2017-10570. 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 April 25, 2017, to May 8, 2017. The last biweekly notice was published on May 9, 2017.

DATES:

Comments must be filed by June 22, 2017. A request for a hearing must be filed by July 24, 2017.

ADDRESSES:

You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0120. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: T-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0120, 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-2017-0120.
• 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-2017-0120, 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 by July 24, 2017. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

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

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on obtaining 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 (CNS), Units 1 and 2, York County, South Carolina

Date of amendment request: December 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16350A422.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.4, “Residual Heat Removal (RHR) and Coolant Circulation—High Water Level,” and TS 3.9.5, “Residual Heat Removal (RHR) and Coolant Circulation—Low Water Level.” Condition A of TS 3.9.4 applies when RHR requirements are not met, and includes four required actions. Required Action A.4 requires, within 4 hours, the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes revise Required Action A.4 and add new Required Actions A.5, A.6.1, and A.6.2 to clarify that the intent of the required actions is to establish containment closure. Each of these required actions will have a completion time of 4 hours. Condition B of TS 3.9.5 applies when no RHR loop is in operation, and includes three required actions. Required Action B.3 requires the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes are the same as the proposed changes to TS 3.9.4, consisting of a revision to Required Action B.3 and the addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-197-A, Revision 2, “Require Containment Closure When Shutdown Cooling Requirements Are Not Met.”

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 revise the CNS TS to ensure that the appropriate actions are taken to establish containment closure in the event that Residual Heat Removal requirements are not met during refueling operations. Containment closure would be appropriate for mitigation of a loss of shutdown cooling accident, but it does not affect the initiation of the accident. The containment purge system isolation valves will be capable of being closed automatically on a high containment radiation signal, such that there will be no significant increase in the radiological consequences of a loss of shutdown cooling.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal.

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 the margin of safety?

Response: No.

Currently the Technical Specifications are vague and overly restrictive concerning the requirement for containment closure when shutdown cooling is lost. The proposed changes eliminate unclear requirements and provide a clear way to establish containment closure that meets the [TS] Bases description, which is to prevent radioactive gas from being released from the containment during a loss of shutdown cooling incident. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification 3.1.2, “Core Reactivity,” to revise the Completion Times of Required Action A.1 and A.2 from 72 hours to 7 days. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-142-A, Revision 0, “Increase the Completion Time when the Core Reactivity Balance is Not Within Limit.”

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 extend the Completion Time to take the Required Actions when measured core reactivity is not within the specified limit of the predicted values. The Completion Time to respond to a difference between predicted and measured core reactivity is not an initiator to any accident previously evaluated. The radiological consequences of an accident during the proposed Completion Time are no different from the consequences of an accident during the existing Completion Time.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

The proposed changes provide additional time to investigate and to implement appropriate operating restrictions when measured core reactivity is not within the specified limit of the predicted values. The additional time will not have a significant effect on plant safety due to the conservatisms used in designing the reactor core and performing the safety analyses, and the low probability of an accident or transient which would approach the core design limits during the additional time.

Therefore, 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: Kate B. Nolan, Deputy General Counsel, Duke Energy Corporation, LLC, 550 South Tryon Street—DEC45A, Charlotte, NC 28202-1802.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.5, “Residual Heat Removal (RHR) and Coolant Circulation—High Water Level,” and TS 3.9.6, “Residual Heat Removal (RHR) and Coolant Circulation—Low Water Level.” Condition A of TS 3.9.5 applies when RHR requirements are not met, and includes four required actions. Required Action A.4 requires, within 4 hours, the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes revise Required Action A.4 and add new Required Actions A.5, A.6.1, and A.6.2 to clarify that the intent of the required actions is to establish containment closure. Each of these required actions will have a completion time of 4 hours. Condition B of TS 3.9.6 applies when no RHR loop is in operation, and includes three required actions. Required Action B.3 requires the closure of all containment penetrations providing direct access from containment atmosphere to outside atmosphere. The proposed changes are the same as the proposed changes to TS 3.9.5, consisting of a revision to Required Action B.3 and the addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-197-A, Revision 2, “Require Containment Closure When Shutdown Cooling Requirements Are Not Met.”

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 revise the MNS TS to ensure that the appropriate actions are taken to establish containment closure in the event that Residual Heat Removal requirements are not met during refueling operations. Containment closure would be appropriate for mitigation of a loss of shutdown cooling accident, but it does not affect the initiation of the accident. The containment purge system isolation valves will be capable of being closed automatically on a high containment radiation signal, such that there will be no significant increase in the radiological consequences of a loss of shutdown cooling.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal.

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 the margin of safety?

Response: No.

Currently the Technical Specifications are vague and overly restrictive concerning the requirement for containment closure when shutdown cooling is lost. The proposed changes eliminate unclear requirements and provide a clear way to establish containment closure that meets the [TS] Bases description, which is to prevent radioactive gas from being released from the containment during a loss of shutdown cooling incident. The containment purge system isolation valves will remain capable of being closed automatically on a high containment radiation signal.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.6.3, “Containment Isolation Valves,” to add a Note to TS Limiting Condition for Operation (LCO) 3.6.3 Required Actions A.2, C.2 and E.2 to allow isolation devices that are locked, sealed or otherwise secured to be verified by use of administrative means. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-269-A, Revision 2, “Allow Administrative Means of Position Verification for Locked or Sealed 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes modify MNS TS 3.6.3, “Containment Isolation Valves”. This TS currently includes actions that require penetrations to be isolated and periodically verified to be isolated. A Note is proposed to be added to TS 3.6.3 Required Actions A.2, C.2, and E.2, to allow isolation devices that are locked, sealed, or otherwise secured to be verified by use of administrative means. The proposed changes do not affect any plant equipment, test methods, or plant operation, and is not an initiator of any analyzed accident sequence. The inoperable containment penetrations will continue to be isolated, and hence perform their isolation function. Operation in accordance with the proposed TSs will ensure that all analyzed accidents will continue to be mitigated as previously analyzed.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

The proposed changes will not affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. Affected containment penetrations will continue to be isolated as required by the existing TS.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.8.1, “AC [Alternating Current] Sources—Operating,” to allow greater flexibility in performing Surveillance Requirements (SRs) by modifying Mode restriction notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-283-A, Revision 3, “Modify Section 3.8 Mode Restriction Notes.”

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 modify Mode restriction Notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19 to allow performance of the Surveillance in whole or in part to reestablish Diesel Generator (DG) Operability, and to allow the crediting of unplanned events that satisfy the Surveillance(s) [Requirements]. The emergency diesel generators and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. To manage any increase in risk, the proposed changes require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the current prohibited Modes. The radiological consequences of an accident previously evaluated during the period that the DG is being tested to reestablish operability are no different from the radiological consequences of an accident previously evaluated while the DG is inoperable. As a result, the consequences of any accident previously evaluated are not increased.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

The purpose of Surveillances is to verify that equipment is capable of performing its assumed safety function. The proposed changes will only allow the performance of the Surveillances to reestablish operability, and the proposed changes may not be used to remove a DG from service. In addition, the proposed changes will potentially shorten the time that a DG is unavailable because testing to reestablish operability can be performed without a plant shutdown. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the current prohibited Modes.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.4.12, “Low Temperature Overpressure Protection (LTOP) System,” to increase the time allowed for swapping charging pumps to 1 hour. Additionally, an existing note in the Applicability section of TS 3.4.12 is being reworded and relocated to the Limiting Condition for Operation section of TS 3.4.12 as Note 2. These proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-285-A, Revision 1, “Charging Pump Swap LTOP Allowance.”

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 increase the time allowed for swapping charging pumps from 15 minutes to one hour, and make several other associated administrative changes and clarifications to the TS. These changes do not affect event initiators or precursors. Thus, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated. In addition, the proposed changes do not alter any assumptions previously made in the radiological consequence evaluations nor affect mitigation of the radiological consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR). As such, the consequences of accidents previously evaluated in the UFSAR will not be increased and no additional radiological source terms are generated. Therefore, there will be no reduction in the capability of those structures, systems, and components (SSCs) in limiting the radiological consequences of previously evaluated accidents, and reasonable assurance that there is no undue risk to the health and safety of the public will continue to be provided. Thus, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

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

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

Response: No.

The proposed changes do not involve physical changes to analyzed SSCs or changes to the modes of plant operation defined in the technical specification. The proposed changes do not involve the addition or modification of plant equipment (no new or different type of equipment will be installed) nor do they alter the design or operation of any plant systems. No new accident scenarios, accident or transient initiators or precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed changes do not cause the malfunction of safety-related equipment assumed to be operable in accident analyses. No new or different mode of failure has been created and no new or different equipment performance requirements are imposed for accident mitigation. As such, the proposed changes have no effect on previously evaluated accidents.

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

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

Response: No.

The proposed changes do not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.1.8, “PHYSICS TESTS Exceptions,” to allow the numbers of channels required by the Limiting Condition for Operation (LCO) section of TS 3.3.1, “Reactor Trip System (RTS) Instrumentation,” to be reduced from “4” to “3” to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-315-A, Revision 0, “Reduce Plant Trips Due to Spurious Signals to the Nuclear Instrumentation System (NIS) During Physics Testing.”

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

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

Response: No.

The proposed changes revise TS 3.1.8, “PHYSICS TESTS Exceptions,” to allow the number of channels required by LCO 3.3.1, “RTS Instrumentation,” to be reduced from “4” to “3”, to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. A reduction in the number of required nuclear instrumentation channels is not an initiator to any accident previously evaluated. With the nuclear instrumentation channel placed in bypass instead of in trip, reactor protection is still provided by the nuclear instrumentation system operating in a two-out-of-three channel logic. As a result, the ability to mitigate any accident previously evaluated is not significantly affected. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

The proposed changes reduce the probability of a spurious reactor trip during physics testing. The reactor trip system continues to be capable of protecting the reactor utilizing the power range neutron flux trips operating in a two-out-of-three trip logic. As a result, the reactor is protected and the probability of a spurious reactor trip is significantly reduced.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System,” to expand the TS 3.7.5 Limiting Condition for Operation, Condition A to include the situation when one turbine driven AFW pump is operable in MODE 3, immediately following a refueling outage (if MODE 2 has not been entered), with a 7-day Completion Time. This proposed change is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-340-A, Revision 3, “Allow 7 Day Completion Time for a Turbine-Driven AFW Pump Inoperable.”

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 revise TS 3.7.5, “Auxiliary Feedwater (AFW) System,” to allow a 7 day Completion Time to restore an inoperable AFW turbine-driven pump in MODE 3 immediately following a refueling outage, if MODE 2 has not been entered. An inoperable AFW turbine-driven pump is not an initiator of any accident previously evaluated. The ability of the plant to mitigate an accident is no different while in the extended Completion Time than during the existing Completion Time. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

The proposed changes revise TS 3.7.5, “Auxiliary Feedwater (AFW) System,” to allow a 7 day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3, immediately following a refueling outage, if Mode 2 has not been entered. In Mode 3 immediately following a refueling outage, core decay heat is low and the need for AFW is also diminished. The two operable motor driven AFW pumps are available and there are alternate means of decay heat removal if needed. As a result, the risk presented by the extended Completion Time is minimal.

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

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.4.10, “Pressurizer Safety Valves,” TS 3.4.12, “Low Temperature Overpressure Protection (LTOP) System,” and TS 3.7.4, “Steam Generator Power Operated Relief Valves (SG PORVs),” to revise the Completion Times for Limiting Condition for Operation (LCO) 3.4.10 Required Action B.2, and LCO 3.7.4 Required Action C.2 from 12 to 24 hours and LCO 3.4.12 Required Action G.1 from 8 to 12 hours. The proposed changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-352-A, Revision 1, “Provide Consistent Completion Time to Reach MODE 4.”

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 a more reasonable time to plan and execute required actions, and will 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 changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will 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 will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the MNS Updated Final Safety Analysis Report (UFSAR). The applicable radiological dose acceptance criteria will continue to be met.

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

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

Response: No.

There are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC performs its safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed changes will not alter any assumptions made in the safety analyses.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment.

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

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

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The proposed changes will not have any impact on these barriers. No accident mitigating equipment will be adversely impacted.

Therefore, existing safety margins will be preserved. None of the proposed changes will 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: 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 Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: January 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A069.

Description of amendment request: The amendments would modify Technical Specification (TS) 3.9.6, “Residual Heat Removal (RHR) and Coolant Circulation—Low Water Level,” to add Note 1 to the Limiting Condition for Operation (LCO) Section of TS 3.9.6 to allow the securing of the operating train of RHR for up to 15 minutes to support switching operating trains. The allowance is restricted to three conditions: (a) The core outlet temperature is maintained greater than 10 degrees Fahrenheit below saturation temperature; (b) no operations are permitted that would cause an introduction of coolant into the Reactor Coolant System (RCS) with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1; and (c) no draining operations to further reduce RCS water volume are permitted. Additionally, the amendments would modify the LCO Section of TS 3.9.6 to add Note 2 which would allow one required RHR loop to be inoperable for up to 2 hours for surveillance testing, provided that the other RHR loop is operable and in operation. These proposed changes are consistent with Technical Specification Task Force (TSTF) Travelers TSTF-349-A, Revision 1, “Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation,” TSTF-361-A, Revision 2, “Allow Standby SDC [Shutdown Cooling]/RHR/DHR [Decay Heat Removal] Loop to be Inoperable to Support Testing,” and TSTF-438-A, Revision 0, “Clarify Exception Notes to be Consistent with the Requirement Being Excepted.”

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 add two notes to MNS TS LCO 3.9.6. Note 1 would allow securing the operating train of Residual Heat Removal (RHR) for up to 15 minutes to support switching operating trains, subject to certain restrictions. Note 2 to would allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is Operable and in operation. These provisions are operational allowances. Neither operational allowance is an initiator to any accident previously evaluated. In addition, the proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated.

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

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

Response: No.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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 the margin of safety?

Response: No.

An operational allowance is proposed which would allow securing the operating train of RHR for up to 15 minutes to support switching operating trains, subject to certain restrictions. Considering these restrictions, combined with the short time frame allowed to swap operating RHR trains, and the ability to start an operating RHR train, if needed, the occurrence of an event that would require immediate operation of an RHR train is extremely remote.

An operational allowance is also proposed which would allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation. A similar allowance currently appears in MNS TS 3.4.7, “Reactor Coolant System (RCS) Loops—MODE 5, Loops Filled,” and MNS TS 3.4.8, “RCS Loops—MODE 5, Loops Not Filled,” and the conditions under which the operational allowance would be applied in TS 3.9.6 are not significantly different from those specifications. This operational allowance provides the flexibility to perform surveillance testing, while ensuring that there is reasonable time for operators to respond to and mitigate any expected failures. The purpose of the RHR System is to remove decay and sensible heat from the Reactor Coolant System, to provide mixing of borated coolant, and to prevent boron stratification. Removal of system components from service as described above, and with limitations in place to maintain the ability of the RHR System to perform its safety function, does not significantly impact the margin of safety. Operators will continue to have adequate time to respond to any off-normal events. Removing the system from service, for a limited period of time, with other operational restrictions, limits the consequences to those already assumed in the Updated Final Safety Analysis Report (UFSAR).

Therefore, the proposed changes do 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: 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.

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

Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17089A380.

Description of amendment request: The proposed amendment would revise the PNP Cyber Security Plan (CSP) Milestone 8 full implementation date from December 15, 2017, to May 31, 2020. This amendment request is in support of PNP's transition, starting on October 1, 2018, from an operating power plant to a decommissioned plant.

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 CSP implementation schedule 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, system, and components relied upon to mitigate the consequences of postulated accidents, and has no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Attorney for licensee: Jeanne Cho, Senior Counsel, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: David J. Wrona.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17101A608.

Description of amendment request: The amendment would revise the renewed facility operating license Paragraph 3.G, “Physical Protection.” The amendment would revise the Pilgrim Nuclear Power Station Cyber Security Plan (CSP) implementation schedule for Milestone 8 full implementation date from December 15, 2017, to December 31, 2020.

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

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

Response: No.

The proposed change to the CSP implementation schedule is administrative in nature. The 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 has no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed change to the CSP implementation schedule is administrative in nature. The 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 previously evaluated.

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

Response: No.

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

Therefore, the proposed change does not involve a significant reduction in 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Douglas A. Broaddus.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc. (the licensees), Docket Nos. 50-416 and 72-50, Grand Gulf Nuclear Station, Unit 1 (Grand Gulf), and Independent Spent Fuel Storage Installation (ISFSI), Claiborne County, Mississippi

Date of amendment request: March 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17093A729.

Description of amendment request: The proposed amendment would make an administrative change to the name of South Mississippi Electric Power Association, one of the licensees for Grand Gulf and its ISFSI. Effective November 10, 2016, South Mississippi Electric Power Association changed its corporate name from “South Mississippi Electric Power Association” to “Cooperative Energy, a Mississippi Electric Cooperative.” The corporate name was changed for commercial reasons. The changes proposed herein to the Grand Gulf operating license solely reflects the changed licensee name. This name change is purely administrative in nature. This request does not involve a transfer of control or of an interest in the license.

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. Do the proposed changes “involve a significant increase in the probability or consequences of an accident previously evaluated”?

Response: No.

The proposed amendments simply change the name of a licensee. The name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected.

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

2. Do the proposed changes “create the possibility of a new or different kind of accident from any accident previously evaluated”?

Response: No.

The proposed amendments simply change the name of a licensee. The proposed name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected.

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 the margin of safety”?

Response: No.

The proposed amendments simply change the name of a licensee. The name change is purely administrative. None of the functions or responsibility of any of the Grand Gulf licensees will change as a result of the amendments. The proposed amendments do not alter the design, function, or operation of any plant equipment. As such, the accident and transient analyses contained in the facility updated final safety analysis report will not be affected.

Therefore, the proposed changes do 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 B. Glew, Jr., Associate General Counsel—Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601.

NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi

Date of amendment request: December 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16364A338.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) for GGNS. The amendment would allow for a one cycle extension to the 10-year frequency of the GGNS containment integrated leakage rate test (ILRT) or Type A test and the drywell bypass leak rate test (DWBT). These tests are required by TS 5.5.12, “10 CFR part 50, Appendix J [Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors], Testing Program,” and TS Surveillance Requirement 3.6.5.1.1, respectively. The proposed change would permit the existing ILRT and DWBT frequency to be extended from 10 years to 11.5 years.

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 edits in [brackets]:

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

Response: No.

The proposed amendment to the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years.

The proposed extension does 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. Type B and C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type B and C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The assessment of the [leak-tightness] of the drywell will continue to be performed at least once each operating cycle. The proposed amendment will not change the leakage rate acceptance requirements. As such, the containment will continue to perform its design function as a barrier to fission product releases. In addition, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment and the assessment of the [leak-tightness] of the drywell 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. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

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 Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years. 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 previously evaluated.

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

Response: No.

The proposed amendment to the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years. 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 10 CFR part 50, Appendix J, Testing Program for containment leak rate testing exist to ensure that the degree of containment structural integrity and leak-tightness 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 the extension of the interval for only the Type A containment leakage rate test and the drywell bypass leakage rate test for GGNS. The proposed surveillance interval extension is bounded by the 15-year Type A test interval currently authorized within NEI 94-01, Revision 3-A. 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 the proposed changes to the Type A test interval. In addition to the scheduled performance of DWBT GGNS will continue to monitor the drywell for significant leakage during operation.

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 B. Glew, Jr., Associate General Counsel—Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: April 10, 2017. A publicly-available version is available in ADAMS under Accession No. ML17100A844.

Description of amendment request: The amendment would revise the OCNGS Cyber Security Plan (CSP) Milestone 8 (MS8) full implementation completion date, as set forth in the CSP implementation schedule, and revise the physical protection license condition in the renewed facility operating license. The licensee proposes to revise the CSP MS8 completion date from December 31, 2017, to August 31, 2021.

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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review 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 amendment request proposes a change to the OCNGS CSP MS8 completion date as set forth in the CSP implementation schedule and associated regulatory commitments. The NRC staff has concluded that the proposed change: (1) 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; (2) 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 (3) has no impact on the probability or consequences of an accident previously evaluated. In addition, the NRC staff has concluded that the proposed change to the CSP implementation schedule is administrative in nature.

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 NRC staff has concluded the proposed change: (1) 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; and (2) 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. In addition, the NRC staff has concluded that the proposed change to the OCNGS CSP MS8 implementation schedule is administrative in nature.

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

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

Response: No.

Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The delay of the full implementation date for the OCNGS CSP MS8 has no substantive impact because other measures have been taken which provide adequate protection for the plant during this period of time. Therefore, the NRC staff has concluded that there is no significant reduction in a margin of safety. In addition, the NRC staff has concluded that the proposed change to the OCNGS CSP MS8 implementation schedule is administrative in nature.

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

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: Douglas A. Broaddus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment request: March 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17087A012.

Description of amendment request: The proposed changes would modify Technical Specifications (TS) Section 3.7.2, “Steam Generator Stop Valves (SGSVs),” to incorporate the SGSV actuator trains into the Limiting Condition for Operation and provide associated Conditions and Required Actions. In addition, Surveillance Requirement (SR) 3.7.2.2 would be revised to clearly identify that the SGSV actuator trains are required to be tested in accordance with the SR.

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 provide requirements for SGSVs that have dual actuators which receive signals from separate instrumentation trains. The design and functional performance requirements, operational characteristics, and reliability of the SGSVs and actuator trains are unchanged. There is no impact on the design safety function of the SGSVs to close (as an accident mitigator), nor is there any change with respect to inadvertent closure of an SGSV (as a potential transient initiator). Since no failure mode or initiating condition that could cause an accident (including any plant transient) is created or affected, the change cannot involve a significant increase in the probability of an accident previously evaluated.

With regard to the consequences of an accident and the equipment required for mitigation of the accident, the proposed changes involve no design or physical changes to the SGSVs or any other equipment required for accident mitigation. With respect to SGSV actuator train Completion Times, the consequences of an accident are independent of equipment Completion Times as long as adequate equipment availability is maintained. The proposed SGSV actuator Completion Times take into account the redundancy of the actuator trains and are limited in extent consistent with other Completion Times specified in the TS. Adequate equipment availability would therefore continue to be required by the TS. On this basis, the consequences of applicable, analyzed accidents are not significantly affected by the proposed changes.

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

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

Response: No.

The proposed changes to incorporate requirements for the SGSV actuator trains in TS 3.7.2 do not involve any design or physical changes to the facility, including the SGSVs and actuator trains themselves. No physical alteration of the plant is involved, as no new or different type of equipment is to be installed. The proposed changes do not alter any assumptions made in the safety analyses, nor do they involve any changes to plant procedures for ensuring that the plant is operated within analyzed limits. As such, no new failure modes or mechanisms that could cause a new or different kind of accident from any previously evaluated are being introduced.

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

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

Response: No.

The proposed changes to incorporate requirements for the SGSV actuator trains do not alter the manner in which safety limits or limiting safety system settings are determined. No changes to instrument/system actuation setpoints are involved. The safety analysis acceptance criteria are not affected by this change and the proposed changes will not permit plant operation in a configuration outside the design basis.

Therefore, 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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106.

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: March 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17086A442.

Description of amendment request: The proposed change would relocate cycle specific minimum critical power ratio (MCPR) values to the DAEC core operating limits report (COLR). The proposed amendment would revise the DAEC technical specifications (TS) to modify TS Table 3.3.2.1-1, “Control Rod Block Instrumentation,” Footnotes (a) through (e), and would relocate cycle specific MCPR values previously specified in TS Table 3.3.2.1-1, Footnotes (a) through (e) to TS 5.6.5(a)(4) by reference to the DAEC COLR.

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 is an administrative change that does not 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.

Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on, nor does it contribute in any way to the probability or consequences of transients or accidents. The COLR will continue to be controlled by the NextEra programs and procedures that comply with TS 5.6.5. Transient analyses addressed in the Final Safety Analysis Report will continue to be performed in the same manner with respect to changes in the cycle-dependent parameters obtained from the use of NRC-approved reload design methodologies, which ensures that the transient evaluation of new reloads are bounded by previously accepted analyses.

Therefore, the proposed TS 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 proposed administrative change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on new or different kind of accidents.

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 relocation of cycle-specific Control Rod Block Instrumentation MCPR values from the TS to the COLR. Appropriate measures exist to control the values of these cycle-specific 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.

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.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

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

Date of amendment request: March 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17093A688.

Description of amendment request: The amendments would revise technical specification requirements to operate ventilation systems with charcoal filters from 10 hours to 15 minutes in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month.”

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

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

Response: No.

The proposed change replaces an existing Surveillance Requirement to operate the CREMAFS [Control Room Emergency Makeup Air and Filtration System], FSBEACS [Fuel Storage Building Emergency Air Cleaning System], and SBVS [Shield Building Ventilation System] equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP [Surveillance Frequency Control Program] with a requirement to operate the systems for 15 continuous minutes with heaters operating.

These systems are not accident initiators and therefore, these changes do not involve a significant increase in the probability of an accident. The proposed system and filter testing changes are consistent with current regulatory guidance for these systems and will continue to assure that these systems perform their design function which may include mitigating accidents. Thus, the change does not involve a significant increase in the consequences of an accident.

Therefore, it is concluded that 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 replaces an existing Surveillance Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP with a requirement to operate the systems for 15 continuous minutes with heaters operating.

The change proposed for these ventilation systems does not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are capable of performing their intended safety functions. The change does not create new failure modes or mechanisms and no new accident precursors are generated.

Therefore, it is concluded that this 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 replaces an existing Surveillance Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with electric heaters for at least a continuous 10-hour period in accordance with the SFCP with a requirement to operate the systems for 15 continuous minutes with heaters operating.

The design basis for the ventilation systems' heaters is to heat the incoming air which reduces the relative humidity. The heater testing change proposed will continue to demonstrate that the heaters are capable of heating the air and will perform their design function. The proposed change is consistent with regulatory guidance.

Therefore, it is concluded that 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: William Blair, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: March 29, 2017. A publicly-available version is in ADAMS under Accession No. ML17094A565.

Brief description of amendment request: The proposed amendments would revise the current emergency action levels (EAL) scheme used at PINGP to the EAL scheme contained in NEI 99-01, Revision 6, “Development of Emergency Action Levels.”

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

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

Response: No.

The proposed change to the PINGP EAL scheme does not impact the physical function of plant structures, systems or components (SSC) or the manner in which the SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions. Therefore, the proposed change does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an event. The Emergency Plan, including the associated EALs, is implemented when an event occurs and cannot increase the probability of an accident. Further, the proposed change does not reduce the effectiveness of the Emergency Plan to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR part 50, Appendix E.

Therefore, the proposed EAL scheme 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 any physical alteration to the plant, that is, no new or different type of equipment will be installed. The proposed change also does not change the method of plant operation and does not alter assumptions made in the safety analysis. Therefore, the proposed change will not create new failure modes or mechanisms that could result in a new or different kind of accident. The emergency plan, including the associated EAL scheme, is implemented when an event occurs and is not an accident initiator.

Therefore, the proposed EAL scheme 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.

Margin of safety is provided by the ability of accident mitigation SCCs to perform at their analyzed capability. The change proposed in this license amendment request does not modify any plant equipment and there is no impact to the capability of the equipment to perform its intended accident mitigation function. The proposed change does not impact operation of the plant or its response to transients or accidents. Additionally, the proposed changes will not change any criteria used to establish safety limits or any safety system settings. The applicable requirements of 10 CFR 50.47 and 10 CFR part 50, Appendix E will continue to be met.

Therefore, the proposed EAL scheme 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 requests involve no significant hazards consideration.

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: March 27, 2017, as supplemented by letter dated April 28, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17086A364 and ML17118A092, respectively.

Description of amendment request: The amendment would amend the Hope Creek Generating Station (Hope Creek) Technical Specifications (TSs) to revise and relocate the pressure-temperature (P-T) limit curves to a licensee-controlled pressure and temperature limits report (PTLR). The request was submitted in accordance with guidance provided in NRC Generic Letter 96-03, “Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protections System Limits,” dated January 31, 1996.

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 license amendment adopts the NRC approved methodology described in Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044), “Pressure Temperature Limits Report Methodology for Boiling Water Reactors.” The Hope Creek PTLR was developed based on the methodology and template provided in the BWROG LTR.

10 CFR part 50, Appendix G establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants.

Implementing this NRC approved methodology does not reduce the ability to protect the RCPB as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

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

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

Response: No.

The change in methodology for calculating P-T limits and the relocation of those limits to the PTLR do not alter or involve any design basis accident initiators. RCPB integrity will continue to be maintained in accordance with 10 CFR part 50, Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner.

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 changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not affect the function of the RCPB or its response during plant transients. Calculating the Hope Creek P-T limits using the NRC approved SI methodology ensures adequate margins of safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation P-T limits for the RCPB are determined. There are no changes to the setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

Therefore, it is concluded that 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: Jeffrie J. Keenan, PSEG Nuclear LLC—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: April 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17102B032.

Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix C (and plant-specific Tier 1) and Updated Final Safety Analysis Report (UFSAR) Tier 2 that describe: (1) The inspection and analysis of, and specifies the maximum calculated flow resistance acceptance criteria for, the fourth-stage automatic depressurization system loops; (2) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the allowable flow resistance acceptance criteria for, the in-containment refueling water storage tank (IRWST) injection line; (3) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the maximum flow resistance acceptance criteria for, the containment recirculation line; (4) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that specifies acceptance criteria for the maximum flow resistance between the IRWST drain line and the containment; and 5) removes licensing basis text from UFSAR Tier 2 that discusses the operation of swing check valves in current operating plants.

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 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 do not adversely affect the physical design and operation of the in-containment refueling water storage tank (IRWST) injection, drain, containment recirculation, or fourth-stage automatic depressurization system (ADS) valves, including as-installed inspections and maintenance requirements as described in the Updated Final Safety Analysis Report (UFSAR). Inadvertent operation or failure of the fourth-stage ADS valves are considered as an accident initiator or part of an initiating sequence of events for an accident previously evaluated. However, the proposed change to the test methodology and calculated flow resistance for the fourth-stage ADS lines does not adversely affect the probability of inadvertent operation or failure. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected.

The proposed changes do not adversely affect the ability of IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The designs of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria.

The proposed changes do not adversely 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 might 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 adversely affect the physical design and operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves, including as-installed inspections, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves is not adversely affected. These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.

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

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

Response: No.

The proposed changes maintain existing safety margins. The proposed changes verify and maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while updating the acceptance criteria for verifying the design features necessary to ensure the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves perform the design functions required to meet the existing safety margins in the safety analyses. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR.

These changes do not adversely affect any design code function, design analysis, safety analysis input or result, or design/safety margin.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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 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.

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

Date of amendment request: April 1, 2016, as supplemented by letters dated July 21, September 9, and October 26, 2016.

Description of amendment request: The amendments revised the Technical Specifications (TSs) for PVNGS, by modifying the requirements regarding the degraded and loss of voltage relays that are planned to be modified to be more aligned with designs generally implemented in the industry. Specifically, the licensing basis for degraded voltage protection will be changed from reliance on a TS initial condition that ensures adequate post-trip voltage support of accident mitigation equipment to crediting automatic actuation of the degraded and loss of voltage relays to ensure proper equipment performance.

Date of issuance: April 27, 2017.

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

Amendment Nos.: Unit 1-201, Unit 2-201, and Unit 3-201. A publicly-available version is in ADAMS under Accession No. ML17090A164; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register : May 24, 2016 (81 FR 32803). The supplements dated July 21, September 9, and October 26, 2016, 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 April 27, 2017.

No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324; Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina

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

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

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

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

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

Date of amendment request: June 23, 2016.

Brief description of amendments: The amendments modified the technical specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9 to the TS for BSEP, ONS, and RNP. The same changes were added as LCO 3.0.10 to the TS for CNS and MNS. For HNP, TS requirements for unavailable barriers were modified by adding LCO 3.0.6 to the TS. The changes are consistent with Technical Specification Task Force Traveler (TSTF)-427, Revision 2, “Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY,” subject to stated variations.

Date of issuance: April 26, 2017.

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

Amendment Nos: 274/302 (BSEP), 288/284 (CNS), 155 (HNP), 295/274 (MNS), 402/404/403 (ONS), and 251 (RNP). A publicly-available version is in ADAMS under Accession No. ML17066A374; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-71 and DPR-62 (BSEP), NPF-35 and NPF-52 (CNS), NPF-63 (HNP), NPF-9 and NPF-17 (MNS), DPR-38, DPR-47, DPR-55 (ONS), and DPR-23 (RNP): Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : August 16, 2016 (81 FR 54614).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 26, 2017.

No significant hazards consideration comments received: No.

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

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

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

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

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

Date of amendment request: September 27, 2016, as supplemented by letters dated November 22, 2016, and April 20, 2017.

Description of amendment request: The amendments revised Technical Specification Surveillance Requirements to require operating ventilation systems with charcoal filters for 15 continuous minutes every 31 days or at a frequency controlled in accordance with the Surveillance Frequency Control Program. The amendments are consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month,” as published in the Federal Register on September 20, 2012 (77 FR 58428), with variations due to plant-specific differences.

Date of issuance: May 8, 2017.

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

Amendment Nos.: 275 (Unit 1) and 303 (Unit 2) for the Brunswick Steam Electric Plant; 289 (Unit 1) and 285 (Unit 2) for the Catawba Nuclear Station; 296 (Unit 1) and 275 (Unit 2) for the McGuire Nuclear Station; 156 (Unit 1) for the Shearon Harris Nuclear Power Plant; and 252 (Unit No. 2) for the H. B. Robinson Steam Electric Plant. A publicly-available version is in ADAMS under Accession No. ML17055A647; documents related to these amendments are listed in the Safety Evaluations enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-71 and DPR-62, for the Brunswick Steam Electric Plant, Units 1 and 2; NPF-35 and NPF-52, for the Catawba Nuclear Station, Units 1 and 2; NPF-9 and NPF-17, for the McGuire Nuclear Station, Units 1 and 2; NPF-63, for the Shearon Harris Nuclear Power Plant, Unit 1; and DPR-23, for the H. B. Robinson Steam Electric Plant, Unit No. 2: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : January 17, 2017 (82 FR 4929). The supplemental letter dated April 20, 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 evaluations of the amendments are contained in Safety Evaluations dated May 8, 2017.

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: February 10, 2016, as supplemented by letters dated October 10 and December 16, 2016; and January 31, February 7, February 16, and March 29, 2017.

Brief description of amendment: The amendment revised Technical Specification 5.5.11, “Primary Containment Leakage Rate Testing Program,” to increase the containment integrated leakage rate test program Test A interval from 10 to 15 years.

Date of issuance: April 25, 2017.

Effective date: As of the date of issuance and shall be implemented prior to the startup from the 2017 refueling outage.

Amendment No.: 193. A publicly-available version is in ADAMS under Accession No. ML17103A235; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-22: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : April 26, 2016 (81 FR 24663). The supplemental letters dated October 10 and December 16, 2016; and January 31, February 7, February 16, and March 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 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 April 25, 2017.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo County, California

Date of application for amendments: June 17, 2015, as supplemented by letters dated August 31, October 22, November 2, November 6, and December 17, 2015; and February 1, February 10, April 21, June 9, September 15, October 6, and December 27, 2016.

Brief description of amendments: The amendments revised the licensing bases to adopt the alternative source term (AST) as allowed by 10 CFR 50.67, “Accident source term.” The AST methodology, as established in NRC Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” July 2000 (ADAMS Accession No. ML003716792), is used to calculate the offsite and control room radiological consequences of postulated accidents for DCPP, Units 1 and 2. The amendments revised Technical Specification (TS) 1.1, “Definitions,” for the definition of Dose Equivalent I-131; TS 3.4.16, “RCS [Reactor Coolant System] Specific Activity,” to revise the noble gas activity limit; TS 3.6.3, “Containment Isolation Valves,” to require the 48-inch containment purge supply and exhaust valves to be sealed closed during Modes 1, 2, 3, and 4; TS 5.5.11, “Ventilation Filter Testing Program (VFTP),” to change the allowable methyl iodide penetration testing criteria for the auxiliary building system charcoal filter; TS 5.5.19, “Control Room Habitability Program,” to replace “whole body or its equivalent to any part of the body,” with “Total Effective Dose Equivalent,” which is the dose criteria specified in 10 CFR 50.67, and Appendix D, “Additional Conditions,” for Facility Operating License Nos. DPR-80 and DPR-82 for DCPP, Units 1 and 2, to add additional license conditions.

Date of issuance: April 27, 2017.

Effective date: As of its date of issuance and shall be implemented within 365 days from the date of issuance.

Amendment Nos.: Unit 1-230; Unit 2-232. A publicly-available version is in ADAMS under Accession No. ML17012A246; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : The license amendment request was originally noticed in the Federal Register on October 13, 2015 (80 FR 61486). As a result of the supplemental letters dated October 22, November 2, November 6, and December 17, 2015; and February 1, February 10, April 21, June 9, and September 15, 2016, the notice was reissued in its entirety to include the revised scope, description of the amendment request, and proposed no significant hazards consideration determination on November 8, 2016 (81 FR 78664).

The supplemental letters dated October 6 and December 27, 2016, 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 April 27, 2017.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: June 7, 2016.

Brief description of amendment: The amendment modified the Technical Specifications to allow the use of Component Cooling System (CCS) pump 2B-B to support Train 1B operability when the normally aligned CCS pump C-S is removed from service.

Date of issuance: April 27, 2017.

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

Amendment No.: 113. A publicly-available version is in ADAMS under Accession No. ML17081A263; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-90: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : September 13, 2016 (81 FR 62932).

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

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 11th day of May 2017.

For the Nuclear Regulatory Commission.

Kathryn M. Brock,

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