ML18295A744

Biweekly Federal Register Notice - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: 11/06/18

ML18295A744
Person / Time
Site:	Peach Bottom, Browns Ferry, Palo Verde, Harris, Point Beach, Oyster Creek, Watts Bar, Byron, Pilgrim, Braidwood, Brunswick, Vogtle, Cook, Comanche Peak, LaSalle
Issue date:	10/25/2018
From:	Craig Erlanger Division of Operating Reactor Licensing
To:
Blechman P
References
NRC-2018-0246
Download: ML18295A744 (43)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2018-0246]

Biweekly Notice Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

SUMMARY

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from October 6, 2018, to October 22, 2018. The last biweekly notice was published on October 23, 2018.

DATES: Comments must be filed by December 6, 2018. A request for a hearing must be filed by January 7, 2019.

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-2018-0246. Address questions about Docket IDs in

Regulations.gov to Jennifer Borges; telephone: 301-287-9127; e-mail:

Jennifer.Borges@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.

• Mail comments to: May Ma, Office of Administration, Mail Stop: TWFN A60M, 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, e-mail: Paula.Blechman@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2018-0246 facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0246.

• NRCs Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select Begin Web-based ADAMS Search. For problems with ADAMS, please contact the NRCs Public Document Room (PDR) reference staff at 1-800-397-4209, 2

301-415-4737, or by e-mail 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.

• NRCs PDR: You may examine and purchase copies of public documents at the NRCs PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments Please include Docket ID NRC-2018-0246 facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.

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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 Commissions regulations in section 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination.

Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any 4

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 Commissions Agency Rules of Practice and Procedure in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRCs regulations are accessible electronically from the NRC Library on the NRCs Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRCs 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 petitioners right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioners 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 petitioners 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 5

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 partys admitted contentions, including the opportunity to present evidence, consistent with the NRCs 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 6

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 petitioners 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 7

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 NRCs 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 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 e-mail 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 8

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 NRCs 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 NRCs 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 NRCs 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 e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRCs 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 NRCs adjudicatory E-Filing system may seek assistance by contacting the NRCs Electronic Filing Help Desk through the Contact Us link located on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals.html, by e-mail to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-9

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 NRCs 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 NRCs 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 10

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 NRCs 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: August 30, 2018. A publicly-available version is in ADAMS under Accession No. ML18242A395.

Description of amendment request: The proposed amendments would add new Required Actions (RAs) and Completion Times (CTs) for three inoperable Control Room air conditioning (AC) subsystems to Technical Specification (TS) 3.7.4, Control Room Air Conditioning (AC) System.

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.

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The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed).

The proposed change adds new RAs and CTs for three inoperable Control Room AC subsystems. The equipment qualification temperature of the control room equipment is not affected. Future changes to the Bases or licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, Changes, tests and experiments, to ensure that such changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated.

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

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

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

Response: No.

The proposed change adds new RAs and CTs for three inoperable Control Room AC subsystems. The change does not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal pant operation. The proposed TSs continue to require maintaining the control room temperature within the design limits.

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

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

Response: No.

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The proposed change adds new RAs and CTs for three inoperable Control Room AC subsystems. Instituting the proposed change will continue to maintain the control room temperature within design limits. Changes to the Bases or licensee-controlled document are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that the control room temperature will be maintained within design limits.

Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Undine Shoop.

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Wake and Chatham Counties, North Carolina Date of amendment request: August 13, 2018. A publicly-available version is in ADAMS under Accession No. ML18226A022.

Description of amendment request: The amendment would revise the Emergency Plan Emergency Action Level (EAL) scheme for HNP associated with the fission product barrier degradation EAL thresholds, and the cold shutdown/refueling system malfunction EAL thresholds.

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:

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

Response: No.

The proposed changes affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not reduce the effectiveness of the HNP Emergency Plan or the HNP Emergency Response Organization. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes do not impact the consequence of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase 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 affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. These changes do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce any accident initiator or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

These changes affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety.

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

Attorney for licensee: David Cummings, Associate General Counsel, Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Undine Shoop.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station (PNPS), Plymouth County, Massachusetts Date of amendment request: August 1, 2018. A publicly-available version is in ADAMS under Accession No. ML18218A184.

Description of amendment request: The amendment would revise the PNPS Emergency Plan and Emergency Action Level (EAL) scheme to support a permanently shutdown and defueled condition at PNPS.

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

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

Response: No.

The proposed changes to the PNPS Emergency Plan and EAL scheme do not impact the function of facility structures, systems, or components. The proposed changes do not affect accident initiators or precursors, nor do they alter design assumptions that could increase the probability or consequences of previously evaluated accidents. The proposed changes do not prevent the ability of the on-shift staff and emergency response organization to perform their intended functions to mitigate the consequences 15

of any accident or event that will be credible in the permanently defueled condition.

The probability of occurrence of previously evaluated accidents is not increased because most previously analyzed accidents can no longer occur and the probability of the few remaining credible accidents are unaffected by the proposed amendment.

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

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

Response: No.

The proposed changes reduce the scope of the PNPS Emergency Plan and EAL scheme commensurate with the hazards associated with a permanently shut down and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment that could create the possibility of a new or different kind of accident. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new or different kinds of accident initiators are created.

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.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the PNPS Emergency Plan and EAL scheme and do not impact operation of the facility or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of facility operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised Emergency Plan will continue to provide the necessary response staff commensurate with the reduction in consequences of radiological events that will be possible at PNPS when the facility 16

is in the permanently defueled condition and therefore, there is no reduction in the margin of safety.

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

The NRC staff has reviewed the licensees 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: Susan H. Raimo, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue, NW, Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Douglas A. Broaddus.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station (PNPS), Plymouth County, Massachusetts Date of amendment request: September 13, 2018. A publicly-available version is in ADAMS under Accession No. ML18260A085.

Description of amendment request: The amendment would revise the Renewed Facility Operating License (RFOL) and the associated Technical Specifications (TSs) to Permanently Defueled Technical Specifications consistent for a facility in a permanently shutdown and defueled condition. The amendment would revise certain requirements contained within the RFOL and TS and remove the requirements that would no longer be applicable upon docketing the certification of permanent fuel removal from the reactor vessel at PNPS. The amendment would also make administrative and editorial changes, such as renumbering of pages, where appropriate, and condense and reduce the number of pages.

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

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

Response: No.

The proposed amendment would not take effect until PNPS has permanently ceased operation, entered a permanently defueled condition, and met the decay requirements established in the analysis of the Fuel Handling Accident (FHA). The proposed amendment would modify the PNPS [RF]OL and TS by deleting the portions of the OL and TS that are no longer applicable to a permanently defueled facility, while modifying the other sections to correspond to the permanently defueled condition. This change is consistent with the criteria set forth in 10 CFR 50.36 for the contents of TS.

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

The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility structures, systems, and components (SSCs) or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled status of PNPS has no impact on the remaining applicable DBAs, i.e., the FHA and the radioactive waste handling accident (High Integrity Container (HIC) Drop Event).

The removal of LCOs [limiting conditions of operations] or SRs

[surveillance requirements] that are related only to the operation of 18

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

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

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

The deletion and modification of provisions of the administrative controls do not directly affect the design of SSCs necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the fuel pool. The changes to the administrative controls do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shut down and defueled condition of the reactor.

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

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

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

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Response: No.

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

The proposed deletion of requirements of the PNPS OL and TS do not affect systems credited in the accident analyses for the FHA or the HIC Drop Event at PNPS. The proposed OL and TS will continue to require proper control and monitoring of safety significant parameters and activities.

The TS regarding SFP water level and fuel storage required is retained to preserve the current requirements for safe storage of irradiated fuel. The restriction on the SFP water level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA.

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

Therefore, the proposed 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.

Because the 10 CFR Part 50 license for PNPS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel after the certifications required by 10 CFR 50.82(a)(1) are docketed for PNPS as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible. The only remaining credible accidents are the FHA and a radioactive waste handling accident (HIC Drop Event). The proposed amendment does not 20

adversely affect the inputs or assumptions of any of the design basis analyses that impact the remaining DBAs.

The proposed changes are limited to those portions of the OL and TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the PNPS OL and TS are not credited in the existing accident analyses for the remaining DBAs; and as such, do not contribute to the margin of safety associated with the accident analyses.

Postulated design basis accidents involving the reactor will no longer be possible because the reactor will be permanently shut down and defueled and PNPS will no longer be authorized to operate the reactor.

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

The NRC staff has reviewed the licensees 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: Susan H. Raimo, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue, NW, Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC, Docket No. STN 50-455, Byron Station, Unit No. 2, Ogle County, Illinois Date of amendment request: March 8, 2018. A publicly-available version is in ADAMS under Accession No. ML18067A431.

Description of amendment request: The amendment would add a License Condition to the Byron Station, Unit No. 2, Renewed Facility Operating License, Appendix C, Additional Conditions, that authorizes use of two lead test assemblies (LTAs) containing a limited number of accident tolerant fuel (ATF) lead test rods (LTRs) during Byron, Unit No. 2, Refueling Cycles 22, 23, and 24.

21

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 involves only a very small number of LTRs, which will be conservatively designed from a neutronic standpoint, and are thermal-hydraulically and mechanically compatible with all plant Systems, Structures and Components (SSCs). The fuel pellets and fuel rods themselves will have no impact on accident initiators or precursors. There will not be a significant impact on the operation of any plant SSC or on the progression of any operational transient or design basis accident. There will be no impact on any procedure or administrative control designed to prevent or mitigate any accident.

The Westinghouse Encore and ADOPT' (with and without chromium-coated cladding) LTAs are of the same design as the co-resident fuel in the core, with the exception of containing a limited number of LTRs in place of the standard fuel rods. The LTAs will be placed in nonlimiting core locations. The Byron Station, Unit 2, [Refueling] Cycle, 22, 23 and 24 reload designs will meet all applicable design criteria. Evaluations of the LTAs will be performed as part of the [refueling] cycle specific reload safety analysis to confirm that the acceptance criteria of the existing safety analyses will continue to be met. Operation of the Westinghouse Encore and ADOPT' fuel will not significantly increase the predicted radiological consequences of accidents currently postulated in the Updated Final Safety Analysis Report.

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

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

Response: No.

The proposed change involves the use of a very small number of LTRs in two LTAs which are very similar in all aspects to the co-resident fuel, as noted in Question 1. The proposed change does not change the design function or operation of any SSC, and does 22

not introduce any new failure mechanism, malfunction, or accident initiator not considered in the current design and licensing bases.

The Byron Station Unit 2 reactor cores will be designed to meet all applicable design and licensing basis criteria. Demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of accident. The reload core designs for the [refueling]

cycles in which the Westinghouse LTAs will operate (i.e.,

[Refueling] Cycles 22, 23 and 24) will demonstrate that the use of the LTAs in nonlimiting core locations is acceptable. The relevant design and performance criteria will continue to be met and no new single failure mechanisms will be created. The use of Westinghouse LTAs does not involve any alteration to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors.

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

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

Response: No.

Operation of Byron Station Unit 2 with two Westinghouse LTAs containing a limited number of LTRs, placed in nonlimiting core locations, does not change the performance requirements on any system or component such that any design criteria will be exceeded. The current limits on core operation defined in the Byron Station Technical Specifications will remain applicable to the subject LTAs during [Refueling] Cycles 22, 23 and 24.

Westinghouse analytical codes and methods will be used, and supplemented as necessary using conservative assumptions, to confirm that all applicable limits associated with the LTAs (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, nuclear limits such as Shutdown Margin, transient analysis limits and accident analysis limits) remain bounded by the current analysis of record.

To further assure no reduction in the margin of safety, the LTRs will be designed with reduced uranium enrichment and will be placed in non-limiting core locations as noted above. With respect to non-fuel SSCs, there is no reduction in the margin of safety for any safety limit, limiting safety system setting, limiting condition of operation, instrument setpoint, or any other design parameter.

Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: September 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18275A023.

Description of amendment request: The amendment would revise the PBAPS, Units 2 and 3, design and licensing basis described in the Updated Final Safety Analysis Report (UFSAR) to reduce the design pressure rating of the High Pressure Service Water (HPSW) system. This change will provide additional corrosion margin in the HPSW system pipe wall thickness, increasing the margin of safety for the existing piping. This one-time change would be implemented starting in the fall of 2019 and would expire for both units on December 31, 2020.

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.

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The HPSW system does not initiate any accidents discussed in Chapter 14 of the PBAPS, Units 2 and 3 UFSAR. A shutdown cooling (RHR [residual heat removal] system) malfunction leading to a moderator temperature decrease could result from mis-operation of the cooling water controls for the RHR heat exchangers, as described in UFSAR Section 14.5.2.4. The resulting temperature decrease causes a slow insertion of positive reactivity into the core. However, the proposed change to the HPSW system design pressure will not affect the initiator for this accident. The proposed reduction of the HPSW system design pressure has been evaluated for effects on system piping and components using appropriate codes and standards. The proposed changes do not introduce any failure mechanisms that would initiate a previously analyzed accident. The HPSW and RHR systems remain capable of performing their UFSAR-described design functions for accident mitigation. Moreover, the design and operability requirements currently addressed by the PBAPS Technical Specifications (TS) are unaffected and the design basis radiological analysis of analyzed accidents is unchanged. Thus, the consequences of analyzed accidents are not increased.

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

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

Response: No.

The proposed changes will reduce the design and operating pressure in a portion of the HPSW system. This change will not introduce a new mode of plant operation. The system flowrate and heat removal rate for design basis events are not changed.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. All accident analysis criteria continue to be met and there are no adverse effects on any safety-related system.

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

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

Response: No.

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The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated and the setpoints for the actuation of equipment relied upon to respond to an event. The reduction in HPSW system design pressure permits continued operation of the HPSW and RHR systems in accordance with the plant safety analysis. The core and containment heat removal functions of the HPSW and RHR systems are not affected. The proposed change does not alter the safety limits or safety analysis assumptions associated with the operation of the plant.

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

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

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

NRC Branch Chief: James G. Danna.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2, Berrien County, Michigan Date of amendment request: September 5, 2018. A publicly-available version is in ADAMS under Accession No. ML18250A185.

Description of amendment request: The proposed change would modify technical specification (TS) Section 5.5.15, Battery Monitoring and Maintenance Program, to align with the latest Institute of Electrical and Electronics Engineers (IEEE) Standard.

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:

26

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

Response: No.

The design of the protection systems will be unaffected. The reactor protection system and engineered safety feature actuation system will continue to function in a manner consistent with the plant design basis. All design, material and construction standards that were applicable prior to the request are maintained.

The proposed amendment will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR).

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 previously evaluated?

Response: No.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of the proposed change. All systems, structures, and components previously required for the mitigation of an event remain capable of fulfilling their intended design function. The proposed change has no adverse effects on any safety related systems or components and does not challenge the performance or integrity of any safety related system. Further, there are no changes in the method by which any safety-related plant system performs its safety function. This amendment will not affect the normal method of power operation or change any operating parameters.

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

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

Response: No.

The margin of safety is established through equipment design, operating parameters; and the setpoints at which automatic actions are initiated. The equipment margins will be maintained in accordance with the plant-specific design bases. The proposed 27

changes will not adversely affect operation of plant equipment.

These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient Direct Current (DC) capacity to support operation of mitigation equipment is ensured.

The changes associated with the Battery Maintenance and Monitoring Program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety-related loads in accordance with analysis assumptions.

The TS changes maintain the same level of equipment performance stated in the UFSAR and the current TSs.

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

The proposed change does not involve a significant reduction in a margin of safety because the proposed changes do not reduce the margin of safety that exists in the present CNP TS or UFSAR.

The operability requirements of the TS are consistent with the initial condition assumptions of the safety analyses.

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

The NRC staff has reviewed the licensees 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: Robert B. Haemer, Senior Nuclear Counsel, Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: David J. Wrona.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: July 30, 2018. A publicly-available version is in ADAMS under Accession No. ML18214A730.

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Description of amendment request: The amendments would revise the requirements on control and shutdown rods, and rod and bank position indication in Technical Specification (TS) 3.1.4, Rod Group Alignment Limits; TS 3.1.5, Shutdown Bank Insertion Limits; TS 3.1.6, Control Bank Insertion Limits; and TS 3.1.7, Rod Position Indication. The changes provide time to repair rod movement failures that do not affect rod operability, provide time for analog position indication instruments to read accurately after rod movement, correct conflicts between the TS, and increase consistency and improve the presentation.

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

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

Response: No.

Control and shutdown rods are assumed to insert into the core to shut down the reactor in evaluated accidents. Rod insertion limits ensure that adequate negative reactivity is available to provide the assumed shutdown margin (SDM). Rod alignment and overlap limits maintain an appropriate power distribution and reactivity insertion profile.

Control and shutdown rods are initiators to several accidents previously evaluated, such as rod ejection. The proposed change does not change the limiting conditions for operation for the rods or make any technical changes to the Surveillance Requirements (SRs) governing the rods. Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated.

Revising the TS Actions to provide a limited time to repair rod movement control has no effect on the SDM assumed in the accident analysis as the proposed Action require verification that SDM is maintained. The effects on power distribution will not cause a significant increase in the consequences of any accident previously evaluated as all TS requirements on power distribution continue to be applicable.

29

Therefore, the assumptions used in any accidents previously evaluated are unchanged and there is no significant increase in the consequences.

The consequences of an accident that might occur during the one-hour period provided for the analog rod position indication to stabilize after rod movement are no different from the consequences of the accident under the existing actions with the rod declared inoperable.

The proposed change to resolve the conflicts in the TS ensure that the intended Actions are followed when equipment is inoperable. Actions taken with inoperable equipment are not assumptions in the accidents previously evaluated and have no significant effect on the consequences.

The proposed change to increase consistency within the TS has no effect on the consequences of accidents previously evaluated as the proposed change clarifies the application of the existing requirements and does not change the intent.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed).

The change does not alter assumptions made in the safety analyses. The proposed change does not alter the limiting conditions for operation for the rods or make any technical changes to the SRs governing the rods. The proposed change to actions maintains or improves safety when equipment is inoperable and does not introduce new failure modes.

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 to allow time for rod position indication to stabilize after rod movement and to allow an alternative method of verifying rod position has no effect on the safety margin, as actual 30

rod position is not affected. The proposed change to provide time to repair rods that are operable but immovable does not result in a significant reduction in the margin of safety because all rods must be verified to be operable, and all other banks must be within the insertion limits. The remaining proposed changes to make the requirements internally consistent and to eliminate unnecessary actions do not affect the margin of safety as the changes do not affect the ability of the rods to perform their specified safety function.

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

The NRC staff has reviewed the licensees 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: Debbie Hendell, Managing Attorney - Nuclear, Florida Power &

Light Company, Mail Stop: LAW/JB, 700 Universe Boulevard, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee Date of amendment request: July 23, 2018. A publicly available version is in ADAMS under Accession No. ML18205A492.

Description of amendment request: The amendments would revise the Units 1 and 2 Technical Specification (TS) 4.2.1, Fuel Assemblies, to allow the use of Optimized ZIRLO' fuel rod cladding material. They would also revises Units 1 and 2 TS 5.9.5, Core Operating Limits Report (COLR), to add Westinghouse Electric Company Topical Reports WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO',

to the list of analytical methods used to determine the core operating limits approved by 31

the NRC. In addition, the amendments would correct the spelling of the word Zircaloy in WBN Unit 1 TS 4.2.1 only, add the word clad after the proposed phrase Optimized ZIRLO', capitalize the word Zirlo, and add a registered trademark designator to the word ZIRLO.

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

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

Response: No.

The proposed amendment will allow the use of Optimized ZIRLO clad nuclear fuel at WBN Units 1 and 2. The NRC approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, which addresses Optimized ZIRLO fuel rod cladding and demonstrates that Optimized ZIRLO fuel rod cladding has essentially the same properties as currently licensed ZIRLO fuel rod cladding. The use of Optimized ZIRLO fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO meets the fuel design acceptance criteria and hence does not significantly affect the consequences of an accident.

Therefore, the proposed TS change does not result in a significant increase in the probability or consequences of an accident previously evaluated within the WBN [Unit 1 and] Unit 2 UFSAR

[Updated Final Safety Analysis Report].

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

Response: No.

The use of Optimized ZIRLO fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A demonstrated that the material properties of Optimized ZIRLO fuel rod cladding are similar to those of ZIRLO fuel rod cladding.

Therefore, Optimized ZIRLO fuel rod cladding will perform 32

similarly to ZIRLO fuel rod cladding, thus precluding the possibility of the fuel rod cladding becoming an accident initiator and causing a new or different kind of accident.

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

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

Response: No.

WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of the Optimized ZIRLO fuel rod cladding are similar to those of ZIRLO fuel rod cladding.

Optimized ZIRLO fuel rod cladding is expected to perform similarly to ZIRLO fuel rod cladding for normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The use of Optimized ZIRLO fuel rod cladding will not result in adverse changes to the operation or configuration of the facility.

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

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

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

NRC Branch Chief: Undine Shoop.

33

III. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices.

The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Vistra Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant (CPNPP), Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: September 5, 2018, as supplemented by letters dated September 20 and October 3, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML18250A186, ML18267A059, and ML18277A207, respectively.

Brief description of amendment request: The amendments would revise the CPNPP Technical Specification 3.8.4, DC [Direct Current] Sources - Operating, by adding a new REQUIRED ACTION to CONDITION B and an extended COMPLETION TIME, on a one-time basis to repair two affected battery cells on the CPNPP Unit 1, Train B safety-related batteries.

Date of publication of individual notice in Federal Register: October 10, 2018 (83 FR 50971).

34

Expiration date of individual notice: October 24, 2018 (public comments); December 10, 2018 (hearing requests).

IV. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations. The Commission has made appropriate findings as required by the Act and the Commissions 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 Commissions related letter, Safety 35

Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the Obtaining Information and Submitting Comments section of this document.

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

Maricopa County, Arizona Date of amendment request: July 19, 2017, as supplemented by letters dated May 9, July 13, and August 10, 2018.

Brief description of amendments: The amendments modified the licensing basis by the addition of a license condition to allow the implementation of the provisions of 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, for Palo Verde. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced.

Date of issuance: October 10, 2018.

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

Amendment Nos.: 207 (Unit 1), 207 (Unit 2), and 207 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML18243A280; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

36

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

Date of initial notice in Federal Register: September 26, 2017 (82 FR 44850). The supplements dated May 9, July 13, and August 10, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated October 10, 2018.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Date of amendment request: September 1, 2017, as supplemented by letters dated April 4, 2018, June 13, 2018, and September 13, 2018.

Brief description of amendments: The amendments revised the licensing basis by the addition of a license condition to allow for the implementation of the provisions of 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment 37

determined to be of high safety significance, requirements will not be changed or will be enhanced.

Date of issuance: October 22, 2018.

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

Amendment Nos: Braidwood - 198/198 and Byron - 204/204. A publicly-available version is in ADAMS under Accession No. ML18264A092; documents related to these amendments are listed in the related Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: November 21, 2017 (82 FR 55404).

The supplements dated April 4, 2018, June 13, 2018, and September 13, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated October 22, 2018.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of amendment request: December 13, 2017, as supplemented by letter dated June 18, 2018.

38

Brief description of amendments: The amendments revised the LSCS, Units 1 and 2, Technical Specifications to adopt Technical Specifications Task Force (TSTF)-542, Reactor Pressure Vessel Water Inventory Control.

Date of issuance: October 15, 2018.

Effective date: As of the date of issuance and shall be implemented for LSCS, Units 1 and 2 prior to initial entry into Mode 4 during the LSCS Unit 2 refueling outage in 2019 (i.e., L2R17), which is currently scheduled to occur in February 2019.

Amendment Nos.: 230 (Unit 1) and 216 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18226A202; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: February 13, 2018 (83 FR 6223). The supplemental letter dated June 18, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated October 15, 2018.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Date amendment request: August 29, 2017, as supplemented by letter dated February 13, 2018.

39

Brief description of amendment: The amendment revised the site emergency plan and emergency action level scheme for the permanently shutdown and defueled condition.

Date of issuance: October 17, 2018.

Effective date: The amendment is effective 12 months (365 days) following the permanent cessation of power operations and shall be implemented within 60 days of the effective date, but no later than March 28, 2021.

Amendment No.: 294. A publicly-available version is in ADAMS under Accession No. ML18221A400; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-16: Amendment revised the emergency plan and emergency action level scheme.

Date of initial notice in Federal Register: October 24, 2017 (82 FR 49238). The supplemental letter dated February 13, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated October 17, 2018.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: July 28, 2017, as supplemented by letters dated January 23, March 23, June 21, and August 9, 2018.

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Description of amendment: The amendment authorized the Southern Nuclear Operating Company to change the VEGP Units 3 and 4 plant-specific Combined License (COL)

Appendix A, Technical Specifications (TS) as incorporated into the VEGP Units 3 and 4 COLs. The amendment consisted of changes to the COL Appendix A TS related to reactivity controls and other miscellaneous changes. The amendment revised the COL Appendix A, plant-specific TS by modifying the TS to make them consistent with the design, licensing basis, and other related TS.

Date of issuance: August 23, 2018.

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

Amendment Nos.: 138 (Unit 3) and 137 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML18100A110; documents related to the amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: December 5, 2017 (82 FR 57469). The supplemental letters dated January 23, March 23, June 21 and August 9, 2018 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazard determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in the Safety Evaluation dated August 23, 2018.

No significant hazards consideration comments received: No.

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Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant (Browns Ferry), Units 1, 2, and 3, Limestone County, Alabama Date of amendment request: May 3, 2018.

Brief description of amendment: The amendments revised the Browns Ferry, Units 1, 2, and 3, Renewed Facility Operating Licenses to provide a correction to previously submitted information in relation to their approved fire protection program under 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. Specifically, the amendments modified the Browns Ferry licenses to reflect changes to Item 3.3.4 in Table B-1, Transition of Fundamental Fire Protection Program & Design Elements, of Attachment A in the NRC-approved amendments regarding NFPA 805 dated March 27, 2013.

Date of issuance: October 9, 2018.

Effective date: As of the date of issuance and shall be implemented immediately.

Amendment Nos.: 306 (Unit 1); 329 (Unit 2); and 289 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML18241A319; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: The amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: July 17, 2018 (83 FR 33270).

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The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated October 9, 2018.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 25th day of October, 2018.

For the Nuclear Regulatory Commission.

/RA/

Craig G. Erlanger, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

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