2017-14743. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from June 20, 2017 to July 3, 2017. The last biweekly notice was published on July 5, 2017.

DATES:

Comments must be filed by August 17, 2017. A request for a hearing must be filed by September 18, 2017.

ADDRESSES:

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

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0158. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0158, facility name, unit number(s), plant docket number, application date, and subject, when contacting the NRC about the availability of information for this action. You may obtain publicly available information related to this action by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0158.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2017-0158, 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 posts all comment submissions at http://www.regulations.gov as well as entering 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 submissions into ADAMS.

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

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

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

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

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

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

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

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

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

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

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

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

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

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

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

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

Date of amendment request: June 14, 2017. A publicly-available version is in ADAMS under Accession No. ML17165A555.

Description of amendment request: The amendments would modify the completion date for implementation of Milestone 8 of the Cyber Security Plan (CSP). The proposed amendments would extend the CSP Milestone 8 completion date from September 30, 2017, to December 31, 2017.

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 to the PVNGS Cyber Security Plan implementation schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components (SSCs) relied upon to mitigate the consequences of postulated accidents, and has no impact on the probability or consequences of an accident previously evaluated.

Therefore, the proposed change will 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 to the PVNGS Cyber Security Plan implementation schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the SSCs relied upon to mitigate the consequences of postulated accidents, and does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

Plant safety margins are established through limiting conditions for operation, limiting safety systems settings, and safety limits specified in the [T]echnical [S]pecifications [TSs]. The proposed change to the PVNGS Cyber Security Plan implementation schedule is administrative in nature. Since the proposed change is administrative in nature, there are no changes to these established safety margins.

Therefore, the proposed change does not involve a significant reduction in a margin of safety as defined in the basis for any TS.

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: Michael G. Green, Senior Regulatory Counsel, Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, Phoenix, AZ 85072-2034.

NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: March 30, 2017, as supplemented by letter dated May 11, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17095A530 and ML17139D352, respectively.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) in accordance with the NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-448, Revision 3, “Control Room Habitability,” with variations from the TSTF to account for plant-specific configuration and licensing basis differences. The amendments would modify the TSs for the control room ventilation system (CRVS) booster fans and would establish a control room envelop (CRE) habitability program in TS 5.5, “Programs and Manuals.” The NRC staff issued “Notice of Availability of Technical Specification Improvement to Modify Requirements Regarding Control Room Envelope Habitability Using the Consolidated Line Item Improvement Process,” associated with TSTF-448, Revision 3, in the Federal Register on January 17, 2007 (72 FR 2022). The notice included a model safety evaluation, a model no significant hazards consideration determination, and a model license amendment request. In its application dated March 30, 2017, as supplemented by letter dated May 11, 2017, the licensee affirmed the applicability of the model no significant hazards consideration determination, which is presented in the following section.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee affirmed the applicability of the model no significant hazards consideration, which is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the [CRVS], which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the [CRVS] is the CRE boundary. The [CRVS] is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the [CRVS] is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the [CRVS] will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased.

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

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the [CRVS], or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice.

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

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down 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: Lara S. Nichols, Vice President Nuclear & EHS Legal Support, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202-1802.

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: May 15, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17139D261.

Description of amendment request: The proposed amendment would replace the Permanently Defueled Emergency Plan and its associated Permanently Defueled Emergency Action Level (EAL) Technical Bases Document with the Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and its associated ISFSI EAL Technical Bases Document, for the Vermont Yankee Nuclear Power Station (VY). The proposed changes would reflect the complete removal of all fuel from the spent fuel pool (SFP) and permit specific reductions in the size and makeup of the Emergency Response Organization due to the elimination of the design-basis accident related to the spent fuel (fuel handling accident). As described in the Post Shutdown Decommissioning Activities Report, spent fuel will remain in the SFP until it meets the criteria for transfer, the existing ISFSI is expanded, and the spent fuel can be safely transferred in an efficient manner to the expanded ISFSI, an activity that is currently scheduled for completion in late 2018.

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 modify the VY facility operating license by revising the emergency plan and EAL scheme. VY has permanently ceased power operations and is permanently defueled. The proposed amendment is conditioned on all spent nuclear fuel being removed from wet storage in the SFP and placed in dry storage within the ISFSI. Occurrence of postulated accidents associated with spent fuel stored in a SFP is no longer credible in a SFP devoid of fuel. The proposed amendment has no effect on plant structures, systems, or components (SSC) and therefore can neither affect the capability of any plant SSC to perform its design function nor increase the likelihood of the malfunction of any plant SSC. The proposed amendment would have no effect on any of the previously evaluated accidents in the VY Defueled Safety Analysis Report or the Holtec HI-STORM 100 Final Safety Analysis Report.

Because VY has permanently ceased power operations, the generation of fission products has largely ceased and the remaining source term continues to decay. This source term decay continues to significantly reduce the consequences of previously evaluated postulated accidents. Furthermore, previously generated source term materials such as reactor water cleanup resins have been removed from the site in accordance with applicable regulations and permitting requirements.

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 amendment constitutes a revision of the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at VY.

The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment as a result of the proposed amendment. Similarly, the proposed amendment would not physically change any SSC involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures. The credible events for the ISFSI remain unchanged.

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

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

Response: No.

Because the 10 CFR part 50 license for VY no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the postulated accidents associated with reactor operation are no longer credible. In addition, with all spent nuclear fuel transferred out of wet storage from the SFP and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible during dry storage of spent nuclear fuel. Therefore, there are no credible events that would result in radiological releases beyond the site boundary exceeding the exposure levels in U.S. EPA's “Protective Action Guide and Planning Guidance for Radiological Incidents,” dated January 2017.

The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not affect either the way in which the plant SSCs perform their safety function or their design margins. Because there is no change to the physical design of the facility, there is no change to these margins.

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: Ms. Susan Raimo, Senior Counsel, Entergy Services, Inc., 101 Constitution Ave. NW., Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Bruce Watson.

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

Date of amendment request: April 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17129A612.

Description of amendment request: The amendments would modify the completion date for implementation of Milestone 8 of the Cyber Security Plan (CSP). The proposed amendments would extend the CSP Milestone 8 full implementation date from December 31, 2017, to December 31, 2022.

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 the CSP Implementation Schedule does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and has no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed change to the CSP Implementation Schedule does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the CSP Implementation Schedule does not involve these items. In addition, the milestone date delay for full implementation of the CSP has no substantive impact because other measures have been taken which provide adequate protection during this period of time. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety.

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: 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, Westchester County, New York

Date of amendment request: April 7, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17104A039.

Description of amendment request: The amendment would revise Technical Specification 3.5.4, “Refueling Water Storage Tank (RWST),” such that the non-seismically qualified piping of the Boric Acid Recovery System be connected to the RWST seismic piping. This change will only be applicable until the end of the Indian Point Nuclear Generating Unit No. 2 Refueling Outage 2R23.

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 use of the non-seismic Boric Acid Recovery System (BARS) to recirculate and filter the RWST water does not involve any changes or create any new interfaces with the reactor coolant system or main steam system piping. Therefore, the connection of the BARS Purification Loop to the RWST would not affect the probability of these accidents occurring. The BARS is not credited for safe shutdown of the plant or accident mitigation. Administrative controls ensure that the BARS can be isolated as necessary and in sufficient time to assure that the RWST volume will be adequate to perform the safety function as designed. Since the RWST will continue to perform its safety function and overall system performance is not affected, the consequences of the accident are not increased.

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

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

Response: No.

The design of the RWST and the SFP [Spent Fuel Pool] Purification Loop has been revised to allow recirculation and purification using the BARS for a short period of time (not to exceed 30 days per fuel cycle) for the next fuel cycle. The BARS takes RWST water in and processes it out without additional connections that could affect other systems and without an impact from its installation. Procedures for the operation of the plant, including the BARS, will not create the possibility of a new or different type of accident. Contingent upon manual operator action, a BARS line break will not result in a loss of the RWST safety function. Similarly, an active or passive failure in the BARS will not affect safety related structures, systems or components.

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 SFP Purification Loop and recirculation and purification of the RWST water using the BARS is not credited for safe shutdown of the plant or accident mitigation. RWST volume will be maximized prior to purification and timely operator action can be taken to isolate the non-seismic system from the RWST to assure it can perform its function. This will result in no significant reduction in the margin of safety.

Therefore the proposed change does not significantly reduce 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.

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

Date of amendment request: April 9, 2017. A publicly-available version is in ADAMS under Accession No. ML17100A269.

Description of amendment request: The amendments would revise Technical Specification (TS) Section 4.2.1, “Fuel Assemblies,” and Section 5.6.3, “Core Operating Limits Report (COLR),” to allow the use of Optimized ZIRLO^TM as an approved fuel rod cladding material. In the letter dated April 9, 2017, the licensee also requested an exemption from certain requirements of 10 CFR 50.46 and 10 CFR part 50, appendix K, in accordance with 10 CFR 50.12, to support the license amendments.

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 allow the use of Optimized ZIRLO^TM clad nuclear fuel at BVPS. The NRC approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO^TM,” prepared by Westinghouse Electric Company LLC (Westinghouse), which addresses Optimized ZIRLO^TM fuel rod cladding and demonstrates that Optimized ZIRLO^TM fuel rod cladding has essentially the same properties as currently licensed ZIRLO® fuel rod cladding. The use of Optimized ZIRLO^TM fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The fuel cladding itself is not an accident initiator and does not affect accident probability. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes and do not affect the fuel cladding design. Use of Optimized ZIRLO^TM meets the fuel design acceptance criteria and hence does not significantly affect the consequences of an accident.

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 use of Optimized ZIRLO^TM fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes and do not affect the fuel cladding design. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A demonstrated that the material properties of Optimized ZIRLO^TM fuel rod cladding are similar to those of ZIRLO® fuel rod cladding. Therefore, Optimized ZIRLO^TM fuel rod cladding will perform similarly to ZIRLO® fuel rod cladding, thus precluding the possibility of the fuel rod cladding becoming an accident initiator and causing 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 accident previously evaluated.

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

Response: No.

The proposed amendment will not involve a significant reduction in the margin of safety. NRC-approved Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of the Optimized ZIRLO^TM fuel rod cladding are similar to those of ZIRLO® fuel rod cladding. Optimized ZIRLO^TM fuel rod cladding is expected to perform similarly to ZIRLO® fuel rod cladding for normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The use of Optimized ZIRLO^TM fuel rod cladding will not result in adverse changes to the operation or configuration of the facility. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes that do not affect the fuel cladding design.

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: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: James G. Danna.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: June 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17164A191.

Description of amendment request: The requested amendments propose changes to the Updated Final Safety Analysis Report in the form of departures from the plant-specific Design Control Document (DCD) Tier 2 information, and involve changes to related plant-specific DCD Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. In addition, revisions are proposed to COL Appendix A, Technical Specifications. The proposed changes revise the COLs concerning standardizing the Protection and Safety Monitoring System (PMS) setpoint nomenclature. No changes are proposed to setpoint values or PMS alarms and actuations. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, Design Certification Rule, is also requested for the plant-specific Tier 1 departures.

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

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

Response: No.

No setpoint values or PMS actuations are proposed to be changed by this activity. Nor are any values assumed in the safety analysis changed. This is an administrative change to standardize the PMS setpoint designators. The proposed amendment does not affect the prevention and mitigation of abnormal events, e.g., accidents, anticipated operation occurrences, earthquakes, floods, turbine missiles, and fires or their safety or design analyses. This change does not involve containment of radioactive isotopes or any adverse effect on a fission product barrier. There is no impact on previously evaluated accidents.

These proposed changes have no adverse impact on the support, design, or operation of mechanical and fluid systems. The response of systems to postulated accident conditions is not adversely affected and remains within response time assumed in the accident analysis. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. Consequently, 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 requested amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

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

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

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

Response: No.

No setpoint values or PMS actuations are proposed to be changed by this activity. This is an administrative change to standardize the PMS setpoint designators. The proposed changes would not affect any safety-related design code, function, design analysis, safety analysis input or result, or existing design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes.

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

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

Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.

NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, Georgia

Date of amendment request: June 9, 2017. A publicly-available version is in ADAMS under Accession No. ML17163A174.

Description of amendment request: The requested amendments propose changes to combined license (COL) Appendix C (and plant-specific Tier 1) Table 2.7.2-2 to revise the minimum chilled water flow rates to the supply air handling units serving the Main Control Room and the Class 1E electrical rooms, and the unit coolers serving the normal residual heat removal system and chemical and volume control system pump rooms. The proposed COL Appendix C (and plant-specific Design Control Document (DCD) Tier 1) changes require additional changes to corresponding Tier 2 component data information in Updated Final Safety Analysis Report Chapters 6 and 9. Because this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 DCD, the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).

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 COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, Updated Final Safety Analysis Report (UFSAR) Table 9.2.7-1, and associated UFSAR design information to identify the revised equipment parameters for the nuclear island nonradioactive ventilation system (VBS) air handling units (AHUs) and radiologically controlled area (RCA) ventilation system (VAS) unit coolers and reduced chilled water system (VWS) cooling coil flow rates does not adversely impact the plant response to any accidents which are previously evaluated. The function of the cooling coils to provide chilled water to the VBS AHUs and VAS unit coolers is not credited in the safety analysis.

No safety-related structure, system, component (SSC) or function is adversely affected by this change. The change does not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the plant-specific UFSAR are not affected. The proposed changes do not involve a change to the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected. The proposed changes do not increase the probability or consequences of an accident previously evaluated as the VWS, VBS and VAS do not provide safety-related functions and the functions of each system to support required room environments are not changed.

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

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

Response: No.

The proposed changes to COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design information to identify the revised equipment parameters for VBS AHUs and VAS unit coolers and reduced VWS cooling coil flow rates do not affect any safety-related equipment, and do not add any new interfaces to safety-related SSCs. The VWS function to provide chilled water is not adversely impacted. The function of the VAS to provide ventilation and cooling to maintain the environment of the serviced areas within the design temperature range is not adversely impacted by this change. No system or design function or equipment qualification is affected by these changes as the change does not modify the operation of any SSCs. The changes do not introduce a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment. Revised equipment parameters, including the reduced cooling coil flow rates, do not adversely impact the function of associated components.

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

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

Response: No.

The changes to COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design information do not affect any other safety-related equipment or fission product barriers. The requested changes will not adversely affect compliance with any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes as previously evaluated accidents are not impacted.

Therefore, the proposed amendment does not involve a significant reduction in 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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.

NRC Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: May 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17144A408.

Description of amendment request: The amendments would revise Surveillance Requirement 3.3.1.3 to change the thermal power at which the surveillance may be performed.

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 to the TS [Technical Specification] does not affect the initiators of any analyzed accident. In addition, operation in accordance with the proposed amendment to the TS ensures that the previously evaluated accidents will continue to be mitigated as analyzed. The proposed amendment does not adversely affect the design function or operation of any structures, systems, and components important to safety.

The probability or consequences of accidents previously evaluated in the UFSAR [Updated Final Safety Analysis Report] are unaffected by this proposed amendment because there is no change to any equipment response or accident mitigation scenario. There are no new or additional challenges to fission product barrier integrity.

Therefore, it is concluded that 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?

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

No new accident scenarios, failure mechanisms, or limiting single failures are introduced due to the proposed amendment. The proposed amendment does not challenge the performance or integrity of any safety-related system.

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

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

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed amendment will have no affect on the availability, operability, or performance of the safety-related systems and components. No change is being made to the requirement to perform the surveillance. The NOTE in the surveillance is being changed to clarify when the initial surveillance after refueling is to be performed. The Technical Specification Limiting Condition for Operation (LCO) limits are not being changed.

The proposed amendment will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis.

Therefore, it is concluded that 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: May 5, 2017. A publicly-available version is in ADAMS under Accession No. ML17128A120.

Description of amendment request: The requested amendments propose changes to more clearly define the boundaries and seismic requirements for the portion of the fire protection system (FPS) piping that is required to remain functional following a safe shutdown earthquake (SSE) (i.e., the “seismic standpipe system”). The proposed changes also include the removal of SSE requirements from pipe lines that do not need to remain functional following an SSE (specifically, the FPS piping that is part of the non-seismic FPS containment spray system and the FPS open tray system).

Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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

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

Response: No.

The proposed modification changes would clarify the boundaries for the portion of the nonsafety-related FPS required to remain functional following a SSE for manual firefighting in areas with SSE equipment, and the addition of two new open-nozzle suppression systems with associated system isolation valves to provide adequate spray coverage to accommodate the final cable tray location, configuration and quantity. These changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not adversely affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not adversely affected. Therefore, the proposed amendment does not involve an increase in the probability or consequences of an accident previously evaluated.

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 clarification of the boundaries for the portion of the nonsafety-related FPS required to remain functional following a SSE for manual firefighting in areas with equipment required for safe shutdown following an SSE 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 changes affect the physical design and operation of the FPS, including as-installed inspections, testing, and maintenance requirements, as described in the Updated Final Safety Analysis Report (UFSAR) due to the addition of two open-nozzle suppression systems with associated system isolation valves. However, the additional open-nozzle suppression systems with associated system isolation valves are similar in design and function as the existing cable tray suppression systems and raceway covers. Therefore, the operation of the FPS is not affected. These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that results in significant fuel cladding failures.

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

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

Response: No.

The proposed clarification of the boundaries for the portion of the FPS required to remain functional following a SSE, and the addition of two new open-nozzle suppression systems with associated system isolation valves do not affect any safety or accident analysis as the FPS is a nonsafety-related system. The only function of the FPS following a design basis earthquake is to provide water for hose valves for manual firefighting in safe shutdown equipment areas. The proposed changes continue to meet the existing design basis, design function, regulatory criterion, or analyses. Therefore, the proposed changes satisfy the same design functions in accordance with the codes and standards currently stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: May 9, 2017. A publicly-available version is in ADAMS under Accession No. ML17129A608.

Description of amendment request: The requested amendments propose to depart from approved AP1000 Design Control Document (DCD) Tier 2 information (text, tables, and figures) as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information, and also propose to depart from involved plant-specific Tier 1 information (and associated combined license (COL) Appendix C information) and from involved plant-specific Technical Specifications as incorporated in Appendix A of the COL. Specifically, the proposed amendments would revise the licensing basis information to reflect design changes to the main control room emergency habitability system (VES) to address the main control room envelope temperature response.

Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific DCD Tier 1 material departures.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. The VES design changes involve: (1) Addition of an automatic and manual, Class 1E, electrical load shed of nonessential nonsafety-related equipment within the main control room envelope (MCRE); and (2) adding a description of the requirements for maintaining habitability of the MCRE beyond 72 hours following a Design Basis Accident to the design and licensing basis. Neither planned or inadvertent operation nor failure of the VES is an accident initiator or part of an initiating sequence of events for an accident previously evaluated. For example, if VES actuation occurs from a loss of power to the plant in a station blackout condition, the additional added features including Wall Panel Information System displays would not be available regardless of the load shed feature. This condition was originally evaluated as part of the AP1000 design certification and no changes are proposed to the plant station blackout response. No additional re-evaluation of other probability or consequences from failures are required to support this change. Therefore, the probabilities of the accidents evaluated in the UFSAR are not affected.

The proposed changes do not have an adverse impact on the ability of the VES to perform its design functions. The design of the VES continues to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the changes maintain the capability of the VES to mitigate the consequences of an accident in conformance with the applicable regulatory acceptance criteria, and there is no adverse effect on any safety-related SSC or function used to mitigate an accident. The changes do not affect the prevention and mitigation of other abnormal events, e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The VES design changes involve: (1) Addition of an automatic and manual, Class 1E, electrical load shed of nonessential nonsafety-related equipment within the MCRE; and (2) adding a description of the requirements for maintaining habitability of the MCRE beyond 72 hours following a DBA to the design and licensing basis. Although a new failure mode of the VES is created by the addition of the MCR Load Shed Panels, neither planned nor inadvertent operation nor failure of the VES is an accident initiator or part of an initiating sequence of events for a new or different kind of accident. In addition, these proposed changes do not adversely affect any other VES or SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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

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

Response: No.

The changes to the VES description and associated COL Appendix A Technical Specification changes provide continued verification that; the VES design functions to maintain heat loads inside the MCRE within design-basis assumptions to limit the heat up of the room, a 72-hour supply of breathable-quality air for the occupants of the MCRE is readily available, and the MCRE pressure boundary is maintained at a positive pressure with respect to the surrounding areas. The changes support the system's intended design functions and continue to meet the regulatory requirements for protecting public health and safety.

The proposed changes also maintain existing safety margins. The proposed changes do not adversely affect VES design requirements and design functions. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while adding additional design features and controls that maintain VES design functions required to meet the existing safety margins. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

Date of amendment request: April 6, 2017. A publicly-available version is in ADAMS under Accession No. ML17097A425.

Description of amendment request: The amendment would revise the Final Safety Analysis Report (FSAR) to allow bypassing of thermal overload protection during motor-operated valve surveillance 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:

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

Response: No.

Keeping the thermal overload protection (TOP) devices bypassed during surveillance testing does not introduce the possibility of a change in the frequency of an accident because failure of a single safety-related motor-operated valve (MOV) is not, by itself, an initiator of any previously evaluated design basis accident. Valves are active components that either position to “open” or “close” as required to fulfill safety functions. As such, safety-related MOVs are subject to single active failures, but such failures are not accident initiators. (For safety-related systems, redundancy in the design ensures that failure of a valve to open or to close on demand, as applicable, will not prevent fulfillment of the safety function(s). However, the associated safety functions are for accident mitigation/response, and while an MOV failure can affect such functions (without loss of the overall function), a single MOV failure cannot by itself initiate any accident previously evaluated in the FSAR.)

Furthermore, the change does not result in an increase in the consequences of an accident previously evaluated in the FSAR. The proposed change would permit MOV TOP devices to remain bypassed during surveillance stroke testing but not during valve maintenance. In regard to the bypassing of TOP devices during testing, the potential for valve damage is of greater concern during valve maintenance activities (when work has been done on the affected valve(s)) than it is for surveillance stroke tests. It may be assumed that the low probability of valve damage resulting from—or occurring during—surveillance valve stroke tests (with the TOP devices bypassed) does not change the single-failure assumptions already considered in the plant's design and accident analyses. As previously noted, redundancy in the design of safety-related systems ensures that failure of a valve to open or close on demand, as applicable, will not prevent fulfillment of the safety function(s). Accordingly, it may be concluded that the provisions for bypassing TOP devices during MOV surveillance testing does not require any changes to assumptions regarding MOV availability, single-failure protection, or the associated systems' capabilities for performing accident mitigation functions. With no changes to such assumptions, the proposed change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR.

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.

NRC [RG] 1.106, Revision 1 [“Thermal Overload Protection for Electric Motors on Motor-Operated Valves”], requires the removal of MOV thermal overload relay bypass jumpers during both maintenance and periodic tests. The regulatory guide's position is that having the thermal overload protection enabled during periodic tests of an MOV is desired to prevent valve motor damage. The concern is that the motor may be damaged if the thermal overload protection is not in force.

Keeping the [TOP] devices bypassed during surveillance testing does not introduce the possibility of an accident of a different type than any previously evaluated in the FSAR. Although there could be a slight increase in the probability of valve damage due to the proposed change, any such failure would not be of a different kind or nature than what may already be experienced by an MOV. Thus, no new failure modes or initiators of a different type of accident are introduced. The single active failure of a[n] [MOV] is already considered in the accident analysis assumptions described in the FSAR, and the failure of a single MOV is not by itself an accident initiator.

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.

No, this change does not affect design basis limits for a fission product barrier. No changes to the accident analyses, including any associated assumptions, are required or being made for the proposed change. Because of redundancy incorporated into the plant design (for single-failure protection), the failure of a single [MOV] will not result in the loss of any overall safety function.

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: John O'Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., Washington, DC 20037.

NRC Branch Chief: Robert J. Pascarelli.

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

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

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

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

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

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: December 9, 2016.

Brief description of amendment: The amendment approved the removal of the existing cyber security license condition from the facility operating license.

Date of issuance: June 22, 2017.

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

Amendment No.: 254. A publicly-available version is in ADAMS under Package Accession No. ML17096A279; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. DPR-72: The amendment revised the license.

Date of initial notice in Federal Register: January 31, 2017 (82 FR 8868).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 22, 2017.

No significant hazards consideration comments received: No.

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River Unit 3 (CR-3) Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: August 31, 2016.

Brief description of amendment: The amendment approved an amendment to the CR-3 Facility Operating License and the Permanently Defueled Technical Specifications to reflect removal of all CR-3 spent nuclear fuel from the spent fuel pools and its transfer to dry cask storage within the independent spent fuel storage installation (ISFSI).

Date of issuance: June 27, 2017.

Effective date: The date Duke Energy Florida, LLC submits written notification that all spent fuel has been transferred from the spent fuel pool to the ISFSI and shall be implemented within 60 days.

Amendment No.: 255.

Facility Operating License No. DPR-72: The amendment revised the license.

Date of initial notice in Federal Register: October 25, 2016 (81 FR 73432).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 27, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 12, 2016, as supplemented by letter dated November 17, 2016.

Brief description of amendment: The amendment reduced the minimum reactor dome pressure associated with the critical power correlation from 785 pounds per square inch gauge (psig) to 686 psig in Technical Specification 2.1.1, “Reactor Core SLs [Safety Limits],” and associated bases.

Date of issuance: June 27, 2017.

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

Amendment No.: 242. A publicly-available version is in ADAMS under Accession No. ML17131A071; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: The license amendment request was originally noticed in the Federal Register on October 25, 2016 (81 FR 73433). Subsequently, by letter dated November 17, 2016, 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 April 25, 2017 (82 FR 19102), which superseded the original notice in its entirety.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 27, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: May 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17139C739.

Brief description of amendment: The amendment revised the Emergency Action Level HU1.5 for James A. FitzPatrick Nuclear Power Plant by replacing the phrase “Lake water level >249.2 ft” with the phrase “A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.”

Date of issuance: June 30, 2017.

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

Amendment No.: 315. A publicly-available version is in ADAMS under Accession No. ML17153A018; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-59: The amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: May 30, 2017 (82 FR 24742).

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

Date of amendment request: May 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17139C739.

Brief description of amendments: The amendments revised the Emergency Action Level HU1.5 for Nine Mile Point Nuclear Station, Units 1 and 2, by replacing the phrase “Lake water level >249.3 ft” with the phrase “A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.”

Date of issuance: June 30, 2017.

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

Amendment Nos.: 228 (Unit 1) and 162 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17152A320; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-63 and NPF-69: Amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: May 30, 2017 (82 FR 24746).

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

No significant hazards consideration comments received: Yes. The comment is addressed in the Safety Evaluation referenced above.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station (Oyster Creek), Ocean County, New Jersey

Date amendment request: February 20, 2017.

Brief description of amendment: The amendment deleted from the Oyster Creek facility operating license certain license conditions that impose specific requirements on the decommissioning trust fund agreement. The provisions of 10 CFR 50.75(h) that specify the regulatory requirements for decommissioning trust funds will apply to Oyster Creek.

Date of issuance: June 23, 2017.

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

Amendment No.: 291. A publicly-available version is in ADAMS under Accession No. ML17067A042; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-16: Amendment revised the Facility Operating License.

Date of initial notice in Federal Register: March 28, 2017 (82 FR 15381).

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: August 22, 2016.

Brief description of amendment: The amendment revised Technical Specification 4.2.1, “Reactor Core, Fuel Assemblies,” and Technical Specification 5.6.5, “Reporting Requirements, Core Operating Limits Report (COLR),” paragraph b, to allow the use of Optimized ZIRLO^TM fuel cladding material. The amendment is also supported by an exemption from certain requirements of 10 CFR 50.46 and 10 CFR part 50, appendix K,

Date of issuance: June 21, 2017.

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

Amendment No.: 125. A publicly-available version is in ADAMS under Accession No. ML17131A066; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-18: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: November 8, 2016 (81 FR 78648).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 21, 2017.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-289 and 50-320, Three Mile Island Nuclear Station, Units 1 and 2, Dauphin County, Pennsylvania

Date of amendment request: July 15, 2016, as supplemented by letter dated February 13, 2017.

Brief description of amendment: The amendment approved changes to the emergency plan that involve on-shift emergency response staffing modifications.

Date of issuance: June 23, 2017.

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

Amendment No.: 291. A publicly-available version is in ADAMS under Accession No. ML17137A393; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-50: Amendment revised the emergency plan.

Date of initial notice in Federal Register: The license amendment request was originally noticed in the Federal Register on October 25, 2016 (81 FR 73435). The supplement dated February 13, 2017, expanded the scope of the application as originally noticed; therefore, the NRC staff renoticed the application in the Federal Register on April 11, 2017 (82 FR 17458).

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

No significant hazards consideration comments received: No.

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

Date of amendment request: October 27, 2016.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.8.3, “Diesel Fuel Oil, Lube Oil, and Starting Air,” by removing the current stored diesel fuel oil and lube oil numerical volume requirements from the TS and replacing them with diesel operating time requirements consistent with NRC-approved Revision 1 to Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-501, “Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.”

Date of issuance: June 29, 2017.

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

Amendment No.: 177. A publicly-available version is in ADAMS under Accession No. ML17163A354; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: December 20, 2016 (81 FR 92869).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 29, 2017

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station (CNS), Nemaha County, Nebraska

Date of amendment request: August 26, 2016.

Brief description of amendment: The amendment revised the CNS Technical Specifications (TSs) to eliminate TS 5.5.6, “Inservice Testing Program,” to remove requirements duplicated in the American Society of Mechanical Engineers Code for Operations and Maintenance of Nuclear Power Plants Case OMN-20, “Inservice Test Frequency.” A new defined term, “Inservice Testing Program,” was added to TS Section 1.1, “Definitions.” The licensee stated that the change to the TSs is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” which was made available to the TSTF via NRC letter dated December 11, 2015 (ADAMS Accession No. ML15317A071), with no proposed technical variations or deviations. However, in some cases, the CNS TSs use different section titles or numbering for SRs than the Standard Technical Specifications on which TSTF-545 was based. The licensee changed the TSTF-545 numbering to be consistent with the CNS TS numbering.

Date of issuance: June 20, 2017.

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

Amendment No.: 259. A publicly-available version is in ADAMS under Accession No. ML17144A082; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-46: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: November 8, 2016 (81 FR 78649).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 20, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 31, 2016, as supplemented by letter dated February 16, 2017.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.8.7 by removing the site-specific Required Actions and associated Completion Times, thus reverting to the standard TS language contained in NUREG-1431, “Standard Technical Specifications: Westinghouse Plants.”

Date of issuance: June 20, 2017.

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

Amendment Nos.: 219 (Unit 1) and 206 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17130A716; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: October 25, 2016 (81 FR 73436). The supplemental letter dated February 16, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 20, 2017.

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 Units 1 and 2), Salem County, New Jersey

Date of amendment request: August 30, 2016.

Brief description of amendments: The amendments approved adoption of NRC-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” dated October 21, 2015. Specifically, the amendments deleted the Salem Units 1 and 2 Technical Specification (TS) Section 6.8.4.j, “Inservice Testing Program,” and added a new defined term, “INSERVICE TESTING PROGRAM,” to the TSs. All existing references to the “Inservice Testing Program” in the Salem Units 1 and 2 TS SRs are replaced with “INSERVICE TESTING PROGRAM” so that the SRs refer to the new definition in lieu of the deleted program.

Date of issuance: June 28, 2017.

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

Amendment Nos.: 319 (Unit No. 1) and 300 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17165A214; 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 TSs.

Date of initial notice in Federal Register: November 8, 2016 (81 FR 78651).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 28, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 20, 2016.

Brief description of amendment: The amendment approved adoption of NRC-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” dated October 21, 2015. Specifically, the amendment deleted the Hope Creek Technical Specification (TS) Section 6.8.4.i, “Inservice Testing Program,” and added a new defined term, “INSERVICE TESTING PROGRAM,” to the TSs. All existing references to the “Inservice Testing Program” in the Hope Creek TS SRs are replaced with “INSERVICE TESTING PROGRAM” so that the SRs refer to the new definition in lieu of the deleted program.

Date of issuance: June 28, 2017.

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

Amendment No.: 205. A publicly-available version is in ADAMS under Accession No. ML17164A355; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-57: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: October 25, 2016 (81 FR 73437).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 28, 2017.

No significant hazards consideration comments received: No.

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, Burke County, Georgia

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

Date of amendment request: December 1, 2016.

Brief description of amendments: The amendments modified the Technical Specification (TS) requirements in Section 1.3 and Section 3.0 regarding Limiting Conditions for Operation (LCO) and Surveillance Requirement (SR) usage. The changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-529, Revision 4, “Clarify Use and Application Rules.”

Date of issuance: June 27, 2017.

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

Amendment Nos.: Farley—211 (Unit 1) and 208 (Unit 2); Vogtle—187 (Unit 1) and 168 (Unit 2); and Hatch—285 (Unit No. 1) and 230 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17137A041; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-2, NPF-8, NPF-68, NPF-81, DPR-57, and NPF-5: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: February 28, 2017 (82 FR 12135).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 27, 2017.

No significant hazards consideration comments received: No.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, Burke County, Georgia

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

Date of amendment request: July 28, 2016.

Brief description of amendments: The amendments modified the technical specifications (TSs) to eliminate Section 5.5.8, “Inservice Testing Program,” for Farley and Vogtle, and eliminate Section 5.5.6, “Inservice Testing Program,” for Hatch. A new defined term, “Inservice Testing Program,” is added to the TS Definitions section. This request is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing”.

Date of issuance: June 30, 2017.

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

Amendment Nos.: Farley—212 (Unit 1) and 209 (Unit 2); Vogtle—187 (Unit 1) and 170 (Unit 2); and Hatch—286 (Unit No. 1) and 231 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17152A218; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-2, NPF-8, NPF-68, NPF-81, DPR-57, and NPF-5: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: September 27, 2016 (81 FR 66309).

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

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 50-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

Date of amendment request: August 29, 2016, as supplemented by letter dated February 13, 2017.

Brief description of amendments: The amendments changed Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating Plant, Units 3 and 4. The amendments changed the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2* information. Specifically, the amendment proposed changes to demonstrate the quality and strength of a specific population of welds between stainless steel mechanical couplers (couplers) and embedment plates that did not receive the nondestructive examinations required by the American Institute of Steel Construction N690-1994, “Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities.” Since some of these coupler welds are already installed and embedded in concrete, the licensee proposed to demonstrate the adequacy of these inaccessible coupler welds through previously-performed visual testing examinations of the couplers and static tension testing of a representative sample of accessible, uninstalled couplers produced concurrently with those already installed.

Date of issuance: June 27, 2017.

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

Amendment Nos.: 80 (Unit 3) and 79 (Unit 4). A publicly-available version is in ADAMS under Package Accession No. ML17107A275; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License No. NPF-91 and NPF-92: Amendments revised the UFSAR in the form of departures from the incorporated plant-specific DCD Tier 2* information.

Date of initial notice in Federal Register: November 8, 2017 (81 FR 78666). The supplement, dated February 13, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 27, 2017.

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: October 17, 2016, as supplemented by letter dated March 6, 2017.

Brief description of amendments: The amendments revised selected Technical Specification (TS) Surveillance Requirements (SRs) for alternating current electrical sources because of delays in the startup of Watts Bar Nuclear Plant, Unit 2. Specifically, the amendments revised the TSs to permit a one-time extension of the specified 18-month interval for performing the required SRs.

Date of issuance: June 28, 2017.

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

Amendment Nos.: 114 (Unit 1) and 12 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17138A100; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in the Federal Register: February 28, 2017 (82 FR 12138). The supplemental letter dated March 6, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the 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 June 28, 2017.

No significant hazards consideration comments received: No.

TEX Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant (CPNPP), Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: December 14, 2016.

Brief description of amendments: The amendments revised the licensee name from “TEX Operations Company LLC” to “Vistra Operations Company LLC” in the CPNPP, Unit No. 1, Facility Operating License (FOL) NPF-87; CPNPP, Unit No. 2, FOL (NPF-89); and the title page of the Environmental Protection Plan.

Date of issuance: June 29, 2017.

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

Amendment Nos.: 169 (Unit 1) and 169 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17129A024; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: February 28, 2017 (82 FR 12139).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 29, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 6th day of July 2017.

For the Nuclear Regulatory Commission.

Eric J. Benner,

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