82 Fed. Reg. 42844 (Sep. 12, 2017)

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

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Federal RegisterSep 12, 2017

82 Fed. Reg. 42844 (Sep. 12, 2017)

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 August 15, 2017 to August 28, 2017. The last biweekly notice was published on August 29, 2017.

DATES:

Comments must be filed by October 12, 2017. A request for a hearing must be filed by November 13, 2017.

ADDRESSES:

You may submit comments by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0189. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. 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:

Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-2242; email: Paula.Blechman@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0189, facility name, unit numbers, 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-2017-0189.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2017-0189, facility name, unit numbers, plant docket number, application date, and subject in your comment submission.

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

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment 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 hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public 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 MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

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

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

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

Date of amendment request: June 29, 2017. A publicly available version is in ADAMS under Accession No. ML17180A538.

Description of amendment request: The amendments would adopt changes, with variations, based on the NRC-approved safety evaluation of Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control,” dated December 20, 2016 (ADAMS Package Accession No. ML16343B066). The revisions would replace existing technical specification (TS) requirements related to “operations with a potential for draining the reactor vessel” (OPDRVs) with new requirements on reactor pressure vessel water inventory control (RPV WIC) to protect Safety Limit 2.1.1.3, which requires reactor vessel water level to be greater than the top of active irradiated fuel.

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

The proposed change reduces the probability of an unexpected draining event, which is not a previously evaluated accident, by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event. The proposed change reduces the consequences of an unexpected draining event, which is not a previously evaluated accident, by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed. The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

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

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

Response: No.

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

Thus, based on the above, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

Response: No.

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

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

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

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

NRC Branch Chief: Undine Shoop.

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

Date of amendment request: July 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17208A428.

Description of amendment request: The proposed amendment would revise certain staffing and training requirements, reports, programs, and editorial changes in the Technical Specifications (TSs) Table of Contents; Section 1.0, “Use and Application”; and Section 5.0, “Administrative Controls,” that will no longer be applicable once PNP is permanently defueled.

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

Response: No.

The proposed amendment would not take effect until the PNP Certified Fuel Handler Training and Retraining Program has been approved by the NRC, and PNP has permanently ceased operation and entered a permanently defueled condition. The proposed changes would revise the PNP TS by modifying the definitions, in TS Section 1.0, and administrative controls, in TS Section 5.0, to correspond to the permanently defueled condition. Additionally, certain portions of the administrative control sections are deleted because they are no longer applicable to a permanently defueled facility.

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

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

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

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

Response: No.

The proposed amendment has no impact on facility systems, structures, and components (SSCs) affecting the safe storage of spent nuclear fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent nuclear fuel itself. The proposed amendment does not result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled, and PNP will no longer be authorized to operate the reactor or retain or place fuel in the reactor vessel.

The proposed amendment does not affect systems credited in the PNP accident analysis for a[n] FHA, or for mitigating accident releases from the failure of tanks containing radioactive liquids or from a spent fuel cask drop. The proposed changes will continue to require proper control and monitoring of safety significant parameters and activities.

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

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

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

Response: No.

The proposed amendment involves deleting and/or modifying certain TS requirements once the PNP has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for PNP will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). Therefore, the occurrence of postulated accidents associated with reactor operation are no longer credible.

The only remaining credible accidents are the fuel handling accident (FHA), the failure of tanks containing radioactive liquids, and a spent fuel cask drop accident. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact these analyzed conditions.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: July 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17199F854.

Description of amendment request: The proposed change would revise the design value for the spent fuel storage pool in Technical Specification (TS) 4.3.2, “Drainage,” to an appropriate value, consistent with the original design basis.

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

Response: No.

No physical changes to the facility will occur as a result of this proposed amendment. The proposed changes will not alter the physical design. The proposed change will revise the current TS 4.3.2 value for the SFP [spent fuel pool] level design to be consistent with the original design basis value and the applicable regulatory requirements. The proposed value will continue to ensure that inadvertent draining of the SFP will not result in the uncovering of spent fuel, as well as provide adequate shielding for personnel protection.

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

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

Response: No.

The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function. The proposed change will revise the current TS 4.3.2 value for the SFP level design to be consistent with the original design basis value and the applicable regulatory requirements. The proposed value will continue to ensure that inadvertent draining of the SFP will not result in the uncovering of spent fuel, as well as provide adequate shielding for personnel protection.

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

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

Response: No.

The proposed change conforms to NRC regulatory guidance regarding the content of plant Technical Specifications. The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant.

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 the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: July 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17200D096.

Description of amendment request: The amendments would replace existing technical specification (TS) requirements related to “operations with a potential for draining the reactor vessel” (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires RPV water level to be greater than the top of active irradiated fuel. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-542, “Reactor Pressure Vessel Water Inventory Control,” Revision 2 (ADAMS Package Accession No. ML16250A231).

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. Draining of RPV water inventory in OPERATIONAL CONDITION 4 (i.e., cold shutdown) and OPERATIONAL CONDITION 5 (i.e., refueling), is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in OPERATIONAL CONDITION 4 or 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

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

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

The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in OPERATIONAL CONDITIONS 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

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

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

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

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

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.4. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the TAF in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

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

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

NRC Branch Chief: James G. Danna.

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

Date of amendment request: January 23, 2017, as supplemented by letter dated July 3, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17025A399 and ML17184A176, respectively.

Description of amendment request: The license amendment request was originally noticed in the Federal Register on March 28, 2017 (82 FR 15383). The notice is being reissued in its entirety to include the revised scope, description of the amendment request, and proposed no significant hazards consideration determination. As a result of the revised scope, updates to the “Basis for proposed no significant hazards consideration determination” section of this notice are delineated by brackets.

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

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

Response: No.

Limiting the MODE 1 applicability for RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤15% of RATED THERMAL POWER, is an administrative change in nature and does not alter the manner in which the functional unit is operated or maintained. The proposed changes do not represent any physical change to plant [structures, systems, and components (SSC(s))], or to procedures established for plant operation. The subject RPS functional unit is not an event initiator nor is it credited in the mitigation of any event or credited in the [probabilistic risk assessment (PRA)]. As such, the initial conditions associated with accidents previously evaluated and plant systems credited for mitigating the consequences of accidents previously evaluated remain unchanged.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 [ADAMS Accession No. ML12102A165] (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

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

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

Response: No.

Limiting the MODE 1 applicability for the RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤15% of RATED THERMAL POWER, is an administrative change in nature and does not involve the addition of any plant equipment, methodology or analyses. The proposed changes do not alter the design, configuration, or method of operation of the subject RPS functional unit or of any other SSC. More specifically, the proposed changes neither alter the power rate-of-change trip function nor its ability to bypass and reset as required. The subject RPS functional unit remains capable of performing its design function.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

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.

Limiting the MODE 1 applicability for RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤15% of RATED THERMAL POWER is an administrative change in nature. The proposed changes neither involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings nor do they adversely impact plant operating margins or the reliability of equipment credited in safety analyses.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

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

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

NRC Branch Chief: Undine Shoop.

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

Date of amendment request: June 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17195A569.

Description of amendment request: The amendments would modify the Technical Specification (TS) requirements for mode change limitations in TS 3.0.4 and TS 4.0.4 based on Technical Specifications Tasks Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-359, Revision 9, “Increase Flexibility in MODE Restraints” (ADAMS Accession No. ML031190607).

The NRC issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration determination, using the consolidated line item improvement process (CLIIP). Subsequently, on April 4, 2003, the NRC published a Notice of Availability for TSTF-359, Revision 8, in the Federal Register (68 FR 16579). That notice announced the availability of this TS improvement through the CLIIP. The NRC subsequently made two modifications in response to comments, as well as one editorial change, which have been incorporated into TSTF-359, Revision 9. The changes proposed in the licensee's submittal are, therefore, based on TSTF-359, Revision 9.

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

Response: No.

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS Action. Being in a TS Action is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on Actions as allowed by the proposed LCO 3.0.4 are no different than the consequences of an accident while relying on Actions for other reasons, such as equipment inoperability. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS while in a TS Action, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

Response: No.

The proposed change allows entry into a mode or other specified condition in the applicability of a TS while in a TS Action. The TS allow operation of the plant without the full complement of equipment through the Actions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of Actions that must be performed within the prescribed completion times. The net effect of being in a TS Action on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS Actions to be entered and the associated required actions and completion times to be used in new circumstances. This use is predicated upon performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing Actions in similar circumstances without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: Undine Shoop.

National Institute of Standards and Technology (NIST), Docket No. 50-184, Center for Neutron Research Test Reactor, Montgomery County, Maryland

Date of amendment request: March 2, 2017 (two letters), as supplemented by letters dated March 29, 2017, and May 25, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17068A163, ML17068A164, ML17097A243, and ML17153A172, respectively.

Description of amendment request: The proposed amendment would modify the NIST test reactor's technical specifications (TSs) to remove limitations in the present version of the TSs that prohibit use of a test procedure and to change the organizational chart in the TSs. In addition, the proposed amendment would modify the NIST test reactor's license to allow transfer of instrumentation calibration and testing sources from the NIST's material license to the reactor license.

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

Response: No.

No, the proposed amendment would not increase the probability or consequences of an accident previously evaluated. The proposed amendment removes conformance conflicts within the Technical Specifications that would occur when operating the reactor as permitted under TSs 2.2(4). The conflicts are removed from the TSs by adding exception statements. When the reactor is operated under the NRC approved conditions in TSs 2.2(4), steady state thermal hydraulic analysis shows that operation at less than 500 kW [kilowatt] with natural circulation results in a critical heat flux ratio and onset of flow instability ratio greater than 2. Transient analysis of reactivity insertion accidents shows that the fuel cladding temperature remains far below the safety limit. The limit of 10 kw was chosen since that was deemed adequate for any operational situation requiring natural circulation operation, such as testing of an unknown core loading.

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

Response: No.

No, the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment removes conformance conflicts within the Technical Specifications that would occur when operating the reactor as permitted under TSs 2.2(4). The conflicts are removed from the TSs by adding exception statements. The accident analysis was discussed in the document, NIST Response to NRC Request for Information (TAC No. MD3410), August 19, 2008, ADAMS Accession Number ML082890338. The request from the NRC was: “. . . Provide justification for 500 kW power operations under natural convection flow by demonstrating that no credible accidents would result in exceeding the safety limit. . . ,” the following was the response by NIST. “This analysis shows that there is ample margin between the maximum clad temperature in any credible accident and the safety limit of 450 °C [degrees Centigrade].” The details of the analysis are presented in the above reference.

The intent with this amendment is to allow, without apparent TSs nonconformance, operation analyzed and evaluated by the NRC. This will allow the use of testing similar to that which was performed in the commissioning of NIST test reactor.

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

Response: No.

No, the proposed amendment would not involve a significant reduction in a margin of safety. This amendment will allow testing when commissioning a core configuration that is unknown in the most conservative manner appropriate. It removes apparent TS conflicts that would force the licensee into situations that would be less conservative and with less margin of safety.

Attorney for licensee: Melissa J. Lieberman, Deputy Chief Counsel for NIST, National Institute of Standards and Technology, 100 Bureau Drive, Gaithersburg, MD 20899.

NRC Branch Chief: Alexander Adams, Jr.

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: April 21, 2017, as supplemented by letter dated August 15, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17111A958, and ML17227A775, respectively.

Description of amendment request: The amendment request proposes 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 proposes to depart from involved plant-specific Tier 1 information (and associated Combined License (COL) Appendix C information). Specifically, the amendment request proposes changes to COL Appendix C (and plant-specific Tier 1) Table 2.2.4-1 and Figure 2.2.4-1 to add two main feedwater thermal relief valves and two start-up feedwater thermal relief valves. The proposed COL Appendix C (and plant-specific DCD Tier 1) changes require additional changes to corresponding Tier 2 information in UFSAR Chapters 3 and 10. 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).

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

Response: No.

The changes to Combined License (COL) Appendix C (and plant-specific Tier 1) Table 2.2.4-1 and Figure 2.2.4-1, and associated Updated Final Safety Analysis Report (USFAR) design information do not adversely impact previously evaluated accidents. The addition of the thermal relief valves to the feedwater lines does not adversely impact the ability to isolate the main and startup feedwater lines following a steam or feedwater line break or steam generator tube rupture. The new thermal relief valves are normally closed and required to open to prevent potential overpressure conditions when ambient temperatures increase in the area. Thermal relief valves added into the feedwater lines operate mechanically and are not activated upon a new engineered safety features (ESF) signal in response to design basis accidents. Isolation capabilities of the main and startup feedwater lines are not adversely affected as ESF signals are not changed. The proposed change does not reduce the temperature of feedwater and does not increase feedwater flow during any operational mode as main feedwater and startup feedwater isolation and control valves are not changed by this activity. Performance of overpressure relief supports the safety-related functions of the isolation and control valves in the main and startup feedwater lines when isolation is required.

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.

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 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design information do not reduce the temperature of feedwater and do not increase feedwater flow during any operational mode such that it would result in a new or different kind of accident from accidents previously evaluated. Conclusions of existing analyses are not changed by this activity as existing feedwater isolation and control valves functions are not changed.

The proposed changes to add thermal relief valves to the main and startup feedwater lines do not adversely affect any safety-related equipment, and do not add any new interfaces to safety-related SSCs that adversely affect safety functions. No system or design function or equipment qualification is adversely affected by these changes as the changes do not modify any SSCs that prevent safety functions from being performed by the existing main feedwater and startup feedwater valves. The changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety-related equipment as feedwater isolation capabilities are not changed. Performance of overpressure relief supports the safety-related functions of the isolation and control valves in the main and startup feedwater lines when isolation is required.

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

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

Response: No.

The proposed changes to COL Appendix C (and plant-specific Tier 1) Table 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design information add thermal relief valves to the main feedwater and startup feedwater lines. These valves are designed to the same codes and standards as the existing piping to which they are connected, including ASME Code Section III, Class C, seismic Category I. The proposed changes do not affect any other safety-related equipment or fission product barriers. The requested changes will not 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 requested changes. There are not any changes to operation of the main feedwater and startup feedwater isolation and control valves when isolation of the lines is required. Operation of the relief valves supports isolation capabilities for the main and feedwater isolation and control valves.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: July 14, 2017. A publicly-available version is in ADAMS under Accession No. ML17195B047.

Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific design control document (DCD) Tier 2 information) and involves related changes to plant-specific Tier 1 (and associated Combined License (COL) Appendix C) information, and COL Appendix A Technical Specifications. Specifically, the requested amendment proposes changes to add a second normal residual heat removal system (RNS) suction relief valve in parallel to the current RNS suction relief valve, with the necessary piping changes. Additionally, a change is proposed to Tier 1 Figure 2.2.1-1, for penetration P19, to accurately depict the orientation of the class break of containment isolation valve RNS-PL-V061.

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 Combined License (COL) Appendix C (and plant-specific Tier 1) Figures 2.2.1-1 and 2.3.6-1, Tables 2.3.6-1, 2.3.6-2 and 2.3.6-4, COL Appendix A, Technical Specification 3.4.14 and associated Updated Final Safety Analysis Report (UFSAR) design information to identify a new normal residual heat removal system (RNS) relief valve, RNS-PL-V020, do not adversely impact accidents previously evaluated in the safety analysis. Transients that are capable of overpressurizing the reactor coolant system (RCS) are categorized as either mass or heat input transients. The relief valves must be capable of passing flow greater than that required for the limiting low-temperature overpressure protection (LTOP) transients while maintaining RCS pressure less than the lowest pressure represented by the pressure/temperature limit curve, 110% of the design pressure of the RNS, or the acceptable RNS relief valve inlet pressure. The restrictions added to COL Appendix A, Technical Specification 3.4.14 to close chemical and volume control system (CVS) makeup line containment isolation valve, CVS-PL-V091, limit flow capacity when the RCS is aligned to the RNS to support LTOP functions and provide reliable operation of the RNS relief valves during mass and heat input transients. When CVS-PL-V091 is open, the RCS is depressurized and an RCS vent of ≥4.15 square inches is established. Transient conditions including mass input and heat input are not changed and probability of events is not increased as the added RNS relief valve, RNS-PL-V020, supports LTOP functions as required by Technical Specification 3.4.14. The current 3-inch RNS relief valve is sufficient to terminate identified transients; however, the added 1-inch RNS relief valve reduces chatter in the current valve during low flow scenarios.

Responses to mass and heat input transients are not changed as LTOP functions to prevent overpressurization of the RCS are not changed by this activity. The added RNS relief valve, RNS-PL-V020, is designed in accordance with the same requirements as the current RNS relief valve, RNS-PL-V021, but with a lower flow capacity and functions at a lower setpoint pressure. Overpressure protection provided by the RNS is not changed. The change does not adversely impact the capability of the RNS to protect the RCS from exceeding pressure and temperature limits in accordance with 10 CFR 50, Appendix G or 110% of the design pressure of the RNS. Changes in piping to accommodate the addition of the valve and reduce inlet piping losses do not impact the consequences or probabilities of previously evaluated accidents. The class break correction for valve RNS-PL-V061, in COL Appendix C (and plant-specific Tier 1) Figure 2.2.1-1 does not impact accidents previously evaluated.

No safety-related structure, system, component (SSC) or function is adversely affected by this change. The change does not involve an interface with any structure, system, or component (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.

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.

Conclusions of existing analyses are not changed by the proposed change as LTOP functions provided by both the current and added RNS relief valves continue to provide the assumed protection for LTOP events. RCS pressure is maintained within limits by the use of both RNS relief valves. The closure of CVS-PL-V091 limits flow and reduces the impact of mass and heat input transients when RNS relief valves are relied upon for overpressure protection.

The proposed change to add the smaller RNS relief valve, RNS-PL-V020, does not adversely affect safety-related equipment, and does not add any new interfaces to safety-related SSCs that adversely affect safety functions. The added RNS relief valve, functions in the same manner as the current RNS relief valve, but has a lower capacity and lifts at a lower pressure. The added RNS relief valve also discharges to the liquid radwaste system (WLS) containment sump. No system or design function or equipment qualification is adversely affected by these changes as the change does not modify any SSCs that prevent safety functions from being performed by the RNS and the current relief valve. The changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety-related equipment. Piping changes to accommodate the installation of the new valve do not create the potential for a new or different kind of accident as the piping requirements are consistent with those of the current relief valve, and subject to the same pipe rupture evaluation requirements. LTOP functions are not changed. The class break correction for valve RNS-PL-V061 does not impact accident analysis or create a new or different kind of accident as the function of the affected equipment and piping is not changed.

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

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

Response: No.

The proposed changes do not affect safety-related equipment or fission product barriers. LTOP functions are not adversely impacted as both the current and added RNS relief valves continue to provide protection from overpressurization. The added RNS relief valve is designed in accordance with [American Society of Mechanical Engineers (ASME)] Code Section III, Class 2, requirements consistent with the current RNS relief valve. Modified piping is constructed consistent with current design requirements for RNS piping. The addition of the valve adds safety margin in regards to transients as the new valve lifts at a lower set pressure than the current valve, causing flow rates to be lower through the RNS piping. Therefore, margin of safety is not reduced. The requested changes will not 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 requested changes. Transient conditions, including mass input and heat input, are not changed and margin of safety is not reduced as the added RNS relief valve supports LTOP functions in the same manner as the current RNS relief valve.

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

Attorney for licensee: Mr. 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 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17151A296.

Description of amendment request: The requested amendment proposes 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 from involved plant-specific Technical Specifications as incorporated in Appendix A of the combined license. Specifically, the proposed changes support the addition of chemicals necessary to achieve proper reactor coolant system (RCS) water quality by allowing an unborated water source through the chemical mixing tank to be unisolated for ≤1 hour for chemical addition to the pressurizer to be performed with reactor coolant pumps (RCPs) not in operation. In order to perform chemical addition to the pressurizer without the mixing provided by forced reactor coolant system (RCS) flow, administrative controls are established such that coolant introduced into the RCS is at a boron concentration greater than or equal to that required to meet the shutdown margin (SDM) boron concentration.

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

Response: No.

Updated Final Safety Analysis Report (UFSAR) 15.4.6, Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant, addresses inadvertent boron dilution events. The principal means of positive reactivity insertion to the core is the addition of unborated, primary-grade water from the demineralized water transfer and storage system (DWS) into the reactor coolant system (RCS) through the reactor makeup portion of the chemical and volume control system (CVS).

These events are primarily evaluated with one or more reactor coolant pumps (RCPs) in operation providing adequate mixing. The changes proposed by this amendment request do not involve operations where the RCPs are in operation. Therefore, there is no increase in the probability or consequences of inadvertent boron dilution events with RCPs operating.

UFSAR Subsection 15.4.6 also describes that when a reactor coolant pump is not operating, the demineralized water isolation valves are closed and an uncontrolled boron dilution transient cannot occur. The proposed amendment adds provisions to allow a specific CVS unborated water source flow path to be opened through the chemical mixing tank to the RCS pressurizer when RCPs are not in operation for the purpose of chemical addition to the pressurizer. The administrative control provisions proposed provide adequate assurance that any injection to the RCS pressurizer would only occur such that injected water is limited to boron concentrations greater than the required concentrations to meet the SDM. With no reduction in SDM, there would be no means of positive reactivity insertion to the core leading to an adverse reactivity event. As such, there is no significant increase in the probability of a previously evaluated boron dilution event as a result of this change.

Since the proposed change does not lead to any positive reactivity insertion, there are no increased 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 administrative control provisions proposed provide adequate assurance that any injection to the pressurizer would only occur such that injected water is limited to boron concentrations greater than the required concentrations to meet the SDM. With no reduction in SDM, there would be no means of positive reactivity insertion to the core leading to an adverse reactivity event. Failure modes involving procedural controls and operator actions are considered in evaluating inadvertent boron dilution events. The possibility of a new or different kind of failure, malfunction, or sequence of events has been evaluated with these proposed changes; events are precluded with the proposed administrative controls and defense in depth features inherent in the AP1000 design.

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

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

Response: No.

The margin of safety is established by maintaining the required SDM during shutdown activities. The proposed changes to the UFSAR and Technical Specifications do not adversely affect the safety-related functions of the RCS or CVS in maintaining adequate SDM. Provisions are proposed for a specific CVS unborated water source flow path to be opened through the chemical mixing tank to the RCS pressurizer when RCPs are not in operation; however, this activity is performed under administrative controls that preclude the potential for a reduction in SDM.

The changes do not affect containment penetrations or any other safety-related equipment or fission product barriers. The requested changes will not 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 requested changes. The existing design and operation of the associated systems are adequate to preclude an inadvertent boron dilution from occurring when RCPs are not in operation.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: July 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17209A185.

Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD information, and also proposes to depart from involved plant-specific Tier 1 information and the associated combined license (COL) Appendix C information. Specifically, the amendment, if approved, would revise the COL documents mentioned previously to reflect the proposed changes to update Reactor Coolant System (RCS) requirements for the reactor vessel head vent (RVHV) mass flow rate. 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.

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

Response: No.

UFSAR Subsections 15.2.7, 15.5.1, and 15.5.2 describe analyses performed for an increase in reactor coolant inventory due to a loss of normal feedwater flow, and for malfunctions of the chemical and volume control system and the core makeup tanks. In each of these evaluated accidents, it is assumed that the operators are alerted to the event due to a high pressurizer water level and take subsequent action to open the reactor vessel head vent valves. When the head vent is opened, the pressurizer water level increase slows and eventually decreases.

Changing the required mass flow rate from 8.2 lbm/sec at a Reactor Coolant System (RCS) pressure of 1250 psia [pounds per square inch absolute] to 9.0 lbm/sec [pounds mass per second] at an RCS pressure of 2500 psia for the reactor vessel head vent (RVHV) flow path does not change the probability of these events occurring. The valves are used to mitigate the events. They are not an initiator of these accidents, or any other accident previously evaluated. Changing the required mass flow rate does not change the consequences of these accidents. The proposed flow rate change is made to be consistent with the latest AP1000 safety analysis. This change does not lead to an increase in the probability of a loss of coolant accident, nor does it cause the RVHV to exceed the capability of the normal makeup system. The changes described above continue to ensure the design is capable of providing adequate flow rate for emergency letdown and the prevention of long term pressurizer overfill.

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 impact the acceptance criteria for RVHV mass flow rate. The required mass flow rate is changed from 8.2 lbm/sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS pressure of 2500 psia to align with the events evaluated in the current safety analysis. The proposed changes do not result in a new accident initiator and do not impact a current accident initiator.

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

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

Response: No.

The proposed changes impact the acceptance criteria for RVHV mass flow rate. The required mass flow rate is changed from 8.2 lbm/sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS pressure of 2500 psia. The proposed changes are made to reflect the updated AP1000 plant safety analysis; the changes are conservative and bound the expected performance of the as-built equipment.

COL Appendix C (plant-specific Tier 1) is proposed to be updated to reflect the new mass flow rate through the RVHV line and the associated system pressure. COL Appendix C (plant-specific Tier 1) is updated to reflect the latest safety analysis, which credits an emergency letdown mass flow rate of 9.0 lbm/sec at an RCS pressure of 2500 psia. At these conditions, long term pressurizer overfill is prevented. RCS calculations show that the expected mass flow rate through the emergency letdown path is 12.34 lbm/sec. Therefore, the safety analysis calculation, and the corresponding mass flow rate and RCS pressure values used in the proposed changes, is conservative and bounded by the expected mass flow rate.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: July 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17212A842.

Description of amendment request: The amendment would revise the staffing and staff augmentation times described in the South Texas Project Emergency Plan. The proposed amendment would increase the Emergency Response Organization (ERO) response times and would modify minimum staffing functions and requirements of the ERO and Operations Support Center staff. The changes also include formatting, clarification, and editorial modifications.

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

Response: No.

The proposed amendment has no effect on normal plant operation or on any accident initiator or precursors and does not impact the function of plant structures, systems, or components. The proposed changes do not alter or prevent the ability of the Emergency Response Organization to perform their intended functions to mitigate the consequences of an accident or event.

Therefore, the proposed STPEGS [South Texas Project Electric Generating Station] Emergency Plan change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment does not impact any accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed change does not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed change revises the on-shift staffing and staff augmentation response times in the STPEGS Emergency Plan. The proposed changes do not alter or prevent the ability of the Emergency Response Organization to perform their intended functions to mitigate the consequences of an accident or event.

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.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change is associated with the STPEGS Emergency Plan staff and staff augmentation and does not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed change does not involve a change in the method of plant operation and no accident analyses will be affected by the proposed change. Safety analysis acceptance criteria are not affected by the proposed change. The revised STPEGS Emergency Plan will continue to provide the necessary response staff with the proposed change. Therefore, the proposed change is determined to not adversely affect the ability to meet the requirements of 10 CFR 50.54(q)(2), 10 CFR 50 Appendix E, or the emergency planning standards described in 10 CFR 50.47(b).

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

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

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

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

Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam Electric Plant Unit No. 2 (Robinson), Darlington County, South Carolina

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

Date of amendment request: April 29, 2016, as supplemented by letters dated October 3, 2016, and January 16, 2017.

Brief description of amendments: The amendments (1) consolidated the Emergency Operations Facilities (EOFs) for Brunswick, Harris, and Robinson with the Duke Energy Progress, LLC (Duke Energy) corporate EOF in Charlotte, North Carolina; (2) decreased the frequency for a multisite drill at Oconee from once per 6 years to once per 8 years; (3) allowed the multisite drill performance with sites other than the Catawba Nuclear Station, McGuire Nuclear Station, or Oconee; (4) changed the Brunswick, Harris, and Robinson augmentation times to be consistent with those of the sites currently supported by the Duke Energy corporate EOF; and (5) decreased the frequency of the unannounced augmentation drill at Brunswick from twice per year to once per year.

Date of issuance: August 21, 2017.

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

Amendment Nos.: 279 and 307 for Brunswick, Units 1 and 2; 160 for Harris, Unit 1; 254 for Robinson Unit No. 2; and 405, 407, and 406 for Oconee, Units 1, 2, and 3. A publicly-available version is in ADAMS under Accession No. ML17188A387; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-71 and DPR-62 for Brunswick, Units 1 and 2; NPF-63 for Harris, Unit 1; DPR-23 for Robinson Unit No. 2; and DPR-38, DPR-47, and DPR-55 for Oconee, Units 1, 2, and 3: The amendments revised the emergency plans.

Date of initial notice in Federal Register: July 5, 2016 (81 FR 43650). The supplemental letters dated October 3, 2016, and January 16, 2017, provided additional information that expanded the scope of the application as originally noticed and changed the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. Accordingly, the NRC published a second proposed no significant hazards consideration determination in the Federal Register on February 14, 2017 (82 FR 10594). This notice superseded the original notice in its entirety.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 28, 2016, as supplemented by letters dated February 23, 2017, and June 21, 2017.

Brief description of amendment: The amendment revised the current emergency action level scheme to one based on Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors” (ADAMS Accession No. ML12326A805). Revision 6 of NEI 99-01 was endorsed by the NRC in a letter dated March 28, 2013.

Date of issuance: August 28, 2017.

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

Amendment No.: 244. A publicly-available version is in ADAMS under Accession No. ML17188A230; 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 Operating License.

Date of initial notice in Federal Register: September 27, 2016 (81 FR 66305). The supplemental letters dated February 23, 2017, and June 21, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the 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 August 28, 2017.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 7, 2016, as supplemented by letter dated March 20, 2017.

Brief description of amendments: The amendments revised the Updated Final Safety Analysis Report (UFSAR) to identify the TORMIS Computer Code as the methodology used for assessing tornado-generated missile protection of unprotected plant structures, systems and components (SSCs) and to describe the results of the Byron Station site-specific tornado hazard analysis.

Date of issuance: August 10, 2017.

Effective date: As of the date of issuance and shall be implemented within 60 days of the date of issuance. The UFSAR changes shall be filed with the NRC in the next periodic update to the UFSAR scheduled for December 15, 2018.

Amendment Nos.: 199 for NPF-37 and 199 for NPF-66. A publicly-available version is in ADAMS under Accession No. ML17188A155; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-37, and NPF-66: The amendments revised the current licensing basis as described in the UFSAR.

Date of initial notice in Federal Register: December 6, 2016 (81 FR 87969). The March 20, 2017, supplement contained clarifying information and did not change the scope of the proposed action or affect the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 10, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 30, 2017.

Brief description of amendments: The amendments modified requirements on control and shutdown rods, and rod and bank position indication for Beaver Valley, Unit No. 2. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1, “Clarification of Rod Position Requirements.” Additional supporting changes to Beaver Valley, Unit Nos. 1 and 2, Technical Specifications were also made.

Date of Issuance: August 16, 2017.

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

Amendment Nos.: 299 (Unit No. 1) and 188 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17221A280; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 11, 2017 (82 FR 32017).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 16, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 21, 2015, as supplemented by letters dated November 13, December 15 (two letters), and December 18, 2015; February 16, March 8, March 9, March 24, March 28, April 4, April 5, April 14, April 22 (two letters), April 27, May 11, May 20 (two letters), May 27, June 9, June 17, June 20, June 24, July 13 (two letters), July 27, July 29 (two letters), August 3 (three letters), September 12, September 21, September 23, October 13, October 28, and October 31, 2016; and January 20, February 3, March 3, and June 12, 2017.

Brief description of amendments: The amendments revised Renewed Facility Operating Licenses (RFOLs) and Technical Specifications (TSs) to authorize an increase of maximum reactor core thermal power level for Browns Ferry Nuclear Plant, Units 1, 2, and 3 to 3,952 megawatt thermal (MWt). These license amendments represent an increase of approximately 14.3 percent above the current licensed thermal power level of 3,458 MWt, which is an increase of approximately 20 percent above the original licensed thermal power level of 3,293 MWt. The NRC considers the requested increase in power level to be an extended power uprate.

Date of issuance: August 14, 2017.

Effective date: As of the date of issuance and shall be implemented prior to startup from the refueling outages of fall 2018 (Unit 1), spring 2019 (Unit 2), and spring 2018 (Unit 3).

Amendment Nos.: 299 (Unit 1), 323 (Unit 2), and 283 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17032A120; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the RFOLs and TSs.

Date of initial notice in Federal Register: July 5, 2016 (81 FR 43666). The supplemental letters dated April 22 (two letters), April 27, May 11, May 20 (two letters), May 27, June 9, June 17, June 20, June 24, July 13, (two letters); July 27, July 29 (two letters), August 3 (three letters), September 12, September 21, September 23, October 13, October 28, and October 31, 2016; and January 20, February 3, March 3, and June 12, 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 the SE dated August 14, 2017.

No significant hazards consideration comments received: Yes, refer to Section 6.0, “Public Comments,” of the SE.

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

Date of amendment request: September 30, 2016, as supplemented by letters dated March 16 and April 26, 2017.

Brief description of amendment: The amendment revised the emergency action level (EAL) scheme used at WCGS. The currently approved EAL scheme is based on Nuclear Management and Resources Council/National Environmental Studies Project (NUMARC/NESP)-007, Revision 2, “Methodology for Development of Emergency Action Levels,” January 1992. The amendment allows WCNOC to adopt an EAL scheme, which is based on the guidance established in Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” November 2012. Revision 6 of NEI 99-01 has been endorsed by the NRC by letter dated March 28, 2013.

Date of issuance: August 28, 2017.

Effective date: As of its date of issuance and shall be implemented by September 30, 2018.

Amendment No.: 218. A publicly-available version is in ADAMS under Accession No. ML17166A409; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: December 6, 2016 (81 FR 87974). The supplemental letters dated March 16 and April 26, 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 August 28, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 30th day of August 2017.

For the Nuclear Regulatory Commission.

Eric J. Benner,

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

[FR Doc. 2017-19214 Filed 9-11-17; 8:45 am]

BILLING CODE 7590-01-P