2017-25901. 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 November 7, 2017, to November 17, 2017. The last biweekly notice was published on November 21, 2017.

DATES:

Comments must be filed by January 4, 2018. A request for a hearing must be filed by February 5, 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0225. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: May Ma, Office of Administration, Mail Stop: OWFN-2-A13, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0225, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0225.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2017-0225, 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 Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

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

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

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

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

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

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

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

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

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

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

Date of amendment request: September 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17270A041.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to reflect the addition of a second qualified offsite power circuit. In addition, the proposed amendment requests approval to change the Updated Final Safety Analysis Report (UFSAR) to allow for the use of automatic load tap changers (LTCs) on the new (230 kilovolt (kV)) and the replacement (115kV) startup transformers.

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 revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO [Limiting Condition for Operation], Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431 [“Standard Technical Specifications—Westinghouse Plants”]. The AC [alternating current] power systems are not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not increased. The consequences of an accident with the proposed LCO requiring two qualified offsite circuits between the offsite transmission network and the onsite emergency AC Electrical Power Distribution System to be operable are no different than the consequences of an accident in Modes 1, 2, 3, and 4 with the existing LCO that requires the single qualified offsite circuit to be operable. The additional 230kV startup transformer will improve the reliability and availability of offsite power to the emergency buses by increasing the amount of available offsite power sources from one to two. The two qualified offsite circuits are designed to mitigate the consequences of previously evaluated accidents. The proposed change to TS 3.8.1 would not change any of the previously evaluated accidents in the UFSAR.

The proposed change will also allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. The only accident previously evaluated where the probability of an accident is potentially affected by the proposed change is a loss of offsite power (LOOP). Failure of a LTC while in the automatic mode of operation that results in decreased voltage to the safety related buses could cause a LOOP if voltage decreased below the degraded grid voltage relay (DGVR) setpoint. The three postulated failure scenarios are: (1) Failure of a primary microcontroller that results in rapidly decreasing voltage supplied to the safety related buses; (2) failure of a primary microcontroller to respond to decreasing grid voltage; and (3) the backup microcontroller overrides the primary microcontroller when not required. For the first scenario, a backup microcontroller is provided for each LTC, which makes this failure unlikely. For the second scenario, operators would have ample time to address the condition utilizing identified procedures since grid voltage changes typically occur relatively slowly. In addition, the frequency of occurrence of all of these failure modes is small, based on the operating history of similar equipment at other plants. Furthermore, in all of the above potential failure modes, operators can take manual control of the LTC to mitigate the effects of the failure. Thus, the probability of a LOOP will not be significantly increased by operation of the LTCs in the automatic mode. The proposed change to allow operation of the LTCs in automatic mode has no effect on the consequences of a LOOP, since the emergency diesel generators (EDGs) provide power to safety related equipment following a LOOP. The design and function of the EDGs are not affected by the proposed change. The LTCs are each equipped with a backup microcontroller, which inhibits gross improper action of the LTC in the event of primary microcontroller failure. Additionally, the operator has procedurally identified actions available to prevent a sustained high voltage condition from occurring. Damage due to overvoltage is time-dependent, requiring a sustained high voltage condition. Therefore, damage to safety related equipment is unlikely, and the consequences of previously evaluated accidents 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 evaluated?

Response: No.

The proposed change revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO, Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431. The proposed change also will allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. All aspects of the proposed change involve electrical transformers that provide offsite power to safety-related equipment for accident mitigation. The proposed change does not alter the design, physical configuration or mode of operation of any other plant structure, system or component. No physical changes are being made to any other portion of the plant, so no new accident causal mechanisms are being introduced. The proposed change also does not result in any new mechanisms that could initiate damage to the reactor or its principal safety barriers (i.e., fuel cladding, reactor coolant system or primary containment).

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 proposed change revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO, Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431. The new 230kV startup transformer will improve the reliability and availability of offsite power to the emergency buses by increasing the amount of available offsite power sources from one to two. Another improvement to the HBRSEP electrical system configuration as a result of the proposed change is that each emergency bus will be normally aligned to independent startup sources and will not require a fast bus transfer on a unit trip. This reduces the risk of loss of power to the emergency buses caused by power transfer and/or equipment failures. The margin of safety is increased with the proposed change to revise TS 3.8.1 to reflect the additional qualified offsite circuit.

The proposed change will also allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. The inputs or assumptions of any of the analyses that demonstrate the integrity of the fuel cladding, reactor coolant system or containment during accident conditions are unaffected by this proposed change. The allowable values for the degraded voltage protection function are unchanged and will continue to ensure that the degraded voltage protection function actuates when required, but does not actuate prematurely to unnecessarily transfer safety related loads from offsite power to the EDGs. Automatic operation of the LTCs increases the margin of safety by reducing the potential for transferring loads to the EDGs during an undervoltage or overvoltage event on the offsite power sources.

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

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

Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Undine Shoop.

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

Date of amendment request: November 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17306A086.

Description of amendment request: The proposed amendment would revise the PNP renewed facility operating license (RFOL) to change the full compliance implementation date for the fire protection program transition license condition. Specifically, the licensee is requesting additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c), “National Fire Protection Association Standard NFPA 805.”

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 PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents, and have no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed changes to the PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

The proposed changes to the PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety.

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

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

Attorney for licensee: William Glew, Associate General Counsel Nuclear, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: October 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A910.

Description of amendment request: The amendment would revise the ANO-1 Technical Specification (TS) 3.7.5, “Emergency Feedwater (EFW) System,” Bases to stipulate the conditions in which the TS 3.7.5, Condition A, 7-day Completion Time should apply to the ANO-1 turbine-driven EFW pump steam supply valves.

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 EFW system is not an initiator of any design basis accident or event and, therefore, the proposed change does not increase the probability of any accident previously evaluated. The proposed change to clarify the conditions in which the current 7-day Completion Time for an inoperable steam supply path to turbine-driven EFW pump does not change the response of the plant to any accidents, since single failure criterion is not applicable when complying with associated TS Actions.

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

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

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

Response: No.

The proposed change does not result in a change in the manner in which the EFW system provides plant protection. Absent a single failure (which is not assumed while in compliance with TS Actions), the EFW system will continue to supply water to the Steam Generators (SGs) to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the SGs, as required. There are no design changes associated with the proposed change. The change to the associated TS Bases does not change any existing accident scenarios, nor create any new or different accident scenarios.

The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the change clarifies the application of the current 7-day Completion Time for an inoperable steam supply path to the turbine-driven EFW pump and does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions, which does not assume an EFW system single failure when complying with TS Actions, and current plant operating practice.

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

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

Response: No.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. The proposed change will not result in plant operation in a configuration outside the design basis. The associated TS will continue to limit the time in which one steam supply path to the turbine-driven EFW pump may be inoperable.

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

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

Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, Washington, DC 20001.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: July 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17209A755.

Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix A, plant-specific Technical Specifications (TS) to make them consistent with the remainder of the design licensing basis and the TS. Specifically, the requested amendment proposes changes to COL Appendix A, the Technical Specification updates for reactivity controls and other miscellaneous changes, and Updated Final Safety Analysis Report (UFSAR) information in various locations.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The change applies to a Diverse Actuation System (DAS) Manual Controls Mode 6 note for operability of the Automatic Depressurization System (ADS) Stage 4 valves that involves revising the note from reactor internals in place to upper internals in place. In accordance with Limiting Condition for Operation (LCO) 3.4.13 ADS—Shutdown, Reactor Coolant System (RCS) Open Applicability and TS 3.3.9, Engineered Safeguards Actuation System Instrumentation, Function 7, the ADS Stage 4 valves are not required to be operable in MODE 6 with the upper internals removed. However, the reactor internals would still be present. The change involves clarification of the note (with no change in required system or device function), such that the appropriate configuration in Mode 6 would be in place and would not conflict with TS 3.4.13 or TS 3.3.9. The revised note previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

The consequences of an accident as a result of the revised note and associated requirements and actions are no different than the consequences of the same accident during the existing ones. As a result, the consequences are not affected by this change.

The proposed change does not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change involves revising the existing LCO 3.1.4 operability to be applicable to Rod Cluster Control Assemblies (RCCAs)with accompanying changes in actions and surveillance requirements (with no change in required system or device function), such that more appropriate, albeit less restrictive, actions would be applied. The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no set points, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this 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 change will not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the LCO 3.1.4 for Rod Group Alignment Limits is made less restrictive by eliminating the worth of the [Gray Rod Cluster Assemblies (GRCAs)] in MODES 1 and 2 with k_eff ≥1, no credit is taken in the current design basis for including their trip reactivity worth. As such, there is no significant reduction in a margin of safety. 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 hazards consideration.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: September 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17257A177.

Description of amendment request: The amendments would revise Technical Specification (TS) 5.5.17, “Containment Leakage Rate Testing Program,” for the Vogtle Electric Generating Plant, Units 1 and 2, to (1) increase the existing Type A integrated leakage rate test interval from 10 to 15 years, (2) extend the Type C containment isolation valve leaking testing to a 75-month frequency, (3) adopt the use of American National Standards Institute/American Nuclear Society 56.8-2002, “Containment System Leakage Testing Requirements,” and (4) adopt a more conservative grace interval of 9 months for Type A, B, and C tests in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J.”

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 activity involves the revision of Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Technical Specification (TS) Section 5.5.17, “Primary Containment Leakage Rate Testing Program,” to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years, and the extension of the Type C local leakage rate test (LLRT) interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events (IE) probabilistic risk analysis (PRA) is 1.79E-03 person-rem/year for Unit 1 and Unit 2. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A states that a very small population is defined as an increase of ≤1.0 person-rem per year or ≤1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325. Moreover, the risk impact when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, “Performance-Based Containment Leak-Test Program,” dated September 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The VEGP Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and (2) time based. Activity-based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. The LLRT requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Section XI, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes exceptions previously granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

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

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

Response: No.

The proposed amendment to the TS 5.5.17, Containment Leakage Rate Testing Program, involves the extension of the VEGP Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. 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.

The proposed amendment also deletes exceptions previously granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is operated.

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 amendment to TS 5.5.17 involves the extension of the VEGP Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leaktightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for VEGP. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusions that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. This exception was for an activity that would have already taken place by the time this amendment is approved; therefore, the deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any margin of safety as a result of this administrative change.

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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: September 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17261B272.

Description of amendment request: The amendment would relocate the defined core plane regions where the radial peaking factor limits are not applicable, from Technical Specification (TS) 4.2.2.2.f to the Core Operating Limits Reports (COLR) for STP Units 1 and 2. The amendment would also revise the COLR Administrative Controls TS to add exclusion zones to the list of limits found in the COLRs, and to revise the description of the methodology used to determine the values. In addition, the proposed amendment requests administrative changes to the TSs.

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 relocation of the F_xy exclusion zones to the COLRs has no impact on the accidents analyzed in the STPNOC UFSAR [Updated Final Safety Analysis Report] and is not an accident initiator. Since the change does not impact any conditions that would initiate an accident, the probability or consequences of previously analyzed events is not increased. The proposed amendment does not change the actions to be taken if a core operating limit is exceeded and there are no physical changes associated with this proposed amendment.

For each core reload, each accident analysis addressed in the STP UFSAR will continue to be examined with respect to changes in the cycle-dependent parameters, which are obtained from the use of NRC-approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses. This examination, which will be conducted per the requirements of 10 CFR 50.59, will ensure that future core reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, there is no impact to the probability or consequences of an accident previously evaluated due to the proposed change.

[The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.]

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

Response: No.

The relocation of the F_xy exclusion zone details from the Technical Specifications to the COLRs will not create the possibility of a new or different kind of accident from any accident previously evaluated. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. No new operator actions are created as a result of the proposed change. The cycle-specific variables are determined using the NRC approved methods and the COLRs are submitted to the NRC to allow the staff to continue to trend the values of these limits. The Technical Specifications will continue to require operation within the core operating limits and appropriate actions will be required if these limits are exceeded.

The relocation of the F_xy exclusion zones to the COLRs has no impact on the accidents analyzed in the STPNOC Updated Final Safety Analysis Report (UFSAR) and is not an accident initiator. Since this change does not impact any conditions that would initiate an accident, there is no possibility of a new or different kind of accident resulting from this change.

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

[The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.]

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

Response: No.

The relocation of the F_xy exclusion zone details from the Technical Specifications to the COLRs will not affect the margin of safety. The margin of safety presently provided by the Technical Specifications remains unchanged. They will be incorporated into the COLR which is submitted to the NRC, therefore appropriate measures exist to control the values of these limits. The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. STPNOC will continue to confirm all safety analysis limits remain bounding on a cycle-specific basis using an NRC-approved methodology. Each core reload will involve a Reload Safety Evaluation to assure that operation of the unit within the cycle specific limits will not involve a significant reduction in the margin of safety.

The proposed amendment does not affect the design of the facility or system operating parameters, does not physically alter safety-related systems and does not affect the method in which safety-related systems perform their functions.

Therefore, the proposed change does not impact margin of safety.

[The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.]

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

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

NRC Branch Chief: Robert J. Pascarelli.

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

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

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

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

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

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 26, 2016, as supplemented by letters dated February 16, July 17, August 8, September 27, October 3, and November 8, 2017.

Brief description of amendments: The amendments revised Technical Specification (TS) 5.5.13, “Primary Containment Leakage Rate Testing Program,” to allow for the permanent extension of the Type A Integrated Leak Rate Testing and Type C Leak Rate Testing frequencies, to change the documents used by LSCS to implement the performance-based leakage testing program, and to delete the information regarding the performance of the next LSCS Type A tests to be performed.

Additionally, the amendments deleted Conditions 2.D.(e) and 2.D.(c), respectively, of the LSCS Unit 1 and Unit 2 Renewed Facility Operating Licenses regarding conducting the third Type A test of each 10-year service period when the plant is shut down for the 10-year inservice inspection.

Date of issuance: November 16, 2017.

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

Amendment Nos.: 226 (Unit 1) and 212 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17283A085; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: February 14, 2017 (82 FR 10597). The supplemental letters dated February 16, July 17, August 8, September 27, October 3, and November 8, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: February 17, 2017, as supplemented by letters dated March 20, 2017; July 13, 2017; August 8, 2017; August 30, 2017; and September 15, 2017.

Brief description of amendments: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications to implement a measurement uncertainty recapture power uprate. Specifically, the amendments authorized an increase in the maximum licensed thermal power level from 3,951 megawatts thermal to 4,016 megawatts thermal, which is an increase of approximately 1.66 percent.

Date of issuance: November 15, 2017.

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

Amendments Nos.: 316 (Unit 2) and 319 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17286A013; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: May 2, 2017 (82 FR 20497). The supplemental letters dated March 20, 2017; July 13, 2017; August 8, 2017; August 30, 2017; and September 15, 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 November 15, 2017.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1 (FCS), Washington County, Nebraska

Date of amendment request: October 25, 2016, as supplemented by letter dated September 25, 2017.

Brief description of amendment: The amendment revised the FCS Updated Safety Analysis Report to change the structural design methodology for the Auxiliary Building at FCS. Specifically, the amendment made the following changes: (1) Use of the ultimate strength design method from the industry standard American Concrete Institute (ACI) 318-63, “Publication SP-10, Commentary on Building Code Requirements for Reinforced Concrete,” for normal operating/service conditions for future designs and evaluations; (2) use higher concrete compressive strength values for Class B concrete, based on original strength test data; (3) use higher reinforcing steel yield strength values, based on original strength test data; and (4) make minor clarifications, including adding a definition of control fluids to the dead load section of the Updated Safety Analysis Report.

Date of issuance: November 17, 2017.

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

Amendment No.: 293. A publicly-available version is in ADAMS under Accession No. ML17278A607; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-40: The amendment revised the Emergency Plan and Emergency Action Level Scheme.

Date of initial notice in Federal Register: January 17, 2017 (82 FR 4930).

The supplemental letter dated September 25, 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 amendment is contained in a safety evaluation dated November 17, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: November 17, 2016, as supplemented by letters dated August 7, 2017, and October 18, 2017.

Brief description of amendments: The amendments revised Technical Specification requirements regarding accident monitoring instrumentation. Specifically, the amendments modified the list of instruments required to be operable based on implementation of Regulatory Guide 1.97, Revision 2, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” In addition, allowed outage times and required actions for inoperable accident monitoring instrumentation channels have been revised to be consistent with NUREG-1431, Revision 4.0, “Standard Technical Specifications—Westinghouse Plants.”

Date of issuance: November 14, 2017.

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

Amendment Nos.: 320 (Unit 1) and 301(Unit 2). A publicly-available version is in ADAMS under Accession No. ML17227A016; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: January 17, 2017 (82 FR 4931). The supplemental letters dated August 7, 2017, and October 18, 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 November 14, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 7, 2017, as supplemented by letters dated March 27, 2017, and July 13, 2017.

Brief description of amendment: The amendment modified Hope Creek Generating Station Technical Specification 6.8.4.f, “Primary Containment Leakage Rate Testing Program,” to extend the Type A reactor containment pressure test interval from one test in 10 years to one test in 15 years, and extend the Type C test interval up to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergency conditions.

Date of issuance: November 8, 2017.

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

Amendment No.: 207. A publicly-available version is in ADAMS under Accession No. ML17291A209; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: December 20, 2016 (81 FR 92869). The supplemental letters dated March 27, 2017, and July 13, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 8, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 17, 2017, as supplemented by letter dated June 29, 2017.

Brief description of amendments: The amendments change technical specifications (TSs) consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” and TSTF-299, Revision 0, “Administrative Controls Program 5.5.2.b Test Interval and Exception.”

Date of issuance: November 8, 2017.

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

Amendment Nos.: 301 (Unit 1), 325 (Unit 2), and 285 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17277A207; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: April 25, 2017 (82 FR 19106). The supplemental letter dated June 29, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 8, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 28, 2017.

Brief description of amendment: The amendment revised the completion date for License Condition 2.C.(5) for Watts Bar Nuclear Plant, Unit 2, regarding the completion of action to resolve the issues identified in Bulletin 2012-01, “Design Vulnerability in Electric Power System” (ADAMS Accession No. ML12074A115), from December 31, 2017, to December 31, 2018, to align with the remainder of the Tennessee Valley Authority fleet and with the nuclear industry.

Date of issuance: November 6, 2017.

Effective date: As of the date of issuance and shall be implemented within 15 days of issuance.

Amendment No.: 17. A publicly-available version is in ADAMS under Accession No. ML17258A328; documents related to this amendment is listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-96: Amendment revised the Facility Operating License.

Date of initial notice in Federal Register: July 5, 2017 (82 FR 31103).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 6, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 27th day of November 2017.

For the Nuclear Regulatory Commission.

Kathryn M. Brock,

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