2017-02795. 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 January 14 to January 30, 2017. The last biweekly notice was published on January 31, 2017.

DATES:

Comments must be filed by March 16, 2017. A request for a hearing must be filed by April 17, 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-2017-0038. 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: OWFN-12-H08, 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-2017-0038, 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-2017-0038.
• 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-2017-0038, 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 posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

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

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

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

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

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

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

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

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

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

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

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

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

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by April 17, 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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

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

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

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

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

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

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

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

Date of amendment request: July 20, 2016. Publicly-available version is in ADAMS under Accession No. ML16209A222.

Description of amendment request: The proposed amendment requests to revise the Technical Specifications (TSs) associated with dry spent fuel storage cask loading and unloading requirements for ONS. Specifically, the license amendment request would revise TS 3.7.12, “Spent Fuel Pool Boron Concentration”; TS 3.7.18, “Dry Spent Fuel Storage Cask Loading and Unloading”; and TS 4.4, “Dry Spent Fuel Storage Cask Loading and Unloading,” to remove certain TS requirements that no longer pertain to the ONS Independent Spent Fuel Storage Facility general license, due to changes in 10 CFR 50.68, “Criticality accident requirements.”

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 square brackets:

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

Response: No.

The proposed changes to Technical Specifications (TSs) 3.7.12, 3.7.18 and 4.4, do not modify the method of nuclear fuel storage or handling at Oconee Nuclear Station (ONS), or make any physical changes to the facility design, material, or construction standards. The proposed change revises the criticality requirements contained in the TSs, as allowed by 10 CFR 50.68(c), that are redundant to regulatory requirements provided in 10 CFR part 72 and the Nuclear Regulatory Commission (NRC)-approved Certificate of Compliance (CoC) for the spent fuel dry shielded canisters utilized at ONS. The proposed change to the TS requirements neither result[s] in operation that will increase the probability of initiating an analyzed event nor alter[s] assumptions relative to mitigation of an accident or transient event. The change has no effect on the process variables, structures, systems, and components that must be maintained consistent with the safety analyses and licensing basis.

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

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

Response: No.

The proposed changes to TSs 3.7.12, 3.7.18 and 4.4, do not modify the method of nuclear fuel storage or handling at ONS, nor make any physical changes to the facility design, material, or construction standards. The change does not alter the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The proposed change to the ONS TS requirements does not adversely impact the results of the ONS safety analyses and is compliant with the current licensing basis.

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

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

Response: No.

The proposed changes to TS 3.7.12, 3.7.18 and 4.4, do not modify the method of nuclear fuel storage or handling at ONS, nor make any physical changes to the facility design, material, or construction standards. The proposed changes comply with NRC approved regulations and the station's Part 72 and 50 licensing basis.

Therefore, the proposed TS 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: Lara S. Nichols, Vice President Nuclear & EHS Legal Support, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202-1802.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: July 21, 2016. Publicly-available version is in ADAMS under Accession No. ML16209A223.

Description of amendment request: The proposed amendment requests to revise the Technical Specifications (TSs) for ONS. Specifically, the license amendment request would revise TS 2.1.1.1, “Reactor Core Safety Limits (SLs),” and TS 5.6.5, “Core Operating Limits Report (COLR),” to allow the use of the COPERNIC fuel performance code.

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 square brackets:

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

Response: No.

The proposed change adds a limit on maximum local fuel pin centerline temperature to [the] ONS Technical Specifications that is based on a[n] NRC reviewed and approved fuel performance code, and does not require a physical change to plant systems, structures or components. Plant operations and analysis will continue to be in accordance with the ONS licensing basis. The peak fuel centerline temperature is the basis for protecting the fuel and is consistent with the safety analysis.

The proposed change also adds a topical report for a[n] NRC reviewed and approved fuel performance code to the list of topical reports in [the] ONS Technical Specifications, which is administrative in nature and has no impact on a plant configuration or system performance relied upon to mitigate the consequences of an accident. The list of topical reports in the Technical Specifications used to develop the core operating limits does not impact either the initiation of an accident or the mitigation of its consequences.

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

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

Response: No.

The proposed change adds a limit on maximum local fuel pin centerline temperature to [the] ONS Technical Specifications that is based on a[n] NRC reviewed and approved fuel performance code, and does not require a physical change to plant systems, structures or components. Specifying maximum local fuel pin centerline temperature ensures that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The addition of a new fuel pin centerline melt temperature versus burnup relationship does not affect any accident initiators that would create a new accident.

The proposed change also adds a topical report for a[n] NRC reviewed and approved fuel performance code to the list of topical reports in [the] ONS Technical Specifications, which is administrative in nature and has no impact on a plant configuration or on system performance. The proposed change updates the list of NRC-approved topical reports used to develop the core operating limits. There is no change to the parameters within which the plant is normally operated. The possibility of a new or different kind of accident is not created.

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

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

Response: No.

The proposed change to TS 2.1.1.1 adds a limit on maximum local fuel pin centerline temperature that is based on an NRC reviewed and approved fuel performance code, and does not require a physical change to plant systems, structures or components. Plant operations and analysis will continue to be in accordance with [the] ONS licensing basis.

Adding the local fuel pin centerline temperature and burnup relationship defined by the NRC reviewed and approved fuel performance code to the ONS Technical Specifications, does not impact the safety margins established in the ONS licensing basis.

The proposed change also adds a topical report for a[n] NRC reviewed and approved fuel performance code to the list of topical reports in [the] ONS Technical Specifications, which is administrative in nature and does not amend the cycle specific parameters presently required by the Technical Specifications. The individual Technical Specifications continue to require operation of the plant within the bounds of the limits specified in the Core Operating Limits Report. The proposed change to the list of analytical methods referenced in the Core Operating Limits Report does not impact the margin of safety.

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

The NRC staff has reviewed the 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: Lara S. Nichols, Vice President Nuclear & EHS Legal Support, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202-1802.

NRC Branch Chief: Michael T. Markley.

Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (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, LLC, 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, LLC, Docket No. 50-261, H.B. Robinson Steam Electric Plant, Unit No. 2 (RNP), Darlington County, South Carolina

Duke Energy Carolinas, LLC, Docket Nos. 52-018 and 52-019, William States Lee III Nuclear Station, Units 1, and 2 (WLS), Cherokee County, South Carolina

Date of amendment request: April 29, 2016, as supplemented by letters dated October 3, 2016, and January 16, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML16120A076, ML16277A521, and ML17017A210, respectively.

Description of amendment request: The NRC staff previously made a proposed determination that the amendment request dated April 29, 2016, involves no significant hazards considerations (81 FR 43650; July 5, 2016). Subsequently, by letter dated January 16, 2017, the licensee provided additional information that expanded the scope of the amendment request as originally noticed. Accordingly, this notice supersedes the previous notice in its entirety.

The amendments would (1) consolidate the Emergency Operations Facilities (EOFs) for BSEP, HNP, and RNP with the Duke Energy Progress, LLC (Duke Energy) corporate EOF in Charlotte, North Carolina; (2) decrease the frequency for a multi-site drill from once per 6 years to once per 8 years; (3) allow the multi-site drill performance with sites other than CNS, MNS, or ONS, (4) change the BSEP, HNP, and RNP augmentation times to be consistent with those of the sites currently supported by the Duke Energy corporate EOF, and (5) decrease the frequency of the unannounced augmentation drill at BSEP from twice per year to once per year. The January 16, 2017, letter also acknowledges the addition of WLS to the Duke Energy corporate EOF with the issuance of the WLS operating license on December 19, 2016 (ADAMS Accession No. ML16333A329).

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 relocate the BSEP, HNP, and RNP EOFs from their present onsite or near-site locations to the established corporate EOF in Charlotte, North Carolina, changes the required response times for supplementing onsite personnel in response to a radiological emergency, decreases the multi-site drill frequency, allows the multi-site drill to be performed with sites other than ONS, MNS, or CNS, and decreases the frequency of augmentation drills at BSEP. The functions and capabilities of the relocated EOFs will continue to meet the applicable regulatory requirements. It has been evaluated and determined that the change in response time does not significantly affect the ability to supplement the onsite staff. In addition, analysis shows that the onsite staff can acceptably respond to an event for longer than the requested time for augmented staff to arrive. The proposed changes have no effect on normal plant operation or on any accident initiator or precursors, and do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not alter or prevent the ability of the emergency response organization to perform its intended functions to mitigate the consequences of an accident or event.

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

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

Response: No.

The proposed changes only impact the implementation of the affected stations' emergency plans by relocating their onsite or near-site EOFs to the established corporate EOF in Charlotte, North Carolina, changing the required response time of responders who supplement the onsite staff, decreasing the multi-site drill frequency, allowing the multi-site drill to be performed with sites other than ONS, MNS, or CNS, and decreasing the frequency of augmentation drills at BSEP. The functions and capabilities of the relocated EOFs will continue to meet the applicable regulatory requirements. It has been evaluated and determined that the change in response time does not significantly affect the ability to supplement the onsite staff. In addition, analysis shows that the onsite staff can acceptably respond to an event for longer than the requested time for augmented staff to arrive. The proposed changes will not change the design function or operation of SSCs. The changes do not impact the accident analysis. The changes do not involve a physical alteration of the plant, a change in the method of plant operation, or new operator actions. The proposed changes do not introduce failure modes that could result in a new accident, and the changes do not alter assumptions made in the safety analysis.

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

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

Response: No.

The proposed changes only impacts the implementation of the affected stations' emergency plans by relocating their onsite or near-site EOFs to the established corporate EOF in Charlotte, North Carolina, changing the required response time of responders who supplement the onsite staff, decreasing the multi-site drill frequency, allowing the multi-site drill to be performed with sites other than ONS, MNS, or CNS, and decreasing the frequency of augmentation drills at BSEP. The functions and capabilities of the relocated EOFs will continue to meet the applicable regulatory requirements. It has been evaluated and determined that the change in response time does not significantly affect the ability to supplement the onsite staff. In addition, analysis shows that the onsite staff can acceptably respond to an event for longer than the requested time for augmented staff to arrive. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the emergency plans and do not impact operation of the plant or its response to transients or accidents. The changes do not affect the Technical Specifications. The changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected. The emergency plans will continue to provide the necessary response staff for emergencies as demonstrated by staffing and functional analyses including the necessary timeliness of performing major tasks for the functional areas of the emergency plans.

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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: December 2, 2016. A publicly-available version is in ADAMS under Accession No. ML16337A249.

Description of amendment request: The amendment would revise HNP Technical Specifications (TSs) to relocate selected figures and values from the TSs to the Core Operating Limits Report (COLR), remove all references to a specific plant procedure as it pertains to the COLR, and adopt Technical Specification Task Force (TSTF)-5, “Delete Safety Limit Violation Notification Requirements,” Revision 1, which deletes duplicate notification, reporting and restart requirements from the Administrative Controls section of TSs.

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

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

[Response: No.]

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative in nature, facilitate improved content and presentation of Administrative controls, and alter only the format and location of cycle-specific parameter figures and limits from the TS to the COLR. This relocation does not result in the alteration of the design, material, or construction standards that were applicable prior to the change. The proposed changes will not result in modification of any system interface that would increase the likelihood of an accident since these events are independent of the proposed change. The proposed amendment will not change, degrade, or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the Final Safety Analysis Report (FSAR).

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 previously evaluated?

[Response: No.]

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there will be no change in types or increase in the amounts of any effluents released offsite.

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

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

[Response: No.]

The proposed changes do not involve a significant reduction in a margin of safety. Previously-approved methodologies will continue to be used in determination of cycle-specific core operating limits that are present in the COLR. The proposed changes are administrative in nature and will not affect the plant design or system operating parameters. As such, there is no detrimental impact on any equipment design parameter and the plant will continue to be operated within prescribed limits.

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: Lara Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: November 8, 2016. A publicly-available version is in ADAMS under Accession No. ML16313A573.

Description of amendment request: The proposed amendment would, on a one-time basis, extend the Completion Time by 7 days for Technical Specification Conditions 3.5.1.A, 3.6.1.5.A, and 3.6.2.3.A. This onetime extension will be used to support preventive maintenance, which replaces the residual heat removal train A subsystem's pump and motor.

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

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

Response: No.

The proposed amendment does not increase the probability of an accident because the residual heat removal (RHR) system cannot initiate an accident. The RHR system provides coolant injection to the reactor core, cooling of the suppression pool water inventory, and drywell sprays following a design basis accident.

The proposed one time completion time (CT) change for RHR train A does not alter the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis for accident mitigation. No changes are proposed in the manner in which the emergency core cooling system (ECCS) provides plant protection or which create new modes of plant operation. In addition, a probabilistic safety assessment (PSA) evaluation concluded that the risk contribution of the increased CT is a very small increase in risk. The proposed change in CT will not affect the probability of any event initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance.

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 previously evaluated?

Response: No.

The proposed amendment will not create the possibility of a new or different kind of accident because inoperability of one RHR subsystem is not an accident precursor. There are no hardware changes nor are there any changes in the method by which any plant system performs a safety function. This request does not affect the normal method of plant operation. The proposed amendment does not introduce new equipment, or new way of operation of the system which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this request.

Therefore, the implementation of the proposed amendment will not create a possibility for an accident of a new or different type than those previously evaluated.

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

Response: No.

Columbia's ECCS is designed with sufficient redundancy such that a low pressure ECCS subsystem may be removed from service for maintenance or testing. The remaining subsystems are capable of providing water and removing heat loads to satisfy the final safety analysis report (FSAR) requirements for accident mitigation or plant shutdown. A PSA evaluation concluded that the risk contribution of the CT extension is very small. There will be no change to the manner in which safety limits or limiting safety system settings are determined nor will there be any change to those plant systems necessary to assure the accomplishment of protection functions. There will be no change to post-LOCA peak clad temperatures.

For these reasons, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: November 9, 2016. A publicly-available version is in ADAMS under Accession No. ML16314A027.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 5.5.10, “Ventilation Filter Testing Program,” to correct and modify the description of the control room ventilation and fuel handling area ventilation systems. In addition, the proposed amendment would correct an editorial omission in TS Limiting Condition for Operation 3.0.9.

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 Palisades Nuclear Plant (PNP) Technical Specifications (TS) are editorial or administrative in nature. The changes make an editorial correction in the TS, and correct and modify the component descriptions in the ventilation filter testing program TS. These changes do 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 changes do not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents, and have no impact on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed changes to the PNP TS are editorial or administrative in nature. The changes make an editorial correction in the TS, and correct and modify the component descriptions within the ventilation filter testing program TS. The proposed changes do 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 changes do 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 do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

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 TS are editorial or administrative in nature and do not impact any safety margins. Because there is no impact on established safety margins as a result of these changes, the proposed change does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: October 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16300A200.

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

Additionally, this license amendment request (LAR) proposes to delete Condition 2.D.(e) of the LSCS Unit 1 Renewed Facility Operating License regarding conducting the third Type A Test of each 10-year service period when the plant is shutdown for the 10-year plant inservice inspection (ISI). Similarly, this LAR proposes to delete Condition 2.D.(c) of the LSCS Unit 2 Renewed Facility Operating License regarding conducting the third Type A test of each 10-year service period when the plant is shutdown for the 10 year plant ISI.

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

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

Response: No.

The proposed amendment to the TS involves the extension of the LSCS Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

The proposed 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. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured as an increase to the total integrated dose risk for all internal events accident sequences for LSCS, is 1.23E-02 person-rem/yr (0.33%) using the EPRI [Electric Power Research Institute] guidance with the base case corrosion included. The change in dose risk drops to 3.15E-03 person-rem/yr (0.08%) when using the EPRI Expert Elicitation methodology. The values calculated per the EPRI guidance are all lower than the acceptance criteria of ≤1.0 person-rem/yr or <1.0% person-rem/yr defined in Section 1.3 of Attachment 3 of this submittal. The results of the risk assessment for this amendment meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The LSCS Type A test history supports this conclusion.

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

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

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

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

Response: No.

The proposed amendment to the TS involves the extension of the LSCS Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

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

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

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

Response: No.

The proposed amendment to TS 5.5.13 involves the extension of the LSCS Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and 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 only the extension of the interval between Type A containment leak rate tests and Type C tests for LSCS. The proposed surveillance interval extension is bounded by the 15-year ILRT Interval and the 75-month Type C test interval currently authorized within NEI [Nuclear Energy Institute] 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for LSCS. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any 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 requested amendments involve no significant hazards consideration.

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

NRC Branch Chief: David J. Wrona.

South Carolina Electric & Gas Company (SCE&G) and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: November 10, 2016. A publicly-available version is in ADAMS under Accession No. ML16316A003.

Description of amendment request: The amendment request proposes to add to License Condition 2.D.(1) of the VCSNS Units 2 and 3 combined licenses (COLs), an Interim Amendment Request process for changes during construction when emergent conditions are present.

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

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

Response: No.

The proposed amendment would add an Interim Amendment Request process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow construction to continue, at SCE&G's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SCE&G to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue a determination that construction could continue pending SCE&G's initiation of the COL-ISG-025 PAR [preliminary amendment request]/LAR [license amendment request] process. The requirement to include a Nuclear Construction Safety Assessment ensures that the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. If the continued construction would result a material decrease in safety, then continued construction would not be authorized.

The proposed amendment does not modify the design, construction, or operation of any plant structures, systems, or components (SSCs), nor does it change any procedures or method of control for any SSCs. Because the proposed amendment does not change the design, construction, or operation of any SSCs, it does not adversely affect any design function as described in the Updated Final Safety Analysis Report.

The proposed amendment does not affect the probability of an accident previously evaluated. Similarly, because the proposed amendment does not alter the design or operation of the nuclear plant or any plant SSCs, the proposed amendment does not represent a change to the radiological effects of an accident, and therefore, does not involve an increase in the consequences of an accident previously evaluated.

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

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

Response: No.

The proposed amendment would add an Interim Amendment Request process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow construction to continue, at SCE&G's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SCE&G to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would NRC issue a determination that construction could continue pending SCE&G's initiation of the COL-ISG-025 PAR/LAR process.

The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment only adds a new screening process and does not change the design, construction, or operation of the nuclear plant or any plant operations.

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 amendment would add an Interim Amendment Request process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow construction to continue, at SCE&G's own risk, in emergent conditions, where a non-conforming condition that has little or no safety significance is discovered and the work activity cannot be adjusted. The Interim Amendment Request process would require SCE&G to submit a Nuclear Construction Safety Assessment which (1) identifies the proposed change; (2) evaluates whether emergent conditions are present; (3) evaluates whether the change would result in any material decrease in safety; and (4) evaluates whether continued construction would make the non-conforming condition irreversible. Only if the continued construction would have no material decrease in safety would the NRC issue determination that construction could continue pending SCE&G's initiation of the COL-ISG-025 PAR/LAR process.

The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The proposed amendment does not alter any design function or safety analysis. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed amendment, thus the margin of safety is not reduced. The only impact of this activity is the addition of an Interim Amendment Request process.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: September 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16259A315.

Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information and a combined license (COL) License Condition which references one of the proposed changes. Additionally, the proposed changes to the UFSAR eliminate pressurizer spray line monitoring during pressurizer surge line first plant only testing. In addition, these proposed changes correct inconsistencies in testing purpose, testing duration, and the ability to leave equipment in place following the data collection period. These changes involve material which is specifically referenced in Section 2.D.(2) of the COL. This submittal requests approval of the license amendment necessary to implement these changes.

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

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

Response: No.

The design functions of the RCS [reactor coolant system] include providing an effective reactor coolant pressure boundary. The proposed changes for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement do not impact the existing design requirements for the RCS. These changes are acceptable as they are consistent with the commitments made for the pressurizer surge line monitoring program for the first plant only, and do not adversely affect the capability of the pressurizer surge line and pressurizer spray lines to perform the required reactor coolant pressure boundary design functions.

These proposed changes to the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis do not have an adverse effect on any of the design functions of the systems. The proposed changes do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors.

Therefore, the requested 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 for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis maintain the required design functions, and are consistent with other Updated Final Safety Analysis Report (UFSAR) information. The proposed changes do not adversely affect the design requirements for the RCS, including the pressurizer surge line and pressurizer spray lines. The proposed changes do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes.

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

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

Response: No.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: November 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16323A020.

Description of amendment request: The amendment request proposes changes to plant-specific Tier 1 information, with corresponding changes to the associated Combined License (COL) Appendix C information, and involves associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR). Specifically, the requested amendment proposes clarifications to a plant-specific Tier 1 (and COL Appendix C) table and a UFSAR table in regard to the inspections of the excore source, intermediate, and power range detectors. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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

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

Response: No.

The proposed change to specify the inspection of the excore source, intermediate, and power range detectors is done to verify that aluminum surfaces are contained in stainless steel or titanium, and avoids the introduction of aluminum into the post-loss of coolant accident (LOCA) containment environment due to detector materials. The proposed change does not alter any safety related functions. The materials of construction are compatible with the post-LOCA conditions inside containment and will not significantly contribute to hydrogen generation or chemical precipitates. The change does not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events.

The change does not impact the support, design, or operation of mechanical and fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions. Consequently, the plant response to previously evaluated accidents or external events is not adversely affected, nor does the proposed change create any new accident precursors.

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 change does not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed change to specify the inspection of the excore source, intermediate, and power range detectors is done to verify that aluminum surfaces are contained in stainless steel or titanium, and avoids the introduction of aluminum into the post-LOCA containment environment due to detector materials. In addition, the proposed change to the ITAAC [inspections, tests, analysis, and acceptance criteria] verified materials of construction does not alter the design function of the excore detectors. The detector canning materials of construction are compatible with the post-LOCA containment environment and do not contribute a significant amount of hydrogen or chemical precipitates. The change to the ITAAC aligns the inspection with the Tier 2 design feature. Consequently, because the excore detectors functions are unchanged, there are no adverse effects on accidents previously evaluated in the UFSAR.

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 change to specify the inspection of the excore source, intermediate, and power range detectors is done to verify that aluminum surfaces are contained in stainless steel or titanium, and avoids the introduction of aluminum into the post-LOCA containment environment, does not alter any safety-related equipment, applicable design codes, code compliance, design function, or safety analysis. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change, thus the margin of safety is not reduced.

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

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

Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 1, 2, and 3, San Diego County, California

Date of amendment request: December 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16355A014.

Description of amendment request: The proposed amendment would revise the Facility Operating Licenses and associated Technical Specifications (TS) for SONGS, Units 1, 2, and 3, to reflect removal of all spent nuclear fuel from the SONGS, Units 2 and 3 spent fuel pools (SFPs) and its transfer to dry cask storage within an onsite independent spent fuel storage installation (ISFSI). The proposed changes would also make conforming changes to the SONGS, Unit 1 TS and combine them with the SONGS, Units 2 and 3 TS. These changes will more fully reflect the permanently shutdown and defueled status of the facility, as well as the reduced scope of structures, systems, and components necessary to ensure plant safety once all spent fuel has been permanently moved to the SONGS ISFSI, an activity which is currently scheduled for completion in 2019.

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

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

Response: No.

The proposed amendment would modify the SONGS, Units 2 and 3 facility operating licenses and TS by deleting the portions of the licenses and TSs that are no longer applicable to a facility with no spent nuclear fuel stored in the SFP, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment becomes effective upon removal of all spent nuclear fuel from the SONGS, Units 2 and 3 SFP and its transfer to dry cask storage within an ISFSI.

Additionally, the proposed change would revise the Unit 1 TSs for consistency with the proposed changes to the Units 2 and 3 TSs. Similar to the changes for Units 2 and 3, the Unit 1 changes delete portions of the TSs that are no longer applicable to a facility with spent fuel no longer stored in the SFP, while modifying the remaining portions to correspond to all nuclear fuel in dry storage. The Unit 1 TSs are also proposed to be combined with the Units 2 and 3 TSs.

The definition of safety-related Structures, Systems, and Components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shutdown the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.43(a)(1) or 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI, none of the SSCs at SONGS, Units 2 and 3 are required to be relied on for accident mitigation. Therefore, none of the SSCs at SONGS, Units 2 and 3 meet the definition of a safety-related SSC stated in 10 CFR 50.2. The proposed deletion of requirements in the TSs is not related to any systems credited in an accident analysis at SONGS, Units 2 and 3.

Chapter 15 of the SONGS, Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) described the design basis accidents (DBAs) related to the SFP. The majority of these postulated accidents are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Manager, or a Certified Operator to prevent occurrence or mitigate the consequences of an accident.

With all of the SONGS 1 operating plant above-ground structures having been demolished and removed, and all Unit 1 spent fuel having been removed from the SFP, there are no remaining design basis accidents or transients in Chapter 8 of the Unit 1 Defueled Safety Analysis Report (DSAR).

The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their potential consequences.

The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility.

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 eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the SFP, and relocate certain administrative controls to the Decommissioning Quality Assurance Program or Licensee Controlled Specifications (LCS).

After the removal of the spent fuel from the Units 2 and 3 SFP and transfer to the ISFSI, there are no spent fuel assemblies that remain in a SFP on site. Coupled with a prohibition against storage of fuel in the Units 2 and 3 SFP (the Unit 1 SFP has been dismantled), the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes.

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

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

Response: No.

The removal of all spent nuclear fuel from the SFPs into storage in casks within an ISFSI, coupled with a prohibition against future storage of fuel within the Units 2 and 3 SFPs (the Unit 1 SFP has been dismantled), removes the potential for fuel related accidents.

The design basis and accident assumptions within the SONGS, Units 1, 2 and 3 UFSARs and the TSs relating to safe management and safe storage of spent fuel in the SFP are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities.

The proposed deletion of TS requirements is not related to any SSCs that will be credited in the accident analysis for an applicable postulated accident. As a result, the proposed deletions do not affect the margin of safety associated with the accident analysis.

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

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

Attorney for licensee: Walker A. Matthews, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Bruce Watson.

Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 1, 2, and 3, San Diego County, California

Date of amendment request: December 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16355A015.

Description of amendment request: The proposed amendment would revise the Permanently Defueled Emergency Plan into an Independent Spent Fuel Storage Facility Installation (ISFSl)-Only Emergency Plan, and revise the Emergency Action Level (EAL) scheme into an ISFSl-Only EAL scheme, for SONGS, Units 1, 2, and 3. The proposed changes would more fully reflect the permanently shutdown and defueled status of the facility, as well as the reduced scope of potential radiological accidents once all spent fuel has been moved to dry cask storage within the onsite SONGS ISFSI, an activity which is currently scheduled for completion in 2019.

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 amendments would modify the SONGS, Units 1, 2 and 3 licenses by revising the emergency plan and revising the EAL scheme. The SONGS units have permanently ceased operation and are permanently defueled. The proposed amendments are conditioned on all spent nuclear fuel being removed from wet storage in the spent fuel pools and placed in dry storage within an ISFSI. Occurrence of postulated accidents associated with spent fuel stored in a spent fuel pool is no longer credible in a spent fuel pool devoid of such fuel. The proposed amendments have no effect on plant systems, structures, and components (SSCs) and no effect on the capability of any plant SSC to perform its design function. The proposed amendments would not increase the likelihood of the malfunction of any plant SSC. The proposed amendments would have no effect on any of the previously evaluated accidents in the SONGS Updated Final Safety Analysis Report (UFSAR).

Since SONGS has permanently ceased operation, the generation of fission products has ceased and the remaining source term continues to decay. This continues to significantly reduce the consequences of previously postulated accidents.

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

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 amendments constitute a revision of the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at SONGS.

The proposed amendments do not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment as a result of the proposed amendments. Similarly, the proposed amendments would not physically change any SSCs involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendments do not create the possibility of a new failure mode associated with any equipment or personnel failures. The credible events for the ISFSI remain unchanged.

Therefore, the proposed amendments 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.

Because the 10 CFR part 50 licenses for SONGS no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. With all nuclear spent fuel transferred out of wet storage from the spent fuel pools and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible. There are no longer credible events that would result in any releases beyond the Exclusion Area Boundary (EAB) exceeding the U.S. Environmental Protection Agency (EPA) Protective Action Guideline (PAG) exposure levels, as detailed in the EPA's “Protective Action Guide and Planning Guidance for Radiological Incidents,” Draft for Interim Use and Public Comment dated March 2013 (PAG Manual).

The proposed amendments do not involve a change in the plant's design, configuration, or operation. The proposed amendments do not affect either the way in which the plant structures, systems, and components perform their safety function or their design margins. Because there is no change to the physical design of the plant, there is no change to any of these margins.

Therefore, the proposed amendments 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: Walker A. Matthews, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Bruce Watson.

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

Date of amendment request: August 31, 2016. A publicly-available version is in ADAMS under Accession No. ML16244A253.

Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2 information and a combined license (COL) License Condition which references one of the proposed changes. Additionally, the proposed changes to the UFSAR eliminate pressurizer spray line monitoring during pressurizer surge line first plant only testing. In addition, these proposed changes correct inconsistencies in testing purpose, testing duration, and the ability to leave equipment in place following the data collection period. These changes involve material which is specifically referenced in Section 2.D.(2) of the COLs. This submittal requests approval of the license amendment necessary to implement these changes.

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

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

Response: No.

The design functions of the RCS [reactor coolant system] include providing an effective reactor coolant pressure boundary. The proposed changes for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement do not impact the existing design requirements for the RCS. These changes are acceptable as they are consistent with the commitments made for the pressurizer surge line monitoring program for the first plant only, and do not adversely affect the capability of the pressurizer surge line and pressurizer spray lines to perform the required reactor coolant pressure boundary design functions.

These proposed changes to the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis do not have an adverse effect on any of the design functions of the systems. The proposed changes do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors.

Therefore, the requested 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 for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis maintain the required design functions, and are consistent with other Updated Final Safety Analysis Report (UFSAR) information. The proposed changes do not adversely affect the design requirements for the RCS, including the pressurizer surge line and pressurizer spray lines. The proposed changes do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes.

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

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

Response: No.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

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

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

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

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: May 25, 2016, as supplemented by letters dated June 15, 2016, and October 18, 2016.

Brief description of amendment: The amendment revised the Millstone Power Station, Unit No. 2, Technical Specifications (TSs) to add the evaluation model EMF-2103(P)(A), Revision 3, “Realistic Large Break LOCA Methodology for Pressurized Water Reactors” (ADAMS Package Accession No. ML16286A579), to the TS Section 6.9.1.8.b list of analytical methods use to establish core operating limits.

Date of issuance: January 24, 2017.

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

Amendment No.: 332. A publicly-available version is in ADAMS under Accession No. ML17025A218; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-65: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: August 30, 2016 (81 FR 59662). The supplemental letters dated June 15, 2016, and October 18, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 24, 2017.

No significant hazards consideration comments received: No.

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

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: August 16, 2016.

Brief description of amendments: The amendments modified license conditions to reflect the transfer of the Master Decommissioning Trust from the Power Authority of the State of New York to Entergy Nuclear Operations, Inc., and deletes other conditions so as to apply the requirements of 10 CFR 50.75(h)(1).

Date of issuance: January 30, 2017.

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

Amendment Nos.: 262 (Indian Point Nuclear Generating Unit No. 3); 313 (James A. FitzPatrick Nuclear Power Plant). A publicly-available version is in ADAMS under Accession No. ML17025A288; documents related to these amendments are listed in the Safety Evaluation enclosed with the letter dated January 27, 2017 (ADAMS Package Accession No. ML16336A488).

Facility Operating License Nos. DPR-64 and DPR-59: Amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: September 27, 2016 (81 FR 66305).

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

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: August 29, 2016, as supplemented by letter dated November 21, 2016.

Brief description of amendment: The amendment revised Technical Specification 5.5.6, “Primary Containment Leakage Rate Testing Program,” to allow permanent extension of Type A and Type C leak rate test intervals through the adoption of Revision 3-A of Nuclear Energy Institute (NEI) 94-01 and the limitations and conditions specified in Revision 2-A of NEI 94-01 as the guidance documents for implementation of performance-based Option B of appendix J to 10 CFR part 50, Option B, “Performance-Based Requirements.” Based on the guidance in Revision 3-A of NEI 94-01, the change allows the maximum interval for the Type A primary containment integrated leakage rate test to extend from once in 10 years to once in 15 years, and the Type C local leak rate test interval to extend to 75 months, provided acceptable performance history and other requirements are maintained.

Date of issuance: January 24, 2017.

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

Amendment No.: 132. A publicly-available version is in ADAMS under Accession No. ML17009A372; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-59: The amendment revised the Renewed Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register: October 11, 2016 (81 FR 70178). The supplemental letter dated November 21, 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 amendment is contained in a Safety Evaluation dated January 24, 2017.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, Ohio

Date of application for amendment: February 17, 2016, as supplemented by letter dated September 6, 2016.

Brief description of amendment: The amendment changed the DBNPS emergency plan by revising the emergency action level scheme.

Date of issuance: January 12, 2017.

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

Amendment No.: 294. A publicly-available version is in ADAMS under Accession No. ML16342C946; documents related to this amendment are listed in the Safely Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-3: The amendment revised the emergency plan.

Date of initial notice in Federal Register: March 15, 2016 (81 FR 13843).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 12, 2017.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: August 29, 2016.

Brief description of amendment: This amendment approves a change to the administrative controls associated with the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.5.4, “Refueling Water Storage Tank.”

Date of issuance: January 18, 2017.

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

Amendment No.: 207. A publicly-available version is in ADAMS under Accession No. ML16348A200; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-12: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: October 11, 2016 (81 FR 70183).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 18, 2017.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San Diego County, California

Date of amendment request: June 16, 2016, as supplemented by letter dated September 6, 2016.

Brief description of amendments: The amendments revised the scheduled implementation date for Milestone 8 of the SONGS, Units 2 and 3, Cyber Security Plan to December 31, 2019, in order to more fully reflect the permanently shutdown status of the facility and accommodate ongoing decommissioning activities.

Date of issuance: January 23, 2017.

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

Amendment Nos.: Unit 2-234 and Unit 3-227: A publicly-available version is in ADAMS under Accession No. ML16252A207; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: August 2, 2016 (81 FR 50735). The supplemental letter dated September 6, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 23, 2017.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: September 8, 2016.

Brief description of amendment: The amendment corrected an error in the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility Operating License No. NPF-8, for Condition 2.C.(23). Specifically, the Unit 2 referenced date representing the start of the 20-year period of extended operation was incorrectly entered as June 25, 2017. The Unit 2 correct date corresponding to the 20-year period of extended operation is March 31, 2021.

Date of issuance: January 23, 2017.

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

Amendment No.: 204. A publicly-available version is in ADAMS under Accession No. ML15329A032; documents related to this amendment is listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-8: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: October 25, 2016 (81 FR 73441).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 23, 2017.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, Houston County, Alabama

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2, Appling County, Georgia

Date of amendment request: March 14, 2016, as supplemented by letters dated May 17, 2016, and October 26, 2016.

Brief description of amendments: The amendments consist of changes that insert generic personnel titles in lieu of plant-specific personnel titles. In addition, the term “plant-specific titles” is replaced with “generic titles” in Technical Specification (TS) 5.2.1.a for each plant. Lastly, this change revised the Hatch, Unit Nos. 1 and 2, TS 5.1 to be consistent with the corresponding Farley, Units 1 and 2, and Vogtle, Units 1 and 2, TS 5.1, and make it consistent with the corresponding Improved Standard Technical Specifications section.

Date of issuance: January 13, 2017.

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

Amendment Nos.: Farley—Unit 1 (207) and Unit 2 (203); Vogtle—Unit 1 (183) and Unit 2 (166); and Hatch—Unit No. 1 (282) and Unit No. 2 (227). A publicly-available version is in ADAMS under Accession No. ML16291A030; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-2, NPF-8, NPF-68, NPF-81, DPR-57, and NPF-5: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: May 24, 2016 (81 FR 32809). The supplemental letters dated May 17, 2016, 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 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 January 13, 2017.

No significant hazards consideration comments received: No.

Susquehanna Nuclear, LLC, Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: January 28, 2016, as supplemented by letters April 6, 2016, and October 10, 2016.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.1, “Residual Heat Removal Service Water (RHRSW) System and the Ultimate Heat Sink (UHS),” and TS 3.8.7, “Distribution Systems—Operating,” to increase the completion time for Conditions A and B of TS 3.7.1, and Condition C of TS 3.8.7, from 72 hours to 7 days, in order to accommodate 480 volt engineered safeguard system load center transformer replacements on the Susquehanna Steam Electric Station, Unit 1. The change is temporary and will be annotated by a note in each TS that specifies the allowance expires on June 15, 2020.

Date of issuance: January 26, 2017.

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

Amendment No.: 248. A publicly-available version is in ADAMS under Accession No. ML17004A250; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-22: The amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: May 24, 2016 (81 FR 32810). The supplemental letter dated October 10, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: August 12, 2016.

Brief description of amendments: The amendments revised Technical Specification (TS) 4.3.1.2, “Fuel Storage Criticality,” for Units 1, 2, and 3, to preclude the placement of fuel in the new fuel storage vaults. This TS change removed the existing TS 4.3.1.2 criticality criteria wording in its entirety, and replaced it with language that specifically restricts the placement of fuel in the new fuel storage vaults.

Date of issuance: January 17, 2017.

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

Amendment Nos.: 296 (Unit 1), 320 (Unit 2), and 280 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML16330A158; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: October 11, 2016 (81 FR 70187).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 17, 2017.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

Date of amendment request: December 8, 2015, as supplemented by letters dated March 11, October 13, December 1, and December 8, 2016.

Brief description of amendment: The amendment revised the Watts Bar Nuclear Plant, Units 1 and 2, Technical Specification (TS) 3.8.1, “AC Sources—Operating,” to extend the Completion Time for one inoperable Diesel Generator from 72 hours to 10 days based on the availability of a supplemental alternating current power source (specifically, the FLEX DG added as part of the mitigating strategies for beyond-design-basis events in response to NRC Order EA-12-049). The amendment also made clarifying changes to certain TS 3.8.1 Conditions, Required Actions, and Surveillance Requirements.

Date of issuance: January 13, 2017.

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

Amendment No.: 110 (Unit 1) and 5 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17006A271; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License Nos. NPF-90 and NPF-96: Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: May 24, 2016 (81 FR 32810). The supplement letters dated October 13, November 1, and December 8, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 13, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 6th day of February 2017.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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