2018-04827. 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 February 13 to February 26, 2018. The last biweekly notice was published on February 27, 2018.

DATES:

Comments must be filed by April 12, 2018. A request for a hearing must be filed by May 14, 2018.

ADDRESSES:

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

• Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0045. Address questions about NRC dockets to Jennifer Borges; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: May Ma, Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2018-0045, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0045.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2018-0045, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

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

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: September 14, 2017. A publicly-available version is in ADAMS under Accession No. ML17261B255.

Description of amendment request: The amendments would modify Catawba Nuclear Station, Units 1 and 2, Technical Specification (TS) 3.7.8, “Nuclear Service Water System (NSWS).” Specifically, the proposed change would add a new Condition D for one NSWS pond return header being inoperable due to the NSWS being aligned for single pond return header operation with a Completion Time (CT) of 30 days. This would involve isolating one train of the NSWS pond return piping at the Auxiliary Building wall and maintaining the discharge crossover lines open between trains in the Auxiliary Building and Emergency Diesel Generator Buildings. This provides a common safety-related discharge path through the single remaining in-service pond return line. This alignment, single pond return header operation, allows a pond return header to be removed from service while a flow path is maintained through both trains of NSWS supplied equipment to the Standby Nuclear Service Water Pond (SNSWP).

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 single pond return header operation configuration for NSWS operation and the associated proposed TS and TS Bases changes have been evaluated to assess their impact on plant operation and to ensure that the design basis safety functions of safety related systems are not adversely impacted. During single pond return header operation, the operating NSWS header will be able to discharge all required NSWS flow from safety related components. PRA [Probabilistic Risk Assessment] has demonstrated that due to the limited proposed time in the single pond return header configuration, the resultant plant risk remains acceptable.

The purpose of this amendment request is to ultimately facilitate inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the Discharge to the SNSWP. Therefore, NRC approval of this request will ultimately help to enhance the long-term structural integrity of the NSWS and will help to ensure the system's reliability for many years.

In general, the NSWS serves as an accident mitigation system and cannot by itself initiate an accident or transient situation. The only exception is that the NSWS piping can serve as a source of floodwater to safety related equipment in the Auxiliary Building or in the diesel generator buildings in the event of a leak or a break in the system piping. The probability of such an event is not significantly increased as a result of this proposed request. Safety related NSWS piping is tested and inspected in accordance with all applicable in-service testing and in-service inspection requirements. Given the negligible influence of flooding events on the NSWS for the submittal configuration (i.e., no dominant contribution from floods), it is judged that the analyses assessing the influence of these events provide an acceptable evaluation of the contribution of the flood risk for the requested CT of 30 days. The proposed 30 day TS Required Action CT has been evaluated for risk significance and the results of this evaluation have been found acceptable. The probabilities of occurrence of accidents presented in the UFSAR [Updated Final Safety Analysis Report] will not increase as a result of implementation of this change. Because the PRA analysis supporting the proposed change yielded acceptable results, the NSWS will maintain its required availability in response to accident situations. Since NSWS availability is maintained, the response of the plant to accident situations will remain acceptable and the consequences of accidents presented in the UFSAR will not increase.

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.

Implementation of this amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed request does not affect the basic operation of the NSWS or any of the systems that it supports. These include the Emergency Core Cooling System, the Containment Spray System, the Containment Valve Injection Water System, the Auxiliary Feedwater System, the Component Cooling Water System, the Control Room Area Ventilation System, the Control Room Area Chilled Water System, the Auxiliary Building Filtered Ventilation Exhaust System, or the Diesel Generators. During proposed single pond return header operation, the NSWS will remain capable of fulfilling all of its design basis requirements.

No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant, which will introduce any new type of accident outside those assumed in the UFSAR.

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 the margin of safety?

Response: No.

Implementation of this amendment will not involve a significant reduction in any margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed TS amendment. During single pond return header operation, the NSWS and its supported systems will remain capable of performing their required functions. No safety margins will be impacted.

The PRA analysis conducted for this proposed amendment demonstrated that the impact on overall plant risk remains acceptable during single pond return header operation. Therefore, there is not a significant reduction in the margin of safety.

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

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

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

Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke Energy Carolinas, LLC, 550 South Tryon Street—DEC45A, Charlotte, NC 28202-1802.

NRC Branch Chief: Michael T. Markley.

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: October 3, 2017. A publicly-available version is in ADAMS under Accession No. ML17277A855.

Description of amendment request: The amendments would revise Surveillance Requirement (SR) 3.8.4.5 contained in Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating.”

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

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

Response: No.

The proposed change to the battery charger amperage requirements of SR 3.8.4.5 contained in TS 3.8.4 does not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SSCs perform their design function. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The DC electrical power system, including the battery chargers, is not an initiator of any accident sequence analyzed in the Updated Final Safety Analysis Report. Rather, the DC electrical power system supports operation of equipment used to mitigate accidents. Specifically, the purpose of the battery chargers is to continuously maintain their respective battery in a charged standby condition while providing power to the system loads. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained.

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 to the battery charger amperage requirements of SR 3.8.4.5 contained in TS 3.8.4 does not require any modification to the plant or change equipment operation. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. Performance of battery testing is not a precursor to any accident previously evaluated. The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators.

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

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

Response: No.

The proposed change to the battery charger amperage requirements of SR 3.8.4.5 contained in TS 3.8.4 does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, Appendix A will continue to be met.

Therefore, the proposed change does not involve any 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.

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

Date of amendment request: December 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17346B280.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) 3.6.4.1, “Secondary Containment,” Surveillance Requirement (SR) 3.6.4.1.1. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, “Revise Secondary Containment Surveillance Requirements” (ADAMS Accession No. ML16277A226).

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 addresses conditions during which the secondary containment SRs are not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four hour Completion Time for an inoperable secondary containment. In addition, the proposed Note for SR 3.6.4.1.1 provides an alternative means to ensure the secondary containment safety function is met. As a result, the consequences of an accident previously evaluated are not significantly increased.

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

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

Response: No.

The proposed change does not alter the protection system design, create new failure modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in 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 change involve a significant reduction in a margin of safety?

Response: No.

The proposed change addresses conditions during which the secondary containment SRs are not met. Conditions in which the secondary containment vacuum is less than the required vacuum are acceptable provided the conditions do not affect the ability of the SGT [Standby Gas Treatment] System to establish the required secondary containment vacuum under post-accident conditions within the time assumed in the accident analysis. This condition is incorporated in the proposed change by requiring an analysis of actual environmental and secondary containment pressure conditions to confirm the capability of the SGT System is maintained within the assumptions of the accident analysis. Therefore, the safety function of the secondary containment is not affected. The allowance for both an inner and outer secondary containment door to be open simultaneously for entry and exit does not affect the safety function of the secondary containment as the doors are promptly closed after entry or exit, thereby restoring the secondary containment boundary.

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

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

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

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: December 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17352B255.

Description of amendment request: The amendment would revise the Environmental Protection Plan to incorporate the terms and conditions of the Incidental Take Statement included in the Biological Opinion issued to Energy Northwest by the National Marine Fisheries Service.

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 changes are administrative in nature and would in no way affect the initial conditions, assumptions, or conclusions of Columbia's accident analyses. In addition, the proposed changes would not affect the operation or performance of any equipment assumed in the accident analyses.

Therefore there is no 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 analyzed?

Response: No.

The changes are administrative in nature and would in no way impact or alter the configuration or operation of the facility and would create no new modes of operation.

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 changes are administrative in nature and would in no way affect plant or equipment operation or the accident analysis.

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

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

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

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: December 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17354A007.

Description of amendment request: The amendment would revise Technical Specification (TS) Limiting Condition for Operation 3.7.13, “Spent Fuel Pit Storage,” and TS 4.3, “Fuel Storage.” Specifically, the proposed changes would (1) resolve a nonconservative TS associated with TS Limiting Condition for Operation 3.7.13, (2) negate the need for the associated compensatory measures, and (3) remove credit for the installed Boraflex panels as a neutron absorber in the criticality analysis of record. The proposed changes in the criticality analysis of record would instead credit empty cells, rod cluster control assemblies (RCCAs), and neutron leakage along the outer two storage rows of the spent fuel pit (SFP).

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 was evaluated for impact on the following previously evaluated events and accidents:

—Multiple Misloads

—Misplaced Assembly

—Dropped Assembly

—Misloaded Assembly

—Over Temperature

—Seismic

—Boron Dilution

—Fuel Handling Accident

—Loss of Spent Fuel Pool Cooling

Multiple Misloads, Misplaced Assembly, Dropped Assembly, Misloaded Accidents

Operation in accordance with the proposed Technical Specifications will not significantly increase the probability of multiple misloads, misplaced assembly, dropped assembly and misloaded assembly accidents because:

a. There are no changes to the equipment for fuel handling or how fuel assemblies are handled, including how fuel assemblies are inserted into and removed from SFP storage locations. There are no changes to how RCCAs will be handled, including how RCCAs are inserted into, or removed from, a fuel assembly.

b. The processes and procedures that are currently in place are sufficiently robust. The proposed Technical Specifications utilize the same basic fuel assembly classification and storage location concepts as those currently in place. However, they do represent a minimal increase in complexity:

—The current TS for fuel storage are complex because the Boraflex neutron absorber built into the SFP racks has degraded. To address this degradation the SFP is divided into four irregularly shaped Regions (Region 1-1, Region 1-2, Region 2-1, and Region 2-2). In addition to the four regions there are six special locations known as peripheral locations in Region 2-2 which are treated as suitable for storage of fuel otherwise designated for Region 1-1 or 1-2. These regions are graphically depicted in the current TS Figure 3.7.13-5.

Each one of these regions has its own rules for fuel placement which are identified in the TS.

—The current Technical Specifications determine a minimum required burnup for each fuel assembly based on initial enrichment, burnup, and cooling time with individual fuel assembly storage location within the SFP restricted based on this minimum required burnup. The minimum required burnup is determined for each of the four regions (1-1, 1-2, 2-1, and 2-2) that utilize a total of ten curves. The proposed assembly categorization is slightly more complex due to the following:

○ The minimum required burnup is dependent on the averaged assembly peaking factor in addition to the initial enrichment, burnup, and cooling time.

○ the minimum required burnup is used to determine the reactivity category of each fuel assembly.

○ the minimum required burnup is adjusted, as necessary, to account for hafnium inserts, a reconstituted fuel assembly with missing stainless steel replacement rods, and a maximum burnup average boron concentration in excess of 950 ppm [parts per million].

—The current Technical Specifications restrict acceptable SFP storage locations to Regions 1-1, 1-2, 2-1 and 2-2 based on minimum required burnup. The proposed Technical Specifications are minimally more complex due to the following:

○ Acceptable storage locations are defined by fuel assembly category and a base configuration is specified. There are five reactivity categories. Certain cell locations in Region 2 require that Category 5 fuel assemblies contain a full length RCCA.

○ the base configurations in Region 1 and Region 2 may be changed in accordance with certain well-defined criteria. An example of a change to a base configuration is that a checkerboard area may be formed in Region 2 where all four sides of the checkerboard are rows of empty cells.

The minimal increase in complexity of current and future fuel categorization and SFP storage restrictions is offset by the significant number of fuel assemblies that have been pre-categorized in TS Tables 3.7.13-2 and Table 3.7.13-3. The minimal increase is also offset by the use of two curves to determine the minimum required burnup (instead of the 10 currently used).

Operation in accordance with the proposed TS will not significantly increase the consequences of multiple misloads, misplaced assembly, dropped assembly and misloaded assembly criticality accidents because the proposed CSA [criticality safety analysis] demonstrates that the acceptance criteria continue to be met for each of these accidents.

Over Temperature Accident

Operation in accordance with the proposed TS will not significantly increase the probability of an over temperature accident because the proposed change does not alter the manner in which the IP2 spent fuel cooling loop is designed, operated, or maintained.

Operation in accordance with the proposed TS will not significantly increase the consequences of an over temperature accident because the proposed CSA demonstrates that the acceptance criteria continue to be met for this accident.

Seismic Event

Operation in accordance with the proposed TS will not significantly increase the probability of a seismic event because there are no elements of the proposed changes that influence the occurrence of any natural event.

Operation in accordance with the proposed TS will not significantly increase the consequences of a seismic event because the proposed changes do not significantly alter the physical arrangement of the spent fuel racks and do not increase the allowable number of fuel assemblies to be stored in the pool. The proposed TS changes require two cell blockers to be in place. These cell blockers have been evaluated and they have a negligible effect on the seismic response of the SFP racks. In addition, the proposed TS changes allow for the placement of miscellaneous non-actinide materials, for example, empty or full trash baskets in fuel positions of any category, in Water Holes and in 50% Water Holes. The placement of miscellaneous materials in the identified locations has been evaluated and has a negligible effect on the seismic response of the SFP racks.

Boron Dilution Accident

Operation in accordance with the proposed TS will not significantly increase the probability of a boron dilution event because the proposed change does not alter the manner in which the IP2 spent fuel cooling system or any other plant system is designed, operated, or maintained, or otherwise increase the likelihood of adding significant quantities of unborated water into the spent fuel pit.

Operation in accordance with the proposed TS will not significantly increase the consequences of a boron dilution event because the TS minimum soluble boron concentration remains unchanged at 2000 ppm and the boron concentration required to ensure k_eff less than or equal to 0.95 has been evaluated at 700 ppm. The proposed CSA demonstrates that the acceptance criteria continue to be met for this accident.

Fuel Handling Accident

Operation in accordance with the proposed TS will not significantly increase the probability of a[n] FHA [fuel handling accident] because the individual fuel assemblies will be moved using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

Operation in accordance with the proposed TS will not significantly increase the consequences of a[n] FHA because the radiological source term of a single fuel assembly will remain the same.

Loss of Spent Fuel Pool Cooling

Operation in accordance with the proposed TS will not significantly increase the probability of a loss of spent fuel pit cooling because the proposed change does not alter the manner in which the IP2 spent fuel cooling loop is designed, operated, or maintained.

Operation in accordance with the proposed TS will not significantly increase the consequences of a loss of spent fuel pit cooling because the proposed change credits empty cells whereas the thermal design basis for the spent fuel pit cooling loop provides for all fuel pit rack locations to be filled at the end of a full core discharge. The proposed TS changes require two cell blockers to be in place. These cell blockers have been evaluated and they have a negligible effect on the thermal response to a loss of spent fuel pool cooling. In addition, the proposed TS changes allow for the placement of miscellaneous non-actinide materials, for example, empty or full trash baskets in fuel positions of any category, in Water Holes and in 50% Water Holes. The placement of miscellaneous materials in the identified locations has been evaluated and has a negligible effect on the thermal response to a loss of spent fuel pool cooling.

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.

Operation in accordance with the proposed TS do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents.

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.

Operation in accordance with the proposed TS does not involve a significant reduction in a margin of safety.

The margin of safety required by 10 CFR 50.68(b)(4) remains unchanged. The evaluations in the CSA confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins for both normal operations and accident conditions. In addition, the SFP seismic and thermal margins are essentially unchanged.

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: Bill Glew, Associate General Counsel, Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA 70113.

NRC Branch Chief: James G. Danna.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3 (IP3), Westchester County, New York

Date of amendment request: December 8, 2017. A publicly-available version is in ADAMS under Accession No. ML17349A131.

Description of amendment request: The amendment would allow for a one-time extension to the 15-year frequency of the IP3 containment leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) or Type A test). Specifically, Technical Specification 5.5.15, “Containment Leakage Rate Testing Program,” would be revised to allow the existing ILRT frequency to be extended one time from 15 to 16 years. The next required ILRT test would be performed no later than the plant restart after the spring 2021 (3R21) refueling outage.

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 involves changes to the IP3 containment leakage rate testing program. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed amendment adopts the NRC accepted guidelines of NEI [Nuclear Energy Institute] 94-01, Revision 3-A, for development of the IP3 performance-based testing program for the Type A testing. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components would limit leakage rates to less than the values assumed in the plant safety analyses. The potential consequences of extending the ILRT interval one-time to 16 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be acceptably small and determined to be within the guidelines published in the NRC Final Safety Evaluation for NEI Topical Report (TR) 94-01, Revision 3-A. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. Entergy has determined that the increase in conditional containment failure probability due to the proposed change would be very small. Therefore, it is concluded that the proposed amendment does not significantly increase the consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for the establishment of a one-time only 16-year interval for the performance of the containment ILRT. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. 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 adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for the establishment of a one-time only 16-year interval for the performance of the containment ILRT. This amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, Type B, and Type C containment leakage tests would be performed at the frequencies established in accordance with the NRC accepted guidelines of NEI 94-01, Revision 3-A. Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment would not degrade in a manner that is not detectable by an ILRT. A risk assessment using the current IP3 PSA [probabilistic safety analysis] model concluded that extending the ILRT test interval one-time from 15 years to 16 years results in a very small change to the risk profile. 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: Bill Glew, Associate General Counsel, Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA 70113.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: January 9, 2017. A publicly-available version is in ADAMS under Accession No. ML18009B037.

Description of amendment request: The proposed change would incorporate a revised alternative source term dose calculation resulting from the removal of a reduction factor credit for dual remote Control Room outside air intakes that had been previously misapplied. This would modify the loss-of-coolant accident (LOCA) dose calculation and the subsequent calculation results as described in the CPS Updated Safety Analysis Report and would revise the affected CPS Technical Specifications.

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

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

Response: No.

The proposed change results in higher Control Room X/Qs [atmospheric dispersion values] which are equivalent to reduced atmospheric dispersion. The increased Control Room X/Qs, in turn, result in higher post-accident Control Room doses. Neither the higher X/Qs, nor the resultant increase in the Control Room doses affect any initiator or precursor of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed change results in an increase in the post-LOCA radiological dose to a Control Room occupant. However, the resultant post-LOCA Control Room dose remains within the regulatory limits of 10 CFR 50.67 [, “Accident source term”] and 10 CFR 50, Appendix A, “General Design Criteria for Nuclear Power Plants” Criterion 19, “Control Room.” Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

In summary, 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 does not alter the design function of operation of the Control Room heating, ventilation, and air-conditioning (HVAC) system, or the ability of this system to perform its design function. The only change is the removal of the Control Room dose reduction factor credit taken for providing a dual remote Control Room air intake. The proposed change does not alter the safety limits, or safety analysis associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators. Rather, this proposed change is the result of an evaluation of the Control Room doses following the most limiting LOCA that can occur at CPS. The proposed change does not introduce any new modes of plant operation. As a result, no new failure modes are introduced.

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

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

Response: No.

The revised post-LOCA dose consequences to a Control Room occupant were calculated in accordance with the requirements of 10 CFR 50.67, [Regulatory Guide (RG)] 1.183, [“Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors”] and NRC SRP [Standard Review Plan] Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.”

The margin of safety is considered to be that provided by meeting the applicable regulatory limits. The increased Control Room X/Qs result in an increase in Control Room dose following the design basis LOCA; however, since the Control Room dose following the design basis accident remains within the regulatory limits, there is not a significant reduction in a margin of safety.

Therefore, operation of CPS in accordance with the proposed change will 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. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: January 24, 2018. A publicly-available version is in ADAMS under Accession No. ML18024A275.

Description of amendment request: The proposed amendments would revise Technical Specifications (TSs) 3.7.2, “Diesel Generator Cooling Water (DGCW) System”; 3.8.1, “AC [Alternating Current] Sources-Operating”; and the associated TS Bases to allow an extended period to install isolation valves to support replacing degraded Core Standby Cooling System (CSCS) piping.

The proposed changes modify TS 3.7.2 to include a 7-day Completion Time (CT) when one or more required DGCW subsystem(s) are inoperable. The proposed changes to TS 3.8.1 include a 7-day CT when a Division 2 Diesel Generator (DG) and the required opposite unit Division 2 DG are inoperable. The proposed changes will only be used during four refueling outages, two for Unit 1 prior to July 1, 2024, and two for Unit 2 prior to July 1, 2023. The current planned schedule, subject to change, is L2R17 (2019), L1R18 (2020), L2R18 (2021), and L1R19 (2022).

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

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

Response: No.

The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. No active or passive failure mechanisms that could lead to an accident are affected. Non-Code line stops required to provide isolation will maintain the availability of the online unit's CSCS. The non-Code line stops being used to isolate the system during the specified refueling outages are being designed to the same or greater pressure rating and seismic requirements as the CSCS piping.

Redundancy is provided by designing the CSCS as multiple independent subsystems. Divisional separation between subsystems assures that no single failure can affect more than one division's subsystem. Therefore, assuming a single failure in any division's subsystem including the subsystem shared between units, two other divisional subsystems in each unit will remain unaffected. This ensures adequate redundancy to supply the minimum required cooling water for safe shutdown of the operating unit or mitigate the consequences of an accident.

The proposed limited use of increased CT's of the operating unit's CSCS maintains the design basis assumptions. 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 involves the temporary installation of new equipment (mechanical line stops) that will be designed and installed to the same or greater pressure rating and seismic design as the CSCS piping. The currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases.

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 does not alter any existing setpoints at which protective actions are initiated and no new setpoints or protective actions are introduced. The design and operation of the CSCS remains unchanged. The proposed change provides a limited period to restore inoperable DGCW subsystems and Division 2 DGs instead of interrupting plant operations, possibly requiring an orderly plant shutdown of the operating unit. The potential to avoid a plant transient in conjunction with maintaining availability of the DGCW subsystems and Division 2 DGs offsets any risk associated with the limited CT. The proposed change does not impact a design basis, limiting safety system setting, or safety limit specified in TSs.

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

NRC Branch Chief: David J. Wrona.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa

Date of amendment request: December 15, 2017. A publicly-available version is in ADAMS under Accession No. ML17363A067.

Description of amendment request: The proposed amendment would revise the Emergency Plan for the DAEC to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” which has been endorsed by the NRC.

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 does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. No actual facility equipment or accident analyses are affected by the proposed changes.

The change revises the NextEra Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, “Methodology for Development of Emergency Action Levels,” but does not alter any of the requirements of the Operating License or the Technical Specifications.

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

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect [sic] on the availability, operability, or performance of safety-related systems and components. The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis.

The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses.

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

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

Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: December 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17353A928.

Description of amendment request: The proposed amendment would separate the Linear Heat Generation Rate (LHGR) requirements and actions from the Average Planar Linear Heat Generation Rate (APLHGR) requirements and actions contained in Technical Specification (TS) 3.2.1. The proposed amendment adds new TS 3.2.3, “Linear Heat Generation Rate (LHGR),” and modifies TS 1.1, “Definitions,” TS 3.4.1, “Recirculation Loops Operating,” and TS 5.6.5, “Core Operating Limits Report (COLR),” to reflect the LHGR change. Modifications associated with TS 3.2.1 and the new TS 3.2.3 are also being added to the actions for TS 3.3.4.1, “End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,” and TS 3.7.7, “The Main Turbine Bypass 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.

The separation of the LHGR requirements and actions from the APLHGR TS is an administrative change. No actions within the TS are changed. The addition of the LCO [limiting condition for operation] for APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO for TS 3.7.7, Main Turbine Bypass System reflect within the TS requirements APLHGR and LHGR actions which are already occurring via the core monitoring processes in place. None of those changes affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner.

LHGR limits have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time.

The proposed change does not modify the limits, change assumptions for the accident analysis, or change operation of the station. Therefore, the proposed change does not involve an increase in the probability or consequences of a previously evaluated accident.

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

Response: No.

The separation of the LHGR requirements and actions from the APLHGR TS is an administrative change. No actions within the TS are changed. The addition of the LCO for APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO for TS 3.7.7, Main Turbine Bypass System reflect within the TS requirements APLHGR and LHGR actions which are already occurring via the core monitoring processes in place. None of those changes affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner.

The proposed change does not modify the limits, change assumptions for the accident analysis, or change operation of the station. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The margin of safety is not affected by the separation of the LHGR requirements and actions from the APLHGR TS. Similarly, the margin of safety is not affected by the addition of the LCO for APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO for TS 3.7.7, Main Turbine Bypass System.

Appropriate measures exist to control the values of these limits since it is required by TS that only NRC-approved methods be used to determine the limits. The proposed change continues to require operation within the core thermal limits as obtained from NRC-approved reload design methodologies and the actions to be taken if a limit is exceeded remain unchanged, again, in accordance with existing TS.

The proposed change does not modify the limits, change assumptions for the accident analysis, or change operation of the station. Therefore, the proposed change has no impact to the margin of safety.

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

Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: December 18, 2017, as supplemented by letter dated February 9, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML17352A502 and ML18040A319, respectively.

Description of amendment request: The amendments would revise Technical Specification (TS) 3/4.3.1, “Reactor Trip System Instrumentation,” and TS 3/4.3.2, “Engineered Safety Feature Actuation System Instrumentation,” to increase the completion times and bypass test times at Salem Nuclear Generating Station, Unit Nos. 1 and 2. The proposed changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Travelers TSTF-411, Revision 1, “Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P),” and TSTF-418, Revision 2, “RPS [Reactor Protection System] and ESFAS [Engineered Safety Feature Actuation System] Test Times and Completion Times (WCAP-14333),” or are supported by plant-specific analysis.

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

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

Response: No.

The proposed changes to the completion times and bypass test time reduce the potential for inadvertent reactor trips and spurious actuations, and therefore do not increase the probability of any accident previously evaluated. The proposed changes to the completion times and bypass test time do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the reactor trip system and engineered safety feature actuation system (RTS and ESFAS) signals. The RTS and ESFAS will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by core damage frequency (CDF) is less than 1.0E-06 per year and the impact on large early release frequency (LERF) is less than 1.0E-07 per year. In addition, for the completion time change, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5.0E-7 and 5.0E-08, respectively. These changes meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and ESFAS will continue to perform their functions with high reliability as originally assumed, and the increase in risk as measured by CDF, LERF, ICCDP, ICLERP is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes are consistent with safety analysis assumptions and resultant consequences.

Therefore, this change does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not result in a change in the manner in which the RTS and ESFAS provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. There are no design changes associated with the license amendment. The changes to completion times and bypass test time do not change any existing accident scenarios, nor create any new or different accident scenarios.

The proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism.

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

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in Regulatory Guides 1.174 and 1.177.

Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: James G. Danna.

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: December 20, 2017. A publicly-available version is in ADAMS under Accession No. ML17354A964.

Description of amendment request: The requested amendment proposes changes to Combined License Appendix C (and to plant-specific Tier 1 information) and associated Tier 2 information to allow a pneumatic test to be used in lieu of a hydrostatic test for the Main Control Room Emergency Habitability System (VES) consistent with American Society of Mechanical Engineers (ASME) Section III.

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 allow for pneumatic testing of the VES ASME Section III components and piping. ASME Section III, ND-6000 contains the requirements for pressure testing of piping and components. ASME Section III, ND-6112.1(a) allows for a pneumatic test to be used in lieu of a hydrostatic test when components, appurtenances or systems cannot be readily dried and traces of the testing medium cannot be tolerated. Due to the design and layout of the VES, it may be difficult to dry the system following a hydrostatic test. Traces of water could result in sending a slug of water through the system or rust to form. Allowing for pneumatic testing continues to meet the ASME Section III code. The proposed changes do not affect the operation of the VES. The VES maintains its design function to maintain control room habitability.

The proposed changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. Therefore, the probabilities of accidents previously evaluated are not affected.

The proposed changes do not affect the prevention and mitigation of other abnormal events (e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles), or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the Updated Final Safety Analysis Report are not affected.

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

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

Response: No.

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

The proposed changes do not affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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

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

Response: No.

The proposed changes allow for pneumatic testing of the VES ASME Section III components and piping. The VES ASME Section III components and piping continue to meet the ASME Section III code. The proposed changes do not have any effect on the ability of the safety-related SSCs to perform their design basis functions. The proposed changes do not affect the ability of the VES to maintain control room habitability.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the requested amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: January 31, 2018. A publicly-available version is in ADAMS under Accession No. ML18031B131.

Description of amendment request: The requested amendment proposes changes to the Technical Specification (TS) 3.4.6, Pressurizer Safety Valve, Applicability to require the pressurizer safety valves (PSVs) to be operable when the TS 3.4.14, Low Temperature Overpressure Protection (LTOP), is not required to be operable. A conforming change to the TS 3.4.6 Actions is also proposed. Additional TS changes necessary to support PSVs operability are proposed for consistency with the TS 3.4.6 change.

The request also proposes moving TS Limiting Condition for Operation Notes regarding reactor coolant pump starts from TS 3.4.4, Reactor Coolant System (RCS) Loops, 3.4.8, Minimum RCS Flow, and 3.4.14, Low Temperature Overpressure Protection (LTOP), to TS 3.4.3, RCS Pressure/Temperature (P/T) Limits.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events.

The proposed changes do not affect the physical design of SSCs related to the TS on Engineered Safety Features Actuation System (ESFAS), RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, or Reactor Vessel head vent (RVHV), as described in the Updated Final Safety Analysis Report (UFSAR). Therefore, the operation of the listed functions and components is not affected. Therefore, the proposed changes do not affect the probability of an accident previously evaluated.

The proposed changes do not affect the physical design of SSCs related to the TS on ESFAS, RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, or RVHV to meet their design functions. The design of the functions and components continue to meet the same regulatory acceptance criteria, codes, and standards as stated in the UFSAR. In addition, the proposed changes maintain the capabilities of the ESFAS, RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, or RVHV to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria.

The proposed changes do not affect the prevention and mitigation of other abnormal events (e.g., anticipated operational occurrences, earthquakes, floods, and turbine missiles), or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

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

The proposed changes do not affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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

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

Response: No.

The proposed changes maintain existing safety margins. The proposed changes verify and maintain the physical design of SSCs related to ESFAS, RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, and RVHV to perform their design functions. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the requested amendment does not involve a significant reduction in 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 hazard consideration.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: February 2, 2018. A publicly-available version is in ADAMS under Accession No. ML18037B114.

Description of amendment request: The requested amendment proposes departures from the generic AP1000 Design Control Document (DCD) for the plant-specific VEGP Combined License (COL) Appendix A Technical Specifications (TS) and related departures from generic DCD Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific DCD Tier 2 information). Specifically, the proposed changes would make administrative changes to COL Appendix A, TS 5.6.3, for the core operating limits report required documentation to include analytical methods which are described elsewhere in the TS and in the UFSAR, and make an editorial change to COL Appendix A TS 5.7.2 for high radiation areas to correct a typographical error.

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 are administrative and editorial changes consistent with the requirements described elsewhere in the TS and in the UFSAR, and do not adversely affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. The proposed changes to the analytical methods approved for maintaining core operating limits do not result in any increase in probability of an analyzed accident occurring, and prevent power oscillations and maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that fuel design limits are not exceeded for events resulting in positive reactivity insertion and reactivity feedback effects, and so that the consequences of postulated accidents are not changed. The proposed changes do not adversely affect the ability of the automatic reactor trips to perform the required safety function to trip the reactor when necessary to protect fuel design limits, and do not adversely affect the probability of inadvertent operation or failure of the automatic reactor trips.

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 are administrative and editorial changes consistent with the requirements described elsewhere in the TS and in the UFSAR, and do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes to the analytical methods approved for maintaining core operating limits do not result in any increase in probability of an analyzed accident occurring, and prevent power oscillations and maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that fuel design limits are not exceeded for events resulting in positive reactivity insertion and reactivity feedback effects, and so that the consequences of postulated accidents are not changed. The proposed changes do not adversely affect the ability of the automatic reactor trips to perform the required safety function to trip the reactor when necessary to protect fuel design limits, and do not adversely affect the probability of inadvertent operation or failure of the automatic reactor trips.

These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that results in significant fuel cladding failures.

Therefore, the 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 are administrative and editorial changes consistent with the requirements described elsewhere in the TS and in the UFSAR, and maintain existing safety margins through continued application of the existing requirements of the UFSAR. The proposed changes maintain the initial conditions and operating limits required by the accident analysis, and the analyses of normal operation and anticipated operational occurrences, so that the existing fuel design limits specified in the UFSAR are not exceeded for events resulting in positive reactivity insertion and reactivity feedback effects, and so that the consequences of postulated accidents are not changed. Therefore, the proposed changes satisfy the same safety functions in accordance with the same requirements as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2, Rhea County, Tennessee

Date of amendment request: January 5, 2018. A publicly-available version is in ADAMS under Accession No. ML18008A257.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.6.3, “Containment Isolation Valves,” and Surveillance Requirement 3.6.3.5 to change the frequency in accordance with the Watts Bar Containment Leakage Rate Testing Program, which is described in TS 5.7.2.19. The proposed change would allow leak rate testing of the containment purge system containment isolation valves to be performed at least once every 30 months, as prescribed in Regulatory Guide 1.163.

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

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

Response: No.

The proposed change deletes the augmented testing requirement for these containment isolation valves and allows the surveillance intervals to be set in accordance with the Containment Leakage Rate Testing Program. This change does not affect the system function or design. The purge valves are not an initiator of any previously analyzed accident. Leakage rates do not affect the probability of the occurrence of any accident. Operating history has demonstrated that the valves do not degrade and cause leakage as previously anticipated. Because these valves have been demonstrated to be reliable, these valves can be expected to perform the containment isolation function as assumed in the accident analyses. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes do not significantly increase the probability of an accident and are consistent with safety analysis assumptions and resultant consequences.

Therefore, the changes do 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.

This change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing any normal plant operation. The change does not alter assumptions made in the safety analyses or licensing basis. Extending the test intervals has no influence on, nor does it contribute in any way to, the possibility of a new or different kind of accident or malfunction from those previously analyzed. No change has been made to the design, function, or method of performing leakage testing. Leakage acceptance criteria have not changed. No new accident modes are created by extending the testing intervals. No safety-related equipment or safety functions are altered as a result of this change.

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 only margin of safety that has the potential of being impacted by the proposed change involves the offsite dose consequences of postulated accidents, which are directly related to the containment leakage rate. The proposed change does not alter the method of performing the tests nor does it change the leakage acceptance criteria. Sufficient data has been collected to demonstrate these resilient seals do not degrade at an accelerated rate. Because of this demonstrated reliability, this change will provide sufficient surveillance to determine an increase in the unfiltered leakage prior to the leakage exceeding that assumed in the accident analysis.

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

NRC Branch Chief: Undine Shoop.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee

Date of amendment request: October 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17284A452.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.3.1, Table 3.3.1-1, “Reactor Trip System (RPS) Instrumentation,” to increase the values for the nominal trip setpoint and the allowable value for Function 14.a, “Turbine Trip—Low Fluid Oil Pressure.” The proposed changes are due to the planned replacement and relocation of the pressure switches from the low pressure auto-stop trip fluid oil header to the high pressure turbine electrohydraulic control (EHC) oil header. The changes are needed due to the higher EHC system operating pressure.

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 reflects a design change to the turbine control system that results in the use of an increased control oil pressure system, necessitating a change to the value at which a low fluid oil pressure initiates a reactor trip on turbine trip. The low fluid oil pressure is an input to the reactor trip instrumentation in response to a turbine trip event. The value at which the low fluid oil initiates a reactor trip is not an accident initiator. A change in the nominal control oil pressure does not introduce any mechanisms that would increase the probability of an accident previously analyzed. The reactor trip on turbine trip function is initiated by the same protective signal as used for the existing auto stop low fluid oil system trip signal. There is no change in form or function of this signal and the probability or consequences of previously analyzed accidents are not impacted.

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

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

Response: No.

The EHC fluid oil pressure rapidly decreases in response to a turbine trip signal. The value at which the low fluid oil pressure switches initiates a reactor trip is not an accident initiator. The proposed TS change reflects the higher pressure that will be sensed after the pressure switches are relocated from the auto stop low fluid oil system to the EHC high pressure header. Failure of the new switches would not result in a different outcome than is considered in the current design basis. Further, the change does not alter assumptions made in the safety analysis but ensures that the instruments perform as assumed in the accident analysis.

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

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

Response: No.

The change involves a parameter that initiates an anticipatory reactor trip following a turbine trip. The safety analyses do not credit this anticipatory trip for reactor core protection. The original pressure switch configuration and the new pressure switch configuration both generate the same reactor trip signal. The difference is that the initiation of the trip will now be adjusted to a different system of higher pressure. This system function of sensing and transmitting a reactor trip signal on turbine trip remains the same. There is no impact to safety analysis acceptance criteria as described in the plant licensing basis because no change is made to the accident analysis assumptions.

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

NRC Branch Chief: Undine Shoop.

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 Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: September 14, 2017, as supplemented by letter dated December 12, 2017.

Brief description of amendments: The amendments modified Technical Specifications (TSs) to allow temporary changes to TSs 3.5.2, “Emergency Core Cooling Systems (ECCS)—Operating”; 3.6.6, “Containment Spray System”; 3.7.5, “Auxiliary Feedwater (AFW) System”; 3.7.6, “Component Cooling Water (CCW) System”; 3.7.7, “Nuclear Service Water System (NSWS)”; 3.7.9, “Control Room Area Ventilation System (CRAVS)”; 3.7.11, “Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)”; and 3.8.1, “[Alternating Current] Sources—Operating,” to permit the “A” Train NSWS to be inoperable for a total of 14 days to address a non-conforming condition on the “A” Train supply piping from the Standby Nuclear Service Water Pond.

Date of issuance: February 15, 2018.

Effective date: These license amendments are effective as of its date of issuance and shall be implemented within 120 days of issuance.

Amendment Nos.: 308 (Unit 1) and 287 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18030A682; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Renewed Licenses and TSs.

Date of initial notice in Federal Register: December 19, 2017 (82 FR 60226).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 15, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 27, 2017.

Brief description of amendment: The amendment revised the Technical Specification (TS) requirements in order to address Generic Letter 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” dated January 11, 2008, as described in Technical Specifications Task Force (TSTF) Traveler TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Date of issuance: February 16, 2018.

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

Amendment No.: 246. A publicly-available version is in ADAMS under Accession No. ML18025A213; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: June 6, 2017 (82 FR 26132).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 16, 2018.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 14, 2017.

Brief description of amendment: The amendment revised the ANO-1 Technical Specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9, which allows a delay time for entering a supported system TS when the inoperability is solely due to an unavailable barrier. The change is consistent with Technical Specification Task Force (TSTF)-427, Revision 2, “Allowance for Non Technical Specification Barrier Degradation Supported System OPERABILITY.” In addition, the amendment corrected a typographical omission on TS page 3.0-3, which was editorial in nature.

Date of issuance: February 26, 2018.

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

Amendment No.: 259. A publicly-available version is in ADAMS under Accession No. ML18033A175; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: October 24, 2017 (82 FR 49236).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2018.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2 (ANO-2), Pope County, Arkansas

Date of amendment request: August 14, 2017.

Brief description of amendment: The amendment revised the ANO-2 Technical Specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9, which allows a delay time for entering a supported system TS when the inoperability is solely due to an unavailable barrier. The change is consistent with Technical Specification Task Force (TSTF)-427, Revision 2, “Allowance for Non Technical Specification Barrier Degradation Supported System OPERABILITY.”

Date of issuance: February 26, 2018.

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

Amendment No.: 309. A publicly-available version is in ADAMS under Accession No. ML18051A589; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-6: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: October 24, 2017 (82 FR 49237).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2018.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2, Calvert County, Maryland

Date of amendment request: March 28, 2017.

Brief description of amendments: The amendments revised the Calvert Cliffs, Units 1 and 2, Technical Specifications (TSs) to change the low level of the refueling water tank to reflect a needed increase in the required borated water volume and change the allowable value of the refueling water tank level-low function.

Date of issuance: February 15, 2018.

Effective date: As of the date of issuance and shall be implemented within 60 days of the end of CC1R24 refueling outage for Calvert Cliffs, Unit 1, and within 60 days of the end of CC2R23 refueling outage for Calvert Cliffs, Unit 2.

Amendment Nos.: 323 (Unit 1) and 301 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18029A195; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: June 19, 2017 (82 FR 27887).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 15, 2018.

No significant hazards consideration comments received: No.

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.

Brief description of amendments: The amendments revised 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).

Date of issuance: February 20, 2018.

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

Amendment Nos.: 278 (Unit 3) and 273 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML18018A559; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: September 12, 2017 (82 FR 42850).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 20, 2018.

No significant hazards consideration comments received: No.

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

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

Date of amendment request: February 13, 2017, as supplemented by letter dated August 11, 2017.

Brief description of amendments: The amendments adopted the NRC-endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors.”

Date of issuance: February 16, 2018.

Effective date: As of the date of issuance and shall be implemented within a 365-day period after issuance.

Amendment Nos.: Salem—322 (Unit No. 1) and 303 (Unit No. 2); Hope Creek—210. A publicly-available version is in ADAMS under Accession No. ML17355A570; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-70, DPR-75, and NPF-57: The amendments revised the emergency action level technical bases documents.

Date of initial notice in Federal Register: March 28, 2017 (82 FR 15384). The supplemental letter dated August 11, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 16, 2018.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: May 26, 2016, as supplemented by letter dated October 26, 2017.

Brief description of amendments: The amendments correct a non-conservative Technical Specification (TS) Surveillance Requirement acceptance criterion for the diesel generator steady-state frequency in Limiting Condition for Operation 3.8.1, “AC [Alternating Current] Sources—Operating.”

Date of issuance: February 12, 2018.

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

Amendment Nos.: 341—Unit 1 and 334—Unit 2. A publicly-available version is in ADAMS under Accession No. ML18026A810; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-77 and DPR-79. Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: August 2, 2016 (81 FR 50740). The supplemental letter dated October 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 amendments is contained in a Safety Evaluation dated February 12, 2018.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, on March 6, 2018.

For the Nuclear Regulatory Commission.

Gregory F. Suber,

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