electric BWR/4 Plants, Volume 1, Revision 4.0,” published April 2012. The proposed change will allow inerting of the primary containment within 24 hours of exceeding 15 percent (%) Rated Thermal Power (RTP), and de-inerting 24 hours prior to reducing reactor power to less than or equal to 15% RTP. Also, a new TS condition will be added to identify required actions if the primary containment oxygen concentration increases to greater than or equal to 4% by volume while in the power operating condition. The proposed change does not alter the physical configuration of the plant, nor does it affect any previously analyzed accident initiators. The accident analysis assumes that a Loss of Coolant Accident (LOCA) occurs at 100% RTP. The consequences of a LOCA at less than or equal to 15% RTP would be much less severe, and produce less hydrogen than a LOCA at 100% RTP.

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

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

Response: No.

The proposed change adopts the STS [Standard Technical Specifications] guidance regarding containment inerting/de-inerting requirements. The proposed change introduces no new mode of plant operation and does not involve any physical modification to the plant. The proposed change is consistent with the current safety analysis assumptions. No setpoints are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced.

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

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

Response: No.

The proposed change revises the Applicability presentation of the Oxygen Concentration TS. No safety limits are affected. The Oxygen Concentration TS requirements assure sufficient safety margins are maintained, and that the design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants' licensing basis. The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety.

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

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

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

NRC Branch Chief: James G. Danna.

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

Date of amendment request: April 26, 2019. A publicly available version is in ADAMS under Accession No. ML19119A249.

Description of amendment request: The amendment request proposes changes to the Combined License (COL) Numbers NPF-91 and NPF-92 for VEGP, Units 3 and 4, and Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 * and Tier 2 information related to the design-specific pre-operational Automatic Depressurization System (ADS) Blowdown Test. The requested amendment involves changes to credit the previously completed ADS Blowdown first three plant tests as described in the licensing basis documents, including COL Condition 2.D.(2)(a). Specifically, the proposed change would revise the COL, License Condition 2.D.(2)(a)2, by removing the requirement to perform the ADS Slowdown first three plant test during preoperational testing.

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

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

Response: No.

The proposed change does not affect the operation of any systems or equipment that initiates an analyzed accident or alter any structures, systems, or components (SSC) accident initiator or initiating sequence of events. The proposed changes remove the requirement to perform the ADS Blowdown first three plant test based on the successful completion of the tests at the lead AP1000 units. The change does not adversely affect any methodology which would increase the probability or consequences of a previously evaluated accident.

The change does not impact the support, design, or operation of mechanical or fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to predicted radioactive releases due to normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor does the proposed change create any new accident precursors.

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

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

Response: No.

The proposed change does not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created.

The proposed change credits previously completed ADS Blowdown first three plant testing based on the successful completion of the tests at the lead AP1000 units. The proposed changes do not adversely affect any design function of any SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or non-safety-related equipment. This activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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

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

Response: No.

The proposed change maintains existing safety margin and provides adequate protection through continued application of the existing requirement in the UFSAR. The proposed change satisfies the same design functions in accordance with the same codes and standards as stated in the UFSAR. This change does not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change.

Since no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by this change, no significant margin of safety is reduced.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: May 1, 2019. A publicly available version is in ADAMS under Accession No. ML19126A309.

Description of amendment request: The amendments would revise the South Texas Project, Units 1 and 2, Technical Specifications in Section 3.0 and Section 4.0 regarding limiting condition for operation (LCO) and surveillance requirement (SR) usage. The proposed changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-529, “Clarify Use and Application Rules,” using the consolidated line item improvement process (ADAMS Accession No. ML16062A271). The model safety evaluation was approved by the NRC in a letter dated April 21, 2016 (ADAMS Accession No. ML16060A441).

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 to Technical Specification LCO 3.0.4 has no effect on the requirement for systems to be Operable and has no effect on the application of Technical Specification actions. The proposed change to Technical Specification SR 4.0.3 states that the allowance may only be used when there is a reasonable expectation the surveillance will be met when performed. Since the proposed change does not significantly affect system Operability, the proposed change will have no significant effect on the initiating events for accidents previously evaluated and will have no significant effect on the ability of the systems to mitigate accidents previously evaluated.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to the Technical Specifications usage rules does not affect the design or function of any plant systems. The proposed change does not change the Operability requirements for plant systems or the actions taken when plant systems are not Operable.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change clarifies the application of Technical Specification LCO 3.0.4 and does not result in changes in plant operation. Technical Specification SR 4.0.3 is revised to allow application of Technical Specification SR 4.0.3 when a Surveillance Requirement has not been previously performed if there is reasonable expectation that the Surveillance Requirement will be met when performed. This expands the use of Technical Specification SR 4.0.3 while ensuring the affected system is capable of performing its safety function. As a result, plant safety is either improved or unaffected.

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 request for amendments involves no significant hazards consideration.

Attorney for licensee: Kym Harshaw, Vice President and General Counsel, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, TX 77483.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: November 26, 2018, as supplemented by letter dated May 13, 2019. Publicly-available versions are in ADAMS under Accession Nos. ML18331A134 and ML19134A233, respectively.

Description of amendment request: The amendments would revise technical specifications (TSs) to support performance of 6.9 kiloVolt and associated 480 Volt shutdown board (SDBD) maintenance. The proposed changes provide operational flexibility for two-unit operation by providing sufficient time to perform preventive maintenance on SDBDs associated with a defueled unit while the opposite unit is operating in Modes 1, 2, 3, or 4.

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

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

Response: No.

The proposed change modifies the required actions for the opposite unit's onsite and offsite AC power sources and electrical distribution system. The opposite unit's AC power sources and electrical distribution system are required to be operable to support the associated unit's required features. In addition, a change is proposed to remove the details regarding the required input power to the vital inverters. This change will not affect the probability of an accident, since the AC power sources, vital inverters, and electrical distribution system are not initiators of any accident sequence analyzed in the WBN dual-unit Updated Final Safety Analysis Report (UFSAR). Rather, the AC power sources, vital inverters, and electrical distribution system support equipment used to mitigate accidents. The consequences of an analyzed accident will not be significantly increased since the minimum requirements for AC power sources, vital inverters, and electrical distribution system will be maintained to ensure the availability of the required power to mitigate accidents assumed in the UFSAR. Operation in accordance with the proposed TS will ensure that sufficient AC power sources, vital inverters, and electrical distribution subsystems are operable, as required to support the unit's required features. Therefore, the mitigating functions supported by the AC power sources, vital inverters, and electrical distribution system will continue to provide the protection assumed by the accident analysis. The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by the proposed changes. Thus, the consequences of previously analyzed accidents will not increase by implementing these changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

The proposed changes involve restructuring the TS for the AC electrical power system to provide more flexibility in performing maintenance on electrical system components. The AC electrical power system is not an initiator to any accident sequence analyzed in the UFSAR. Rather, the AC electrical power system supports equipment used to mitigate accidents. The proposed changes to modify the required actions associated with inoperable opposite unit AC power sources and shutdown boards and proposed changes to the details of the required power supplies to the vital inverters will maintain the same level of equipment performance required for mitigating accidents assumed in the UFSAR. Therefore, operation of the facility in accordance with this proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The equipment margins will be maintained in accordance with the plant-specific design bases as a result of the proposed changes. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient AC capability to support operation of mitigation equipment is ensured. The equipment fed by the AC electrical sources will continue to provide adequate power to safety-related loads in accordance with analysis assumptions. The proposed TS changes maintain the same level of equipment performance stated in the UFSAR and the current TSs. Therefore, the proposed changes do not involve a significant reduction of safety.

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

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

NRC Branch Chief: Undine Shoop.

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

Date of amendment request: November 29, 2018. A publicly-available version is in ADAMS under Accession No. ML18334A389.

Description of amendment request: The amendments would modify the Watts Bar Nuclear Plant, Units 1 and 2, Technical Specification requirements related to direct current (DC) electrical systems to be consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360” (ADAMS Accession No. ML092670242).

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 any accident previously evaluated?

Response: No.

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

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

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

Response: No.

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

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The equipment margins will be maintained in accordance with the plant-specific design bases because of the proposed changes. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery Maintenance and Monitoring Program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety-related loads in accordance with analysis assumptions. TS changes made in accordance with TSTF-500, Revision 2, maintain the same level of equipment performance stated in the FSAR and the current TSs. Therefore, the proposed changes do not involve a significant reduction [in the margin] of safety.

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

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

NRC Branch Chief: Undine Shoop.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendment request: March 18, 2019. A publicly available version is in ADAMS under Accession No. ML19086A113.

Description of amendment request: The amendments would revise Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station, Units 1 and 2, respectively, by revising the Technical Specification (TS) requirements regarding the Emergency Diesel Generators. Specifically, TS 3.8.1, “AC Sources—Operating,” would be revised to reduce the maximum voltage specified in the associated surveillance requirements.

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

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

Response: No.

Modifying the maximum steady-state voltage requirement does not increase the probability of an accident. Verifying proper operation of the EDGs to maintain adequate voltage ensures proper electrical and mechanical system function and does not increase the consequences of an accident.

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

Response: No.

The proposed change would provide more restrictive acceptance criteria to be applied to existing technical specification surveillance tests that demonstrate the capability of the facility EDGs to perform their design function. The proposed acceptance criteria changes would not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. Therefore, the possibility of a new or different kind of accident from any previously evaluated has not been created.

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

Response: No.

The proposed change involves decreasing maximum voltage test acceptance criterion for EDG Surveillance Tests. The conduct of surveillance tests on safety-related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed amendment does not affect EDG performance as described in the design basis analyses, including the capability of the EDG to maintain required voltage for proper operation of plant safety loads. The proposed amendment does not introduce changes to limits established in the accident analyses. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. W.S. Blair, Senior Counsel, Dominion Energy Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

NRC Branch Chief: Michael T. Markley.

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

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

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

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

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

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

Date of application for amendments: July 31, 2015, as supplemented by letters dated April 11, 2016; November 3, 2017; and May 18, June 1, September 21, and October 5, 2018.

Brief description of amendments: The amendments revised certain technical specification (TS) requirements related to Completion Times for Required Actions to provide the option to calculate a longer, risk-informed completion time. The allowance is described in a new program, “Risk Informed Completion Time Program,” that was added to TS Section 5.0, “Administrative Controls.” The methodology for using the Risk-Informed Completion Time Program is described in Nuclear Energy Institute (NEI) Report NEI 06-09, “Risk-Informed Technical Specifications Initiative 4b: Risk-Managed Technical Specifications (RMTS) Guidelines,” Revision 0-A.

Date of issuance: May 29, 2019.

Effective date: As of the date of issuance and shall be implemented within 270 days from the date of issuance.

Amendment Nos.: Unit 1—209; Unit 2—209; Unit 3—209. A publicly-available version is in ADAMS under Accession No. ML19085A525. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register : December 8, 2015 (80 FR 76317). By letter dated November 3, 2017, the licensee supplemented its application. By supplemental letters dated May 18 and June 1, 2018, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC published a second proposed no significant hazards consideration determination in the Federal Register on August 14, 2018 (83 FR 40345), which superseded the original notice in its entirely. The supplemental letters dated September 21, and October 5, 2018, provided additional information that clarified the application, did not expand the scope of the application as noticed, and did not change the staff's second proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 29, 2019.

Dairyland Power Cooperative, Docket No.: 50-409, La Crosse Boiling Water Reactor, La Crosse County, Wisconsin

Date of application for amendment: June 27, 2016, supplemented by letter dated December 1, 2016, May 31, 2018, and November 15, 2018.

Brief description of amendment: The amendment revises the La Crosse Boiling Water Reactor (LACBWR) license to approve the License Termination Plan (LTP). The LACBWR LTP provides the details of the plan for characterizing, identifying, and remediating the remaining residual radioactivity at the LACBWR site to a level that will allow the site to be released for unrestricted use. The LACBWR LTP also describes how the licensee will confirm the extent and success of remediation through radiological surveys, provide financial assurance to complete decommissioning, and ensure the environmental impacts of the decommissioning activities are within the scope originally envisioned in the associated environmental documents. Decommissioning activities at the LACBWR site are scheduled to be complete in 2019, with license termination occurring before the end of 2020.

Date of issuance: May 21, 2019.

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

Amendment No.: 75.

Possession Only License No. DPR-45: The amendment revised the Possession Only License.

Date of initial notice in Federal Register : August 30, 2016 (81 FR 59663). The supplements dated December 1, 2016, May 31, 2018, and November 15, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not affect the applicability of the NRC's generic no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a safety evaluation dated May 21, 2019, which is available in the Agencywide Documents Access and Management System (ADAMS) at Accession No. ML19008A079).

No significant hazards consideration comments received: Not applicable.

Dominion Energy Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: May 3, 2018, as supplemented by letters dated November 29, 2018; March 27, 2019; and May 7, 2019.

Brief description of amendment: The amendment revised the Technical Specifications to reflect the results and constraints of a new criticality safety analysis for fuel assembly storage in the Millstone Power Station, Unit No. 3, fuel storage racks. Specifically, the amendment implemented the following items associated with fuel assembly storage: (1) Increased the Technical Specification minimum spent fuel pool soluble boron concentration, (2) revised allowed storage patterns and initial enrichment/burnup/decay time for fuel assemblies in the spent fuel pool to meet k_eff requirements under normal and accident conditions, (3) permitted the storage of any fuel assembly with certain enrichment that contains a rod cluster control assembly in Region 2 without restriction, and (4) implemented a revised criticality analysis for the new fuel storage racks using the updated methods for the spent fuel pool criticality analysis for consistency.

Date of issuance: May 28, 2019.

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

Amendment No.: 273. A publicly available version is in ADAMS under Accession No. ML19126A000; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-49: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : August 7, 2018 (83 FR 38735). The supplemental letters dated November 29, 2018; March 27, 2019; and May 7, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC 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 28, 2019.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1 (ANO-1), Pope County, Arkansas

Date of amendment request: March 12, 2018, as supplemented by letters dated April 26, October 17, and December 11, 2018.

Brief description of amendment: The amendment revised the ANO-1 Technical Specifications and operating license by relocating certain surveillance frequencies to a licensee-controlled program, consistent with the NRC-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk-Informed TSTF] Initiative 5b.”

Date of issuance: May 22, 2019.

Effective date: As of the date of issuance.

Amendment No.: 264. A publicly available version is in ADAMS under Accession No. ML19098A955; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-51: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : June 5, 2018 (83 FR 26102). The supplemental letters dated October 17 and December 11, 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 May 22, 2019.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., Cooperative Energy, A Mississippi Electric Cooperative, and Entergy Mississippi, LLC, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: April 10, 2018, as supplemented by letters dated October 23, 2018, and March 13, 2019.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control.” The change replaced existing TS requirements related to “operations with a potential for draining the reactor vessel” with new requirements on reactor pressure vessel water inventory control to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

Date of issuance: May 23, 2019.

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

Amendment No: 218. A publicly available version is in ADAMS under Accession No. ML19084A218; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-29: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : June 5, 2018 (83 FR 26103). The supplemental letters dated October 23, 2018, and March 13, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC 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 23, 2019.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-171, 50-277, and 50-278, Peach Bottom Atomic Power Station, Units 1, 2, and 3, York and Lancaster Counties, Pennsylvania

Date of amendment request: May 10, 2018, as supplemented by letters dated November 1 and November 29, 2018.

Brief description of amendments: The amendments revised the emergency response organization positions identified in the emergency plan for each site.

Date of issuance: May 24, 2019.

Effective date: As of the date of issuance and shall be implemented on or before December 31, 2019.

Amendment Nos.: Limerick—235/198 and Peach Bottom—14/325/328. A publicly-available version is in ADAMS under Accession No. ML19078A018. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-39, NPF-85, DPR-12, DPR-44, and DPR-56: Amendments revised the emergency plans.

Date of initial notice in Federal Register : July 17, 2018 (83 FR 33268). The supplemental letters dated November 1 and November 29, 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 24, 2019.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 29, 2018, as supplemented by letter dated October 27, 2018.

Brief description of amendments: The amendments revised Technical Specification (TS) 3/4.3.1, “Reactor Trip System Instrumentation”; TS 3/4.3.2, “Engineered Safety Feature Actuation System Instrumentation”; TS 3/4.7.1.5, “Main Steam Isolation Valves”; and added a new TS for feedwater isolation to better align the TSs with the design-basis analyses and the design of the instrumentation.

Date of issuance: May 31, 2019.

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

Amendment Nos.: 329 (Unit No. 1) and 310 (Unit No. 2). A publicly available version is in ADAMS under Accession No. ML19105B171; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register : August 28, 2018 (83 FR 43907). The supplemental letter dated October 27, 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 31, 2019.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 20, 2017, as supplemented by letters dated February 15, April 9, and October 4, 2018.

Brief description of amendments: The amendments revised Technical Specification (TS) 4.2.1, “Fuel Assemblies,” for Unit 2 to allow up to 1,792 tritium producing burnable absorber rods in the reactor; and revised the Units 1 and 2 TSs related to fuel storage.

Date of issuance: May 22, 2019.

Effective date: As of the date of issuance and shall be implemented prior to startup from the outage where any number of tritium producing burnable absorber rods is inserted in the Watts Bar Nuclear Plant, Unit 2, reactor core not to exceed December 31, 2022.

Amendment Nos.: 125 (Unit 1) and 27 (Unit 2). A publicly available version is in ADAMS under Accession No. ML18347B330; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-90 and NPF-96: The amendments revised the Facility Operating Licenses and TSs.

Date of initial notice in Federal Register : June 8, 2018 (83 FR 26709). The supplement dated October 4, 2018, provided additional information that clarified the application, and did not expand the scope of the application as originally noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 22, 2019.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Unit 1 (Wolf Creek), Coffey County, Kansas

Date of amendment request: January 17, 2017, as supplemented by letters dated March 22, May 4, July 13, October 18, and November 14, 2017; January 15, January 29, April 19, June 19, August 9, November 15 (two letters), and December 6, 2018; and March 5, May 2, and May 15, 2019.

Brief description of amendment: The amendment revised the Wolf Creek Technical Specifications to replace the existing methodology for performing core design, non-loss-of-coolant-accident and loss-of-coolant accident safety analyses with standard Westinghouse Electric Corporation developed and NRC-approved analysis methodologies. In addition, the amendment revised the Wolf Creek licensing basis by adopting the alternative source term (AST) radiological analysis methodology in accordance with 10 CFR 50.67, “Accident source term.” This amendment represented a full scope implementation of the AST as described in Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.”

Date of issuance: May 31, 2019.

Effective date: As of the date of issuance and shall be implemented during startup (prior to entry into Mode 2) from Refueling Outage 23.

Amendment No.: 221. A publicly available version is in ADAMS under Accession No. ML19100A122; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-42. The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : On July 5, 2017, the NRC staff published a proposed no significant hazards consideration (NSHC) determination in the Federal Register (82 FR 31084) for the proposed amendment. Subsequently by letters dated July 13, October 18, and November 14, 2017; January 15, January 29, April 19, June 19, and August 9, 2018, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC published a second proposed NSHC determination in the Federal Register on October 2, 2018 (83 FR 49590), which superseded the original notice in its entirety. The supplemental letters dated November 15 (two letters) and December 6, 2018; and March 5, May 2, and May 15, 2019, provided additional information that clarified the application, did not expand the scope of the application as noticed on October 2, 2018, and did not change the NRC staff's proposed NSHC determination published in the Federal Register dated October 2, 2018.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 31, 2019.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 10th day of June 2019.

For the Nuclear Regulatory Commission.

Craig G. Erlanger,

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