82 FR 10590 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 82, Issue 29 (February 14, 2017)
Federal Register Volume 82, Issue 29 (February 14, 2017)
| Page Range | 10590-10605 | |
| FR Document | 2017-02795 |
[Federal Register Volume 82, Number 29 (Tuesday, February 14, 2017)] [Notices] [Pages 10590-10605] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-02795] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0038] 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: 301-415- 3463; email: [email protected]. 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: 301-415-2242, email: [email protected]. 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 [[Page 10591]] 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 [email protected]. 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 Sec. 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 [[Page 10592]] 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 [email protected], 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 [email protected], 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 [[Page 10593]] 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 [[Page 10594]] 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 [[Page 10595]] 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. [[Page 10596]] 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 [[Page 10597]] 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 [[Page 10598]] 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 [[Page 10599]] 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 [[Page 10600]] 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 [[Page 10601]] 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 [[Page 10602]] 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 [[Page 10603]] 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. [[Page 10604]] 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, [[Page 10605]] 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. [FR Doc. 2017-02795 Filed 2-13-17; 8:45 am] BILLING CODE 7590-01-P
![10590 Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices
financial assistance under the National d. Preservation and Access issue and make immediately effective
Foundation on the Arts and Humanities e. Public Programs any amendment to an operating license
Act of 1965 and make recommendations f. Research Programs or combined license, as applicable,
thereon to the Chairman; and to g. Humanities Medals upon a determination by the
consider gifts offered to NEH and make The remainder of the plenary session Commission that such amendment
recommendations thereon to the will be for consideration of specific involves no significant hazards
Chairman. applications and therefore will be consideration, notwithstanding the
closed to the public. pendency before the Commission of a
DATES: The meeting will be held on
As identified above, portions of the request for a hearing from any person.
Thursday, March 2, 2017, from 10:30
meeting of the National Council on the This biweekly notice includes all
a.m. until 12:30 p.m., and Friday, March
Humanities will be closed to the public notices of amendments issued, or
3, 2017, from 9:00 a.m. until adjourned.
pursuant to sections 552b(c)(4), proposed to be issued, from January 14
ADDRESSES: The meeting will be held at 552b(c)(6) and 552b(c)(9)(b) of Title 5 to January 30, 2017. The last biweekly
Constitution Center, 400 7th Street SW., U.S.C., as amended. The closed sessions notice was published on January 31,
Washington, DC 20506. See will include review of personal and/or 2017.
SUPPLEMENTARY INFORMATION section for proprietary financial and commercial
room numbers. DATES: Comments must be filed by
information given in confidence to the
FOR FURTHER INFORMATION CONTACT:
March 16, 2017. A request for a hearing
agency by grant applicants, and
Elizabeth Voyatzis, Committee must be filed by April 17, 2017.
discussion of certain information, the
Management Officer, 400 7th Street premature disclosure of which could ADDRESSES: You may submit comments
SW., 4th Floor, Washington, DC 20506; significantly frustrate implementation of by any of the following methods:
(202) 606–8322; evoyatzis@neh.gov. proposed agency action. I have made • Federal Rulemaking Web site: Go to
this determination pursuant to the http://www.regulations.gov and search
SUPPLEMENTARY INFORMATION: The
authority granted me by the Chairman’s for Docket ID NRC–2017–0038. Address
National Council on the Humanities is
Delegation of Authority to Close questions about NRC dockets to Carol
meeting pursuant to the National
Advisory Committee Meetings dated Gallagher; telephone: 301–415–3463;
Foundation on the Arts and Humanities
April 15, 2016. email: Carol.Gallagher@nrc.gov. For
Act of 1965 (20 U.S.C. 951–960, as
Please note that individuals planning technical questions, contact the
amended). The Committee meetings of
to attend the public sessions of the individual listed in the FOR FURTHER
the National Council on the Humanities
meeting are subject to security screening INFORMATION CONTACT section of this
will be held on March 2, 2017, as
procedures. If you wish to attend any of document.
follows: The policy discussion session
the public sessions, please inform NEH • Mail comments to: Cindy Bladey,
(open to the public) will convene at
as soon as possible by contacting Ms. Office of Administration, Mail Stop:
10:30 a.m. until approximately 11:30
Katherine Griffin at (202) 606–8322 or OWFN–12–H08, U.S. Nuclear
a.m., followed by the discussion of
kgriffin@neh.gov. Please also provide Regulatory Commission, Washington,
specific grant applications and programs
advance notice of any special needs or DC 20555–0001.
before the Council (closed to the public)
accommodations, including for a sign For additional direction on obtaining
from 11:30 a.m. until 12:30 p.m.
language interpreter. information and submitting comments,
Digital Humanities: Room 4085.
see ‘‘Obtaining Information and
Education Programs: Room 4002. Dated: February 8, 2017. Submitting Comments’’ in the
Federal/State Partnership: Room Elizabeth Voyatzis, SUPPLEMENTARY INFORMATION section of
4089. Committee Management Officer. this document.
Preservation and Access: Room 2002. [FR Doc. 2017–02920 Filed 2–13–17; 8:45 am]
Public Programs/Federal/State FOR FURTHER INFORMATION CONTACT:
Partnership: Room P003.
BILLING CODE 7536–01–P Paula Blechman, Office of Nuclear
Research Programs: Room P002. Reactor Regulation, U.S. Nuclear
In addition, the Humanities Medal Regulatory Commission, Washington DC
Committee (closed to the public) will NUCLEAR REGULATORY 20555–0001; telephone: 301–415–2242,
meet from 2:30 p.m. until 3:30 p.m. in COMMISSION email: Paula.Blechman@nrc.gov.
Room P002. [NRC–2017–0038] SUPPLEMENTARY INFORMATION:
The plenary session of the National I. Obtaining Information and
Council on the Humanities will convene Biweekly Notice; Applications and
Submitting Comments
on March 3, 2017, at 9:00 a.m. in the Amendments to Facility Operating
Conference Center at Constitution Licenses and Combined Licenses A. Obtaining Information
Center. The agenda for the morning Involving No Significant Hazards Please refer to Docket ID NRC–2017–
session (open to the public) will be as Considerations 0038, facility name, unit number(s),
follows: AGENCY: Nuclear Regulatory plant docket number, application date,
A. Minutes of the Previous Meeting Commission. and subject when contacting the NRC
B. Reports ACTION: Biweekly notice. about the availability of information for
1. Chairman’s Remarks this action. You may obtain publicly-
2. Deputy Chairman’s Remarks SUMMARY: Pursuant to Section 189a.(2) available information related to this
3. Presentation by guest speaker of the Atomic Energy Act of 1954, as action by any of the following methods:
sradovich on DSK3GMQ082PROD with NOTICES
Deborah Hess Norris amended (the Act), the U.S. Nuclear • Federal Rulemaking Web site: Go to
4. Congressional Affairs Report Regulatory Commission (NRC) is http://www.regulations.gov and search
5. Reports on Policy and General publishing this regular biweekly notice. for Docket ID NRC–2017–0038.
Matters The Act requires the Commission to • NRC’s Agencywide Documents
a. Digital Humanities publish notice of any amendments Access and Management System
b. Education Programs issued, or proposed to be issued, and (ADAMS): You may obtain publicly-
c. Federal/State Partnership grants the Commission the authority to available documents online in the
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![Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices 10593
all other participants. Filing is storage cask loading and unloading Therefore, the proposed change does not
considered complete by first-class mail requirements for ONS. Specifically, the create the possibility of a new or different
as of the time of deposit in the mail, or license amendment request would kind of accident from any kind of accident
previously evaluated.
by courier, express mail, or expedited revise TS 3.7.12, ‘‘Spent Fuel Pool
3. Does the proposed amendment involve
delivery service upon depositing the Boron Concentration’’; TS 3.7.18, ‘‘Dry a significant reduction in a margin of safety?
document with the provider of the Spent Fuel Storage Cask Loading and Response: No.
service. A presiding officer, having Unloading’’; and TS 4.4, ‘‘Dry Spent The proposed changes to TS 3.7.12, 3.7.18
granted an exemption request from Fuel Storage Cask Loading and and 4.4, do not modify the method of nuclear
using E-Filing, may require a participant Unloading,’’ to remove certain TS fuel storage or handling at ONS, nor make
or party to use E-Filing if the presiding requirements that no longer pertain to any physical changes to the facility design,
officer subsequently determines that the the ONS Independent Spent Fuel material, or construction standards. The
reason for granting the exemption from proposed changes comply with NRC
Storage Facility general license, due to
approved regulations and the station’s Part
use of E-Filing no longer exists. changes in 10 CFR 50.68, ‘‘Criticality 72 and 50 licensing basis.
Documents submitted in adjudicatory accident requirements.’’ Therefore, the proposed TS change does
proceedings will appear in the NRC’s Basis for proposed no significant not involve a significant reduction in a
electronic hearing docket which is hazards consideration determination: margin of safety.
available to the public at https:// As required by 10 CFR 50.91(a), the
adams.nrc.gov/ehd, unless excluded The NRC staff has reviewed the
licensee has provided its analysis of the licensee’s analysis and, based on this
pursuant to an order of the Commission issue of no significant hazards
or the presiding officer. If you do not review, it appears that the three
consideration, which is presented standards of 10 CFR 50.92(c) are
have an NRC-issued digital ID certificate below, with NRC edits in square
as described above, click cancel when satisfied. Therefore, the NRC staff
brackets: proposes to determine that the
the link requests certificates and you
1. Does the proposed change [amendment] amendment request involves no
will be automatically directed to the involve a significant increase in the
NRC’s electronic hearing dockets where significant hazards consideration.
probability or consequences of an accident Attorney for licensee: Lara S. Nichols,
you will be able to access any publicly previously evaluated?
available documents in a particular Vice President Nuclear & EHS Legal
Response: No.
hearing docket. Participants are The proposed changes to Technical
Support, Duke Energy Corporation, 526
requested not to include personal Specifications (TSs) 3.7.12, 3.7.18 and 4.4, do South Church Street—EC07H, Charlotte,
privacy information, such as social not modify the method of nuclear fuel storage NC 28202–1802.
security numbers, home addresses, or or handling at Oconee Nuclear Station (ONS), NRC Branch Chief: Michael T.
personal phone numbers in their filings, or make any physical changes to the facility Markley.
design, material, or construction standards.
unless an NRC regulation or other law Duke Energy Carolinas, LLC, Docket
The proposed change revises the criticality
requires submission of such requirements contained in the TSs, as Nos. 50–269, 50–270, and 50–287,
information. For example, in some allowed by 10 CFR 50.68(c), that are Oconee Nuclear Station, Units 1, 2, and
instances, individuals provide home redundant to regulatory requirements 3 (ONS), Oconee County, South
addresses in order to demonstrate provided in 10 CFR part 72 and the Nuclear Carolina
proximity to a facility or site. With Regulatory Commission (NRC)-approved Date of amendment request: July 21,
respect to copyrighted works, except for Certificate of Compliance (CoC) for the spent
2016. Publicly-available version is in
limited excerpts that serve the purpose fuel dry shielded canisters utilized at ONS.
The proposed change to the TS requirements ADAMS under Accession No.
of the adjudicatory filings and would ML16209A223.
constitute a Fair Use application, neither result[s] in operation that will
increase the probability of initiating an Description of amendment request:
participants are requested not to include analyzed event nor alter[s] assumptions The proposed amendment requests to
copyrighted materials in their relative to mitigation of an accident or revise the Technical Specifications
submission. transient event. The change has no effect on (TSs) for ONS. Specifically, the license
For further details with respect to the process variables, structures, systems, amendment request would revise TS
these license amendment applications, and components that must be maintained 2.1.1.1, ‘‘Reactor Core Safety Limits
see the application for amendment consistent with the safety analyses and
(SLs),’’ and TS 5.6.5, ‘‘Core Operating
which is available for public inspection licensing basis.
Therefore, the proposed change does not Limits Report (COLR),’’ to allow the use
in ADAMS and at the NRC’s PDR. For of the COPERNIC fuel performance
additional direction on accessing involve a significant increase in the
probability or consequences of an accident code.
information related to this document, Basis for proposed no significant
previously evaluated.
see the ‘‘Obtaining Information and 2. Does the proposed amendment create hazards consideration determination:
Submitting Comments’’ section of this the possibility of a new or different kind of As required by 10 CFR 50.91(a), the
document. accident from any accident previously licensee has provided its analysis of the
Duke Energy Carolinas, LLC, Docket evaluated? issue of no significant hazards
Response: No. consideration, which is presented
Nos. 50–269, 50–270, and 50–287, The proposed changes to TSs 3.7.12, 3.7.18
Oconee Nuclear Station Units 1, 2, and below, with NRC edits in square
and 4.4, do not modify the method of nuclear
3 (ONS), Oconee County, South fuel storage or handling at ONS, nor make brackets:
Carolina any physical changes to the facility design, 1. Does the proposed amendment involve
material, or construction standards. The a significant increase in the probability or
sradovich on DSK3GMQ082PROD with NOTICES
Date of amendment request: July 20,
2016. Publicly-available version is in change does not alter the plant configuration consequences of an accident previously
(no new or different type of equipment will evaluated?
ADAMS under Accession No. be installed) or make changes in the methods Response: No.
ML16209A222. governing normal plant operation. The The proposed change adds a limit on
Description of amendment request: proposed change to the ONS TS requirements maximum local fuel pin centerline
The proposed amendment requests to does not adversely impact the results of the temperature to [the] ONS Technical
revise the Technical Specifications ONS safety analyses and is compliant with Specifications that is based on a[n] NRC
(TSs) associated with dry spent fuel the current licensing basis. reviewed and approved fuel performance
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![10594 Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices
code, and does not require a physical change by the NRC reviewed and approved fuel Date of amendment request: April 29,
to plant systems, structures or components. performance code to the ONS Technical 2016, as supplemented by letters dated
Plant operations and analysis will continue Specifications, does not impact the safety October 3, 2016, and January 16, 2017.
to be in accordance with the ONS licensing margins established in the ONS licensing
Publicly-available versions are in
basis. The peak fuel centerline temperature is basis.
the basis for protecting the fuel and is The proposed change also adds a topical ADAMS under Accession Nos.
consistent with the safety analysis. report for a[n] NRC reviewed and approved ML16120A076, ML16277A521, and
The proposed change also adds a topical fuel performance code to the list of topical ML17017A210, respectively.
report for a[n] NRC reviewed and approved reports in [the] ONS Technical Description of amendment request:
fuel performance code to the list of topical Specifications, which is administrative in The NRC staff previously made a
reports in [the] ONS Technical nature and does not amend the cycle specific proposed determination that the
Specifications, which is administrative in parameters presently required by the amendment request dated April 29,
nature and has no impact on a plant Technical Specifications. The individual 2016, involves no significant hazards
configuration or system performance relied Technical Specifications continue to require
upon to mitigate the consequences of an
considerations (81 FR 43650; July 5,
operation of the plant within the bounds of
accident. The list of topical reports in the 2016). Subsequently, by letter dated
the limits specified in the Core Operating
Technical Specifications used to develop the Limits Report. The proposed change to the January 16, 2017, the licensee provided
core operating limits does not impact either list of analytical methods referenced in the additional information that expanded
the initiation of an accident or the mitigation Core Operating Limits Report does not the scope of the amendment request as
of its consequences. impact the margin of safety. originally noticed. Accordingly, this
Therefore, the proposed change does not Therefore, the proposed change does not notice supersedes the previous notice in
involve a significant increase in the involve a significant reduction in a margin of its entirety.
probability or consequences of an accident safety. The amendments would (1)
previously evaluated.
The NRC staff has reviewed the consolidate the Emergency Operations
2. Does the proposed amendment create
the possibility of a new or different kind of licensee’s analysis and, based on this Facilities (EOFs) for BSEP, HNP, and
accident from any accident previously review, it appears that the three RNP with the Duke Energy Progress,
evaluated? standards of 10 CFR 50.92(c) are LLC (Duke Energy) corporate EOF in
Response: No. satisfied. Therefore, the NRC staff Charlotte, North Carolina; (2) decrease
The proposed change adds a limit on proposes to determine that the the frequency for a multi-site drill from
maximum local fuel pin centerline amendment request involves no once per 6 years to once per 8 years; (3)
temperature to [the] ONS Technical allow the multi-site drill performance
significant hazards consideration.
Specifications that is based on a[n] NRC with sites other than CNS, MNS, or
reviewed and approved fuel performance
Attorney for licensee: Lara S. Nichols,
Vice President Nuclear & EHS Legal ONS, (4) change the BSEP, HNP, and
code, and does not require a physical change
to plant systems, structures or components. Support, Duke Energy Corporation, 526 RNP augmentation times to be
Specifying maximum local fuel pin South Church Street—EC07H, Charlotte, consistent with those of the sites
centerline temperature ensures that the fuel NC 28202–1802. currently supported by the Duke Energy
design limits are met. Operations and NRC Branch Chief: Michael T. corporate EOF, and (5) decrease the
analysis will continue to be in compliance Markley. frequency of the unannounced
with NRC regulations. The addition of a new augmentation drill at BSEP from twice
fuel pin centerline melt temperature versus Duke Energy Progress, LLC, Docket Nos.
50–325 and 50–324, Brunswick Steam per year to once per year. The January
burnup relationship does not affect any 16, 2017, letter also acknowledges the
accident initiators that would create a new Electric Plant, Units 1 and 2 (BSEP),
Brunswick County, North Carolina addition of WLS to the Duke Energy
accident.
The proposed change also adds a topical corporate EOF with the issuance of the
Duke Energy Carolinas, LLC, Docket WLS operating license on December 19,
report for a[n] NRC reviewed and approved
fuel performance code to the list of topical
Nos. 50–413 and 50–414, Catawba 2016 (ADAMS Accession No.
reports in [the] ONS Technical Nuclear Station, Units 1 and 2 (CNS), ML16333A329).
Specifications, which is administrative in York County, South Carolina Basis for proposed no significant
nature and has no impact on a plant Duke Energy Progress, LLC, Docket No. hazards consideration determination:
configuration or on system performance. The 50–400, Shearon Harris Nuclear Power As required by 10 CFR 50.91(a), the
proposed change updates the list of NRC- Plant, Unit 1 (HNP), Wake County, licensee has provided its analysis of the
approved topical reports used to develop the issue of no significant hazards
core operating limits. There is no change to
North Carolina
consideration, which is presented
the parameters within which the plant is Duke Energy Carolinas, LLC, Docket
normally operated. The possibility of a new below:
Nos. 50–369 and 50–370, McGuire
or different kind of accident is not created. Nuclear Station, Units 1 and 2 (MNS), 1. Does the proposed change involve a
Therefore, the proposed change does not Mecklenburg County, North Carolina significant increase in the probability or
create the possibility of a new or different consequences of an accident previously
kind of accident from any accident Duke Energy Carolinas, LLC, Docket evaluated?
previously evaluated. Nos. 50–269, 50–270, and 50–287, Response: No.
3. Does the proposed amendment involve Oconee Nuclear Station, Units 1, 2, and The proposed changes relocate the BSEP,
a significant reduction in a margin of safety? 3 (ONS), Oconee County, South HNP, and RNP EOFs from their present
Response: No. Carolina onsite or near-site locations to the established
The proposed change to TS 2.1.1.1 adds a corporate EOF in Charlotte, North Carolina,
limit on maximum local fuel pin centerline Duke Energy Progress, LLC, Docket No. changes the required response times for
temperature that is based on an NRC 50–261, H.B. Robinson Steam Electric supplementing onsite personnel in response
sradovich on DSK3GMQ082PROD with NOTICES
reviewed and approved fuel performance Plant, Unit No. 2 (RNP), Darlington to a radiological emergency, decreases the
code, and does not require a physical change County, South Carolina multi-site drill frequency, allows the multi-
to plant systems, structures or components. site drill to be performed with sites other
Plant operations and analysis will continue Duke Energy Carolinas, LLC, Docket than ONS, MNS, or CNS, and decreases the
to be in accordance with [the] ONS licensing Nos. 52–018 and 52–019, William States frequency of augmentation drills at BSEP.
basis. Lee III Nuclear Station, Units 1, and 2 The functions and capabilities of the
Adding the local fuel pin centerline (WLS), Cherokee County, South relocated EOFs will continue to meet the
temperature and burnup relationship defined Carolina applicable regulatory requirements. It has
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![Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices 10595
been evaluated and determined that the the applicable regulatory requirements. It has Basis for proposed no significant
change in response time does not been evaluated and determined that the hazards consideration determination:
significantly affect the ability to supplement change in response time does not As required by 10 CFR 50.91(a), the
the onsite staff. In addition, analysis shows significantly affect the ability to supplement
licensee has provided its analysis of the
that the onsite staff can acceptably respond the onsite staff. In addition, analysis shows
to an event for longer than the requested time that the onsite staff can acceptably respond issue of no significant hazards
for augmented staff to arrive. The proposed to an event for longer than the requested time consideration, which is presented
changes have no effect on normal plant for augmented staff to arrive. Margin of safety below:
operation or on any accident initiator or is associated with confidence in the ability of 1. Does the proposed change involve a
precursors, and do not impact the function of the fission product barriers (i.e., fuel significant increase in the probability or
plant structures, systems, or components cladding, reactor coolant system pressure consequences of an accident previously
(SSCs). The proposed changes do not alter or boundary, and containment structure) to evaluated?
prevent the ability of the emergency response limit the level of radiation dose to the public. [Response: No.]
organization to perform its intended The proposed changes are associated with The proposed changes do not involve a
functions to mitigate the consequences of an the emergency plans and do not impact significant increase in the probability or
accident or event. operation of the plant or its response to consequences of an accident previously
Therefore, the proposed change does not transients or accidents. The changes do not evaluated. The proposed changes are
involve a significant increase in the affect the Technical Specifications. The administrative in nature, facilitate improved
probability or consequences of an accident changes do not involve a change in the
previously evaluated. content and presentation of Administrative
method of plant operation, and no accident controls, and alter only the format and
2. Does the proposed change create the analyses will be affected by the proposed
possibility of a new or different kind of location of cycle-specific parameter figures
changes. Safety analysis acceptance criteria and limits from the TS to the COLR. This
accident from any accident previously are not affected. The emergency plans will
evaluated? relocation does not result in the alteration of
continue to provide the necessary response the design, material, or construction
Response: No. staff for emergencies as demonstrated by
The proposed changes only impact the standards that were applicable prior to the
staffing and functional analyses including the change. The proposed changes will not result
implementation of the affected stations’ necessary timeliness of performing major
emergency plans by relocating their onsite or in modification of any system interface that
tasks for the functional areas of the
near-site EOFs to the established corporate would increase the likelihood of an accident
emergency plans.
EOF in Charlotte, North Carolina, changing since these events are independent of the
Therefore, the proposed change does not
the required response time of responders proposed change. The proposed amendment
involve a significant reduction in a margin of
who supplement the onsite staff, decreasing will not change, degrade, or prevent actions,
safety.
the multi-site drill frequency, allowing the or alter any assumptions previously made in
multi-site drill to be performed with sites The NRC staff has reviewed the evaluating the radiological consequences of
other than ONS, MNS, or CNS, and licensee’s analysis and, based on this an accident described in the Final Safety
decreasing the frequency of augmentation review, it appears that the three Analysis Report (FSAR).
drills at BSEP. The functions and capabilities standards of 10 CFR 50.92(c) are Therefore, the proposed changes do not
of the relocated EOFs will continue to meet involve a significant increase in the
satisfied. Therefore, the NRC staff
the applicable regulatory requirements. It has probability or consequences of an accident
proposes to determine that the previously evaluated.
been evaluated and determined that the
amendment request involves no 2. Does the proposed change create the
change in response time does not
significantly affect the ability to supplement significant hazards consideration. possibility of a new or different kind of
the onsite staff. In addition, analysis shows Attorney for licensee: Lara S. Nichols, accident from any previously evaluated?
that the onsite staff can acceptably respond Deputy General Counsel, Duke Energy [Response: No.]
to an event for longer than the requested time Corporation, 550 South Tyron Street, The proposed changes do not create the
for augmented staff to arrive. The proposed Mail Code DEC45A, Charlotte, NC possibility of a new or different kind of
changes will not change the design function 28202. accident from any accident previously
or operation of SSCs. The changes do not NRC Branch Chief: Benjamin G. evaluated. The proposed changes do not
impact the accident analysis. The changes do involve any change to the configuration or
Beasley. method of operation of any plant equipment.
not involve a physical alteration of the plant,
a change in the method of plant operation, Duke Energy Progress, LLC, Docket No. Accordingly, no new failure modes have
or new operator actions. The proposed 50–400, Shearon Harris Nuclear Power been defined for any plant system or
changes do not introduce failure modes that Plant (HNP), Unit 1, Wake and Chatham component important to safety nor has any
could result in a new accident, and the new limiting single failure been identified as
Counties, North Carolina
changes do not alter assumptions made in the a result of the proposed changes. Also, there
safety analysis. Date of amendment request: will be no change in types or increase in the
Therefore, the proposed change does not December 2, 2016. A publicly-available amounts of any effluents released offsite.
create the possibility of a new or different version is in ADAMS under Accession Therefore, the proposed changes do not
kind of accident from any accident No. ML16337A249. create the possibility of a new or different
previously evaluated. kind of accident from any accident
Description of amendment request:
3. Does the proposed change involve a previously evaluated.
significant reduction in a margin of safety? The amendment would revise HNP 3. Does the proposed change involve a
Response: No. Technical Specifications (TSs) to significant reduction in a margin of safety?
The proposed changes only impacts the relocate selected figures and values from [Response: No.]
implementation of the affected stations’ the TSs to the Core Operating Limits The proposed changes do not involve a
emergency plans by relocating their onsite or Report (COLR), remove all references to significant reduction in a margin of safety.
near-site EOFs to the established corporate a specific plant procedure as it pertains Previously-approved methodologies will
EOF in Charlotte, North Carolina, changing to the COLR, and adopt Technical continue to be used in determination of
sradovich on DSK3GMQ082PROD with NOTICES
the required response time of responders Specification Task Force (TSTF)-5, cycle-specific core operating limits that are
who supplement the onsite staff, decreasing present in the COLR. The proposed changes
‘‘Delete Safety Limit Violation
the multi-site drill frequency, allowing the are administrative in nature and will not
multi-site drill to be performed with sites Notification Requirements,’’ Revision 1, affect the plant design or system operating
other than ONS, MNS, or CNS, and which deletes duplicate notification, parameters. As such, there is no detrimental
decreasing the frequency of augmentation reporting and restart requirements from impact on any equipment design parameter
drills at BSEP. The functions and capabilities the Administrative Controls section of and the plant will continue to be operated
of the relocated EOFs will continue to meet TSs. within prescribed limits.
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![Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices 10597
accidents, and do not create the possibility of test of each 10-year service period when The integrity of the containment is subject
a new or different kind of accident from any the plant is shutdown for the 10 year to two types of failure mechanisms that can
accident previously evaluated. plant ISI. be categorized as: (1) Activity based and (2)
Therefore, the proposed changes do not Basis for proposed no significant time based. Activity based failure
create the possibility of a new or different mechanisms are defined as degradation due
kind of accident from any accident
hazards consideration determination: to system and/or component modifications or
previously evaluated. As required by 10 CFR 50.91(a), the maintenance. Local leak rate test
3. Does the proposed change involve a licensee has provided its analysis of the requirements and administrative controls
significant reduction in a margin of safety? issue of no significant hazards such as configuration management and
Response: No. consideration, which is presented procedural requirements for system
Plant safety margins are established below: restoration ensure that containment integrity
through limiting conditions for operation, is not degraded by plant modifications or
limiting safety system settings, and safety 1. Does the proposed amendment involve maintenance activities. The design and
limits specified in the technical a significant increase in the probability or construction requirements of the
specifications. The proposed changes to the consequences of an accident previously containment combined with the containment
TS are editorial or administrative in nature evaluated? inspections performed in accordance with
and do not impact any safety margins. Response: No. ASME [American Society of Mechanical
Because there is no impact on established The proposed amendment to the TS Engineers] Section XI and TS requirements
safety margins as a result of these changes, involves the extension of the LSCS Type A serve to provide a high degree of assurance
the proposed change does not involve a containment test interval to 15 years and the that the containment would not degrade in a
significant reduction in a margin of safety. extension of the Type C test interval to 75 manner that is detectable only by a Type A
Therefore, the proposed changes do not months. The current Type A test interval of test. Based on the above, the proposed
involve a significant reduction in a margin of 120 months (10 years) would be extended on extensions do not significantly increase the
safety. a permanent basis to no longer than 15 years consequences of an accident previously
from the last Type A test. The current Type evaluated.
The NRC staff has reviewed the C test interval of 60 months for selected The proposed amendment also deletes
licensee’s analysis and, based on this components would be extended on a
exceptions previously granted to allow one-
review, it appears that the three performance basis to no longer than 75
time extensions of the ILRT test frequency for
standards of 10 CFR 50.92(c) are months. Extensions of up to nine months
LSCS. These exceptions were for activities
satisfied. Therefore, the NRC staff (total maximum interval of 84 months for
that would have already taken place by the
Type C tests) are permissible only for non-
proposes to determine that the routine emergent conditions.
time this amendment is approved; therefore,
amendment request involves no their deletion is solely an administrative
The proposed extension does not involve
significant hazards consideration. action that has no effect on any component
either a physical change to the plant or a
Attorney for licensee: Jeanne Cho, and no impact on how the unit is operated.
change in the manner in which the plant is
Senior Counsel, Entergy Services, Inc., Therefore, the proposed change does not
operated or controlled. The containment is
result in a significant increase in the
440 Hamilton Ave., White Plains, NY designed to provide an essentially leak tight
probability or consequences of an accident
10601. barrier against the uncontrolled release of
radioactivity to the environment for previously evaluated.
NRC Branch Chief: David J. Wrona. 2. Does the proposed change create the
postulated accidents. As such, the
Exelon Generation Company, LLC, containment and the testing requirements possibility of a new or different kind of
Docket Nos. 50–373 and 50–374, LaSalle invoked to periodically demonstrate the accident from any accident previously
integrity of the containment exist to ensure evaluated?
County Station (LSCS), Units 1 and 2, Response: No.
LaSalle County, Illinois the plant’s ability to mitigate the
consequences of an accident, and do not The proposed amendment to the TS
Date of amendment request: October involve the prevention or identification of involves the extension of the LSCS Type A
26, 2016. A publicly-available version is any precursors of an accident. The change in containment test interval to 15 years and the
in ADAMS under Accession No. dose risk for changing the Type A test extension of the Type C test interval to 75
ML16300A200. frequency from three-per-ten years to once- months. The containment and the testing
per-fifteen years, measured as an increase to requirements to periodically demonstrate the
Description of amendment request:
the total integrated dose risk for all internal integrity of the containment exist to ensure
The proposed change revises TS 5.5.13, the plant’s ability to mitigate the
‘‘Primary Containment Leakage Rate events accident sequences for LSCS, is
1.23E–02 person-rem/yr (0.33%) using the consequences of an accident do not involve
Testing Program,’’ to allow for the EPRI [Electric Power Research Institute] any accident precursors or initiators. The
permanent extension of the Type A guidance with the base case corrosion proposed change does not involve a physical
Integrated Leak Rate Testing (ILRT) and included. The change in dose risk drops to change to the plant (i.e., no new or different
Type C Leak Rate Testing frequencies, to 3.15E–03 person-rem/yr (0.08%) when using type of equipment will be installed) or a
change the documents used by LSCS to the EPRI Expert Elicitation methodology. The change to the manner in which the plant is
implement the performance-based values calculated per the EPRI guidance are operated or controlled.
leakage testing program, and to delete all lower than the acceptance criteria of ≤1.0 The proposed amendment also deletes
person-rem/yr or <1.0% person-rem/yr exceptions previously granted to allow one-
the information regarding the time extensions of the ILRT test frequency for
defined in Section 1.3 of Attachment 3 of this
performance of the next LSCS Type A submittal. The results of the risk assessment LSCS. These exceptions were for activities
tests to be performed. for this amendment meet these criteria. that would have already taken place by the
Additionally, this license amendment Moreover, the risk impact for the ILRT time this amendment is approved; therefore,
request (LAR) proposes to delete extension when compared to other severe their deletion is solely an administrative
Condition 2.D.(e) of the LSCS Unit 1 accident risks is negligible. Therefore, this action that does not result in any change in
Renewed Facility Operating License proposed extension does not involve a how the unit is operated.
regarding conducting the third Type A significant increase in the probability of an Therefore, the proposed change does not
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Test of each 10-year service period accident previously evaluated. create the possibility of a new or different
As documented in NUREG–1493, Type B kind of accident from any previously
when the plant is shutdown for the 10-
and C tests have identified a very large evaluated.
year plant inservice inspection (ISI). percentage of containment leakage paths, and 3. Does the proposed change involve a
Similarly, this LAR proposes to delete the percentage of containment leakage paths significant reduction in a margin of safety?
Condition 2.D.(c) of the LSCS Unit 2 that are detected only by Type A testing is Response: No.
Renewed Facility Operating License very small. The LSCS Type A test history The proposed amendment to TS 5.5.13
regarding conducting the third Type A supports this conclusion. involves the extension of the LSCS Type A
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containment test interval to 15 years and the South Carolina Electric & Gas Company The proposed amendment does not affect
extension of the Type C test interval to 75 (SCE&G) and South Carolina Public the probability of an accident previously
months for selected components. This Service Authority, Docket Nos. 52–027 evaluated. Similarly, because the proposed
amendment does not alter the manner in amendment does not alter the design or
and 52–028, Virgil C. Summer Nuclear
which safety limits, limiting safety system set operation of the nuclear plant or any plant
Station (VCSNS) Units 2 and 3, Fairfield SSCs, the proposed amendment does not
points, or limiting conditions for operation County, South Carolina represent a change to the radiological effects
are determined. The specific requirements
Date of amendment request: of an accident, and therefore, does not
and conditions of the TS Containment Leak involve an increase in the consequences of an
Rate Testing Program exist to ensure that the November 10, 2016. A publicly- accident previously evaluated.
degree of containment structural integrity available version is in ADAMS under Therefore, the proposed amendment does
and leak-tightness that is considered in the Accession No. ML16316A003. not involve a significant increase in the
plant safety analysis is maintained. The Description of amendment request: probability or consequences of an accident
overall containment leak rate limit specified The amendment request proposes to add previously evaluated.
by TS is maintained. to License Condition 2.D.(1) of the 2. Does the proposed amendment create
The proposed change involves only the VCSNS Units 2 and 3 combined licenses the possibility of a new or different kind of
extension of the interval between Type A accident from any accident previously
(COLs), an Interim Amendment Request
containment leak rate tests and Type C tests evaluated?
process for changes during construction Response: No.
for LSCS. The proposed surveillance interval when emergent conditions are present. The proposed amendment would add an
extension is bounded by the 15-year ILRT Basis for proposed no significant Interim Amendment Request process to
Interval and the 75-month Type C test hazards consideration determination: Condition 2.D.(1) of the VCSNS 2 and 3 COLs
interval currently authorized within NEI As required by 10 CFR 50.91(a), the to allow construction to continue, at SCE&G’s
[Nuclear Energy Institute] 94–01, Revision 3– own risk, in emergent conditions, where a
licensee has provided its analysis of the
A. Industry experience supports the non-conforming condition that has little or
issue of no significant hazards
conclusion that Type B and C testing detects no safety significance is discovered and the
consideration, which is presented below
a large percentage of containment leakage work activity cannot be adjusted. The Interim
paths and that the percentage of containment
with NRC staff edits in square brackets: Amendment Request process would require
leakage paths that are detected only by Type 1. Does the proposed amendment involve SCE&G to submit a Nuclear Construction
A testing is small. The containment a significant increase in the probability or Safety Assessment which (1) identifies the
inspections performed in accordance with consequences of an accident previously proposed change; (2) evaluates whether
ASME Section XI and TS serve to provide a evaluated? emergent conditions are present; (3)
high degree of assurance that the Response: No. evaluates whether the change would result in
The proposed amendment would add an any material decrease in safety; and (4)
containment would not degrade in a manner
Interim Amendment Request process to evaluates whether continued construction
that is detectable only by Type A testing. The
Condition 2.D.(1) of the VCSNS 2 and 3 COLs would make the non-conforming condition
combination of these factors ensures that the to allow construction to continue, at SCE&G’s irreversible. Only if the continued
margin of safety in the plant safety analysis own risk, in emergent conditions, where a construction would have no material
is maintained. The design, operation, testing non-conforming condition that has little or decrease in safety would NRC issue a
methods and acceptance criteria for Type A, no safety significance is discovered and the determination that construction could
B, and C containment leakage tests specified work activity cannot be adjusted. The Interim continue pending SCE&G’s initiation of the
in applicable codes and standards would Amendment Request process would require COL–ISG–025 PAR/LAR process.
continue to be met, with the acceptance of SCE&G to submit a Nuclear Construction The proposed amendment is not a
this proposed change, since these are not Safety Assessment which (1) identifies the modification, addition to, or removal of any
affected by changes to the Type A and Type proposed change; (2) evaluates whether plant SSCs. Furthermore, the proposed
C test intervals. emergent conditions are present; (3) amendment is not a change to procedures or
The proposed amendment also deletes evaluates whether the change would result in method of control of the nuclear plant or any
exceptions previously granted to allow one any material decrease in safety; and (4) plant SSCs. The proposed amendment only
time extensions of the ILRT test frequency for evaluates whether continued construction adds a new screening process and does not
would make the non-conforming condition change the design, construction, or operation
LSCS. These exceptions were for activities
irreversible. Only if the continued of the nuclear plant or any plant operations.
that would have already taken place by the
construction would have no material Therefore, the proposed amendment does
time this amendment is approved; therefore, decrease in safety would the NRC issue a not create the possibility of a new or different
their deletion is solely an administrative determination that construction could kind of accident from any accident
action and does not change how the unit is continue pending SCE&G’s initiation of the previously evaluated.
operated and maintained. Thus, there is no COL–ISG–025 PAR [preliminary amendment 3. Does the proposed amendment involve
reduction in any margin of safety. request]/LAR [license amendment request] a significant reduction in a margin of safety?
Therefore, the proposed change does not process. The requirement to include a Response: No.
involve a significant reduction in a margin of Nuclear Construction Safety Assessment The proposed amendment would add an
safety. ensures that the proposed amendment would Interim Amendment Request process to
not involve a significant increase in the Condition 2.D.(1) of the VCSNS 2 and 3 COLs
The NRC staff has reviewed the probability or consequences of an accident to allow construction to continue, at SCE&G’s
licensee’s analysis and, based on this previously evaluated. If the continued own risk, in emergent conditions, where a
review, it appears that the three construction would result a material decrease non-conforming condition that has little or
standards of 10 CFR 50.92(c) are in safety, then continued construction would no safety significance is discovered and the
satisfied. Therefore, the NRC staff not be authorized. work activity cannot be adjusted. The Interim
The proposed amendment does not modify Amendment Request process would require
proposes to determine that the the design, construction, or operation of any SCE&G to submit a Nuclear Construction
sradovich on DSK3GMQ082PROD with NOTICES
requested amendments involve no plant structures, systems, or components Safety Assessment which (1) identifies the
significant hazards consideration. (SSCs), nor does it change any procedures or proposed change; (2) evaluates whether
Attorney for licensee: Tamra Domeyer, method of control for any SSCs. Because the emergent conditions are present; (3)
Associate General Counsel, Exelon proposed amendment does not change the evaluates whether the change would result in
Generation Company, LLC, 4300 design, construction, or operation of any any material decrease in safety; and (4)
SSCs, it does not adversely affect any design evaluates whether continued construction
Winfield Road, Warrenville, IL 60555. function as described in the Updated Final would make the non-conforming condition
NRC Branch Chief: David J. Wrona. Safety Analysis Report. irreversible. Only if the continued
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![Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices 10599
construction would have no material As required by 10 CFR 50.91(a), the to remove the pressurizer spray and surge
decrease in safety would the NRC issue licensee has provided its analysis of the line valve leakage requirement as described
determination that construction could issue of no significant hazards in the current licensing basis maintain the
continue pending SCE&G’s initiation of the required design functions, and are consistent
COL–ISG–025 PAR/LAR process.
consideration, which is presented below
with other Updated Final Safety Analysis
The proposed amendment is not a with NRC staff edits in square brackets: Report (UFSAR) information. The proposed
modification, addition to, or removal of any 1. Does the proposed amendment involve changes do not adversely affect the design
plant SSCs. Furthermore, the proposed a significant increase in the probability or requirements for the RCS, including the
amendment is not a change to procedures or consequences of an accident previously pressurizer surge line and pressurizer spray
method of control of the nuclear plant or any evaluated? lines. The proposed changes do not adversely
plant SSCs. The proposed amendment does Response: No. affect the design function, support, design, or
not alter any design function or safety The design functions of the RCS [reactor operation of mechanical and fluid systems.
analysis. Consequently, no safety analysis or coolant system] include providing an The proposed changes do not result in a new
design basis acceptance limit/criterion is effective reactor coolant pressure boundary. failure mechanism or introduce any new
challenged or exceeded by the proposed The proposed changes for removing the accident precursors. No design function
amendment, thus the margin of safety is not requirement to install temporary described in the UFSAR is adversely affected
reduced. The only impact of this activity is instrumentation on the pressurizer spray line by the proposed changes.
the addition of an Interim Amendment during the monitoring of the pressurizer Therefore, the requested amendment does
Request process. surge line for thermal stratification and not create the possibility of a new or different
Therefore, the proposed amendment does thermal cycling during hot functional testing kind of accident from any accident
not involve a significant reduction in a and during the first fuel cycle for the first previously evaluated.
margin of safety. plant only, proposed changes to parameter 3. Does the proposed amendment involve
The NRC staff has reviewed the retention requirements, and proposed change a significant reduction in a margin of safety?
licensee’s analysis and, based on this to remove the pressurizer spray and surge Response: No.
line valve leakage requirement do not impact No safety analysis or design basis
review, it appears that the three
the existing design requirements for the RCS. acceptance limit/criterion is challenged or
standards of 10 CFR 50.92(c) are These changes are acceptable as they are exceeded by the proposed changes, and no
satisfied. Therefore, the NRC staff consistent with the commitments made for margin of safety is reduced. Therefore, the
proposes to determine that the the pressurizer surge line monitoring requested amendment does not involve a
amendment request involves no program for the first plant only, and do not significant reduction in a margin of safety.
significant hazards consideration. adversely affect the capability of the
Attorney for licensee: M. Stanford pressurizer surge line and pressurizer spray The NRC staff has reviewed the
Blanton, Balch & Bingham LLP, 1710 lines to perform the required reactor coolant licensee’s analysis and, based on this
Sixth Avenue North, Birmingham, AL pressure boundary design functions. review, it appears that the three
35203–2015. These proposed changes to the monitoring standards of 10 CFR 50.92(c) are
NRC Branch Chief: Jennifer Dixon- of the pressurizer surge line for thermal satisfied. Therefore, the NRC staff
stratification and thermal cycling during hot proposes to determine that the
Herrity. functional testing and during the first fuel amendment request involves no
South Carolina Electric & Gas Company cycle for the first plant only, proposed
changes to parameter retention requirements,
significant hazards consideration.
and South Carolina Public Service Attorney for licensee: Ms. Kathryn M.
Authority, Docket Nos. 52–027 and 52– and proposed change to remove the
pressurizer spray and surge line valve Sutton, Morgan, Lewis & Bockius, LLC,
028, Virgil C. Summer Nuclear Station, 1111 Pennsylvania Ave. NW.,
leakage requirement as described in the
Units 2 and 3, Fairfield, South Carolina current licensing basis do not have an Washington, DC 20004–2514.
Date of amendment request: adverse effect on any of the design functions NRC Branch Chief: Jennifer Dixon-
September 15, 2016. A publicly- of the systems. The proposed changes do not Herrity.
available version is in ADAMS under affect the support, design, or operation of
mechanical and fluid systems required to South Carolina Electric & Gas Company
Accession No. ML16259A315.
Description of amendment request: mitigate the consequences of an accident. and South Carolina Public Service
The amendment request proposes
There is no change to plant systems or the Authority, Docket Nos. 52–027 and 52–
response of systems to postulated accident 028, Virgil C. Summer Nuclear Station,
changes to the Updated Final Safety conditions. There is no change to the
Analysis Report (UFSAR) in the form of Units 2 and 3, Fairfield, South Carolina
predicted radioactive releases due to
departures from the incorporated plant- postulated accident conditions. The plant Date of amendment request:
specific Design Control Document Tier response to previously evaluated accidents or November 16, 2016. A publicly-
2 information and a combined license external events is not adversely affected, nor available version is in ADAMS under
(COL) License Condition which do the proposed changes create any new Accession No. ML16323A020.
references one of the proposed changes. accident precursors. Description of amendment request:
Additionally, the proposed changes to Therefore, the requested amendment does The amendment request proposes
not involve a significant increase in the
the UFSAR eliminate pressurizer spray changes to plant-specific Tier 1
probability or consequences of an accident
line monitoring during pressurizer surge previously evaluated. information, with corresponding
line first plant only testing. In addition, 2. Does the proposed amendment create changes to the associated Combined
these proposed changes correct the possibility of a new or different kind of License (COL) Appendix C information,
inconsistencies in testing purpose, accident from any accident previously and involves associated Tier 2
testing duration, and the ability to leave evaluated? information in the Updated Final Safety
equipment in place following the data Response: No. Analysis Report (UFSAR). Specifically,
The proposed changes for removing the the requested amendment proposes
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collection period. These changes
involve material which is specifically requirement to install temporary clarifications to a plant-specific Tier 1
instrumentation on the pressurizer spray line
referenced in Section 2.D.(2) of the COL. (and COL Appendix C) table and a
during the monitoring of the pressurizer
This submittal requests approval of the surge line for thermal stratification and UFSAR table in regard to the
license amendment necessary to thermal cycling during hot functional testing inspections of the excore source,
implement these changes. and during the first fuel cycle for the first intermediate, and power range
Basis for proposed no significant plant only, proposed changes to parameter detectors. Pursuant to the provisions of
hazards consideration determination: retention requirements, and proposed change 10 CFR 52.63(b)(1), an exemption from
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![10600 Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices
elements of the design as certified in the design function of the excore detectors. The SONGS, Units 2 and 3 TS. These
10 CFR part 52, appendix D, design detector canning materials of construction changes will more fully reflect the
certification rule is also requested for are compatible with the post-LOCA permanently shutdown and defueled
containment environment and do not
the plant-specific Design Control status of the facility, as well as the
contribute a significant amount of hydrogen
Document Tier 1 material departures. or chemical precipitates. The change to the reduced scope of structures, systems,
Basis for proposed no significant ITAAC aligns the inspection with the Tier 2 and components necessary to ensure
hazards consideration determination: design feature. Consequently, because the plant safety once all spent fuel has been
As required by 10 CFR 50.91(a), the excore detectors functions are unchanged, permanently moved to the SONGS
licensee has provided its analysis of the there are no adverse effects on accidents ISFSI, an activity which is currently
issue of no significant hazards previously evaluated in the UFSAR. scheduled for completion in 2019.
consideration, which is presented below Therefore, the proposed amendment does Basis for proposed no significant
with NRC staff’s edits in square not create the possibility of a new or different hazards consideration determination:
kind of accident from any accident
brackets: previously evaluated.
As required by 10 CFR 50.91(a), the
1. Does the proposed amendment involve 3. Does the proposed amendment involve licensee has provided its analysis of the
a significant increase in the probability or a significant reduction in a margin of safety? issue of no significant hazards
consequences of an accident previously Response: No. consideration, which is presented
evaluated? The proposed change to specify the below:
Response: No. inspection of the excore source, intermediate,
1. Does the proposed amendment involve
The proposed change to specify the and power range detectors is done to verify
a significant increase in the probability or
inspection of the excore source, intermediate, that aluminum surfaces are contained in consequences of an accident previously
and power range detectors is done to verify stainless steel or titanium, and avoids the evaluated?
that aluminum surfaces are contained in introduction of aluminum into the post- Response: No.
stainless steel or titanium, and avoids the LOCA containment environment, does not The proposed amendment would modify
introduction of aluminum into the post-loss alter any safety-related equipment, applicable the SONGS, Units 2 and 3 facility operating
of coolant accident (LOCA) containment design codes, code compliance, design licenses and TS by deleting the portions of
environment due to detector materials. The function, or safety analysis. Consequently, no the licenses and TSs that are no longer
proposed change does not alter any safety safety analysis or design basis acceptance applicable to a facility with no spent nuclear
related functions. The materials of limit/criterion is challenged or exceeded by fuel stored in the SFP, while modifying the
construction are compatible with the post- the proposed change, thus the margin of remaining portions to correspond to all
LOCA conditions inside containment and safety is not reduced. nuclear fuel stored within an ISFSI. This
will not significantly contribute to hydrogen Therefore, the proposed amendment does amendment becomes effective upon removal
generation or chemical precipitates. The not involve a significant reduction in a of all spent nuclear fuel from the SONGS,
change does not affect the operation of any margin of safety. Units 2 and 3 SFP and its transfer to dry cask
systems or equipment that initiate an storage within an ISFSI.
The NRC staff has reviewed the
analyzed accident or alter any structures, Additionally, the proposed change would
systems, and components (SSC) accident licensee’s analysis and, based on this
review, it appears that the three revise the Unit 1 TSs for consistency with the
initiator or initiating sequence of events. proposed changes to the Units 2 and 3 TSs.
The change does not impact the support, standards of 10 CFR 50.92(c) are Similar to the changes for Units 2 and 3, the
design, or operation of mechanical and fluid satisfied. Therefore, the NRC staff Unit 1 changes delete portions of the TSs that
systems. There is no change to plant systems proposes to determine that the are no longer applicable to a facility with
or the response of systems to postulated amendment request involves no spent fuel no longer stored in the SFP, while
accident conditions. There is no change to significant hazards consideration. modifying the remaining portions to
the predicted radioactive releases due to Attorney for licensee: Ms. Kathryn M. correspond to all nuclear fuel in dry storage.
normal operation or postulated accident Sutton, Morgan, Lewis & Bockius, LLC, The Unit 1 TSs are also proposed to be
conditions. Consequently, the plant response combined with the Units 2 and 3 TSs.
1111 Pennsylvania Avenue NW.,
to previously evaluated accidents or external The definition of safety-related Structures,
events is not adversely affected, nor does the Washington, DC 20004–2514.
Systems, and Components (SSCs) in 10 CFR
proposed change create any new accident NRC Branch Chief: Jennifer Dixon- 50.2 states that safety-related SSCs are those
precursors. Herrity. relied on to remain functional during and
Therefore, the proposed amendment does Southern California Edison Company, et following design basis events to assure:
not involve a significant increase in the 1. The integrity of the reactor coolant
probability or consequences of an accident
al., Docket Nos. 50–206, 50–361, and
boundary;
previously evaluated. 50–362, San Onofre Nuclear Generating 2. The capability to shutdown the reactor
2. Does the proposed amendment create Station (SONGS), Units 1, 2, and 3, San and maintain it in a safe shutdown condition;
the possibility of a new or different kind of Diego County, California or
accident from any accident previously Date of amendment request: 3. The capability to prevent or mitigate the
evaluated? December 15, 2016. A publicly-available consequences of accidents which could
Response: No. result in potential offsite exposures
The proposed change does not affect the
version is in ADAMS under Accession comparable to the applicable guideline
operation of any systems or equipment that No. ML16355A014. exposures set forth in 10 CFR 50.43(a)(1) or
may initiate a new or different kind of Description of amendment request: 100.11.
accident, or alter any SSC such that a new The proposed amendment would revise The first two criteria (integrity of the
accident initiator or initiating sequence of the Facility Operating Licenses and reactor coolant pressure boundary and safe
events is created. The proposed change to associated Technical Specifications (TS) shutdown of the reactor) are not applicable
specify the inspection of the excore source, for SONGS, Units 1, 2, and 3, to reflect to a plant in a permanently defueled
intermediate, and power range detectors is removal of all spent nuclear fuel from condition. The third criterion is related to
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done to verify that aluminum surfaces are the SONGS, Units 2 and 3 spent fuel preventing or mitigating the consequences of
contained in stainless steel or titanium, and accidents that could result in potential offsite
pools (SFPs) and its transfer to dry cask
avoids the introduction of aluminum into the exposures exceeding limits. However, after
post-LOCA containment environment due to
storage within an onsite independent all nuclear spent fuel assemblies have been
detector materials. In addition, the proposed spent fuel storage installation (ISFSI). transferred to dry cask storage within an
change to the ITAAC [inspections, tests, The proposed changes would also make ISFSI, none of the SSCs at SONGS, Units 2
analysis, and acceptance criteria] verified conforming changes to the SONGS, Unit and 3 are required to be relied on for accident
materials of construction does not alter the 1 TS and combine them with the mitigation. Therefore, none of the SSCs at
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![10602 Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices
no longer credible. With all nuclear spent of the license amendment necessary to thermal cycling during hot functional testing
fuel transferred out of wet storage from the implement these changes. and during the first fuel cycle for the first
spent fuel pools and placed in dry storage Basis for proposed no significant plant only, proposed changes to parameter
within the ISFSI, a fuel handling accident is hazards consideration determination: retention requirements, and proposed change
no longer credible. There are no longer to remove the pressurizer spray and surge
As required by 10 CFR 50.91(a), the line valve leakage requirement as described
credible events that would result in any
releases beyond the Exclusion Area licensee has provided its analysis of the in the current licensing basis maintain the
Boundary (EAB) exceeding the U.S. issue of no significant hazards required design functions, and are consistent
Environmental Protection Agency (EPA) consideration, which is presented below with other Updated Final Safety Analysis
Protective Action Guideline (PAG) exposure with NRC staff edits in square brackets: Report (UFSAR) information. The proposed
levels, as detailed in the EPA’s ‘‘Protective changes do not adversely affect the design
1. Does the proposed amendment involve
Action Guide and Planning Guidance for requirements for the RCS, including the
a significant increase in the probability or
Radiological Incidents,’’ Draft for Interim Use pressurizer surge line and pressurizer spray
consequences of an accident previously
and Public Comment dated March 2013 (PAG lines. The proposed changes do not adversely
evaluated?
Manual). affect the design function, support, design, or
Response: No.
The proposed amendments do not involve operation of mechanical and fluid systems.
The design functions of the RCS [reactor
a change in the plant’s design, configuration, The proposed changes do not result in a new
coolant system] include providing an
or operation. The proposed amendments do failure mechanism or introduce any new
effective reactor coolant pressure boundary.
not affect either the way in which the plant accident precursors. No design function
The proposed changes for removing the
structures, systems, and components perform described in the UFSAR is adversely affected
requirement to install temporary
their safety function or their design margins. by the proposed changes.
instrumentation on the pressurizer spray line
Because there is no change to the physical Therefore, the requested amendment does
during the monitoring of the pressurizer
design of the plant, there is no change to any not create the possibility of a new or different
surge line for thermal stratification and
of these margins. kind of accident from any accident
thermal cycling during hot functional testing
Therefore, the proposed amendments do and during the first fuel cycle for the first previously evaluated.
not involve a significant reduction in a plant only, proposed changes to parameter 3. Does the proposed amendment involve
margin of safety. retention requirements, and proposed change a significant reduction in a margin of safety?
to remove the pressurizer spray and surge Response: No.
The NRC staff has reviewed the No safety analysis or design basis
licensee’s analysis and, based on this line valve leakage requirement do not impact
the existing design requirements for the RCS. acceptance limit/criterion is challenged or
review, it appears that the three These changes are acceptable as they are exceeded by the proposed changes, and no
standards of 10 CFR 50.92(c) are consistent with the commitments made for margin of safety is reduced.
satisfied. Therefore, the NRC staff the pressurizer surge line monitoring Therefore, the requested amendment does
proposes to determine that the program for the first plant only, and do not not involve a significant reduction in a
amendment request involves no adversely affect the capability of the margin of safety.
significant hazards consideration. pressurizer surge line and pressurizer spray The NRC staff has reviewed the
Attorney for licensee: Walker A. lines to perform the required reactor coolant licensee’s analysis and, based on this
Matthews, Esquire, Southern California pressure boundary design functions. review, it appears that the three
These proposed changes to the monitoring
Edison Company, 2244 Walnut Grove standards of 10 CFR 50.92(c) are
of the pressurizer surge line for thermal
Avenue, Rosemead, California 91770. stratification and thermal cycling during hot satisfied. Therefore, the NRC staff
NRC Branch Chief: Bruce Watson. functional testing and during the first fuel proposes to determine that the
Southern Nuclear Operating Company, cycle for the first plant only, proposed amendment request involves no
Docket Nos. 52–025 and 52–026, Vogtle changes to parameter retention requirements, significant hazards consideration.
and proposed change to remove the Attorney for licensee: M. Stanford
Electric Generating Plant, Units 3 and 4, pressurizer spray and surge line valve
Burke County, Georgia Blanton, Balch & Bingham LLP, 1710
leakage requirement as described in the Sixth Avenue North, Birmingham, AL
Date of amendment request: August current licensing basis do not have an
35203–2015.
31, 2016. A publicly-available version is adverse effect on any of the design functions
of the systems. The proposed changes do not
NRC Branch Chief: Jennifer Dixon-
in ADAMS under Accession No. Herrity.
affect the support, design, or operation of
ML16244A253. mechanical and fluid systems required to
Description of amendment request: III. Notice of Issuance of Amendments
mitigate the consequences of an accident. to Facility Operating Licenses and
The amendment request proposes There is no change to plant systems or the
changes to the Updated Final Safety response of systems to postulated accident Combined Licenses
Analysis Report (UFSAR) in the form of conditions. There is no change to the During the period since publication of
departures from the incorporated plant- predicted radioactive releases due to the last biweekly notice, the
specific Design Control Document postulated accident conditions. The plant Commission has issued the following
(DCD) Tier 2 information and a response to previously evaluated accidents or amendments. The Commission has
combined license (COL) License external events is not adversely affected, nor
do the proposed changes create any new
determined for each of these
Condition which references one of the accident precursors. amendments that the application
proposed changes. Additionally, the Therefore, the requested amendment does complies with the standards and
proposed changes to the UFSAR not involve a significant increase in the requirements of the Atomic Energy Act
eliminate pressurizer spray line probability or consequences of an accident of 1954, as amended (the Act), and the
monitoring during pressurizer surge line previously evaluated. Commission’s rules and regulations.
first plant only testing. In addition, 2. Does the proposed amendment create The Commission has made appropriate
these proposed changes correct the possibility of a new or different kind of
sradovich on DSK3GMQ082PROD with NOTICES
findings as required by the Act and the
inconsistencies in testing purpose, accident from any accident previously Commission’s rules and regulations in
evaluated?
testing duration, and the ability to leave 10 CFR chapter I, which are set forth in
Response: No.
equipment in place following the data The proposed changes for removing the the license amendment.
collection period. These changes requirement to install temporary A notice of consideration of issuance
involve material which is specifically instrumentation on the pressurizer spray line of amendment to facility operating
referenced in Section 2.D.(2) of the during the monitoring of the pressurizer license or combined license, as
COLs. This submittal requests approval surge line for thermal stratification and applicable, proposed no significant
VerDate Sep<11>2014 17:41 Feb 13, 2017 Jkt 241001 PO 00000 Frm 00041 Fmt 4703 Sfmt 4703 E:\FR\FM\14FEN1.SGM 14FEN1](https://thefederalregister.org/doc/82_FR_10619/page/13.svg)
![Federal Register / Vol. 82, No. 29 / Tuesday, February 14, 2017 / Notices 10605
2016, and October 26, 2016, provided No significant hazards consideration Required Actions, and Surveillance
additional information that clarified the comments received: No. Requirements.
application, did not expand the scope of Date of issuance: January 13, 2017.
Tennessee Valley Authority, Docket Effective date: As of the date of
the application as originally noticed,
Nos. 50–259, 50–260, and 50–296, issuance and shall be implemented
and did not change the staff’s original
Browns Ferry Nuclear Plant, Units 1, 2 within 60 days of issuance.
proposed no significant hazards
and 3, Limestone County, Alabama Amendment No.: 110 (Unit 1) and 5
consideration determination as
published in the Federal Register. Date of amendment request: August (Unit 2). A publicly-available version is
The Commission’s related evaluation 12, 2016. in ADAMS under Accession No.
of the amendments is contained in a Brief description of amendments: The ML17006A271; documents related to
Safety Evaluation dated January 13, amendments revised Technical this amendment are listed in the Safety
2017. Specification (TS) 4.3.1.2, ‘‘Fuel Storage Evaluation enclosed with the
No significant hazards consideration Criticality,’’ for Units 1, 2, and 3, to amendment.
comments received: No. preclude the placement of fuel in the Facility Operating License Nos. NPF–
new fuel storage vaults. This TS change 90 and NPF–96: Amendments revised
Susquehanna Nuclear, LLC, Docket No. the Facility Operating Licenses and
50–388, Susquehanna Steam Electric removed the existing TS 4.3.1.2
criticality criteria wording in its Technical Specifications.
Station, Unit 2, Luzerne County, Date of initial notice in Federal
Pennsylvania entirety, and replaced it with language
that specifically restricts the placement Register: May 24, 2016 (81 FR 32810).
Date of amendment request: January of fuel in the new fuel storage vaults. The supplement letters dated October
28, 2016, as supplemented by letters Date of issuance: January 17, 2017. 13, November 1, and December 8, 2016,
April 6, 2016, and October 10, 2016. provided additional information that
Effective date: As of the date of
Brief description of amendment: The clarified the application, did not expand
issuance and shall be implemented
amendment revised Technical the scope of the application as originally
within 60 days of issuance.
Specification (TS) 3.7.1, ‘‘Residual Heat noticed, and did not change the staff’s
Amendment Nos.: 296 (Unit 1), 320
Removal Service Water (RHRSW) original proposed no significant hazards
(Unit 2), and 280 (Unit 3). A publicly-
System and the Ultimate Heat Sink consideration determination as
available version is in ADAMS under
(UHS),’’ and TS 3.8.7, ‘‘Distribution published in the Federal Register.
Accession No. ML16330A158; The Commission’s related evaluation
Systems—Operating,’’ to increase the
documents related to these amendments of the amendment is contained in a
completion time for Conditions A and B
are listed in the Safety Evaluation Safety Evaluation dated January 13,
of TS 3.7.1, and Condition C of TS 3.8.7,
enclosed with the amendments. 2017.
from 72 hours to 7 days, in order to
Renewed Facility Operating License No significant hazards consideration
accommodate 480 volt engineered
Nos. DPR–33, DPR–52, and DPR–68: comments received: No.
safeguard system load center
Amendments revised the Renewed
transformer replacements on the Dated at Rockville, Maryland, this 6th day
Facility Operating Licenses and TSs.
Susquehanna Steam Electric Station, of February 2017.
Unit 1. The change is temporary and Date of initial notice in Federal
For the Nuclear Regulatory Commission.
will be annotated by a note in each TS Register: October 11, 2016 (81 FR
70187). Anne T. Boland,
that specifies the allowance expires on Director, Division of Operating Reactor
June 15, 2020. The Commission’s related evaluation
of the amendment is contained in a Licensing, Office of Nuclear Reactor
Date of issuance: January 26, 2017. Regulation.
Effective date: As of the date of Safety Evaluation dated January 17,
[FR Doc. 2017–02795 Filed 2–13–17; 8:45 am]
issuance and shall be implemented 2017.
No significant hazards consideration BILLING CODE 7590–01–P
within 30 days of issuance.
Amendment No.: 248. A publicly- comments received: No.
available version is in ADAMS under Tennessee Valley Authority, Docket NUCLEAR REGULATORY
Accession No. ML17004A250; Nos. 50–390 and 50–391, Watts Bar COMMISSION
documents related to this amendment Nuclear Plant, Units 1 and 2, Rhea
are listed in the Safety Evaluation [Docket Nos. 52–025 and 52–026; NRC–
County, Tennessee 2008–0252]
enclosed with the amendment.
Facility Operating License No. NPF– Date of amendment request:
22: The amendment revised the December 8, 2015, as supplemented by Southern Nuclear Operating Company,
Renewed Facility Operating License and letters dated March 11, October 13, Inc., Vogtle Electric Generating
TSs. December 1, and December 8, 2016. Station, Units 3 and 4; Fire Pump Head
Date of initial notice in Federal Brief description of amendment: The and Diesel Fuel Day Tank Changes
Register: May 24, 2016 (81 FR 32810). amendment revised the Watts Bar AGENCY: Nuclear Regulatory
The supplemental letter dated October Nuclear Plant, Units 1 and 2, Technical Commission.
10, 2016, provided additional Specification (TS) 3.8.1, ‘‘AC Sources— ACTION: Exemption and combined
information that clarified the Operating,’’ to extend the Completion license amendment issuance.
application, did not expand the scope of Time for one inoperable Diesel
the application as originally noticed, Generator from 72 hours to 10 days SUMMARY: The U.S. Nuclear Regulatory
based on the availability of a Commission (NRC) is granting an
sradovich on DSK3GMQ082PROD with NOTICES
and did not change the staff’s original
proposed no significant hazards supplemental alternating current power exemption to allow a departure from the
consideration determination as source (specifically, the FLEX DG added certification information of Tier 1 of the
published in the Federal Register. as part of the mitigating strategies for generic design control document (DCD)
The Commission’s related evaluation beyond-design-basis events in response and is issuing License Amendment No.
of the amendments is contained in a to NRC Order EA–12–049). The 58 to Combined Licenses (COL), NPF–
Safety Evaluation dated January 26, amendment also made clarifying 91 and NPF–92. The COLs were issued
2017. changes to certain TS 3.8.1 Conditions, to Southern Nuclear Operating
VerDate Sep<11>2014 16:48 Feb 13, 2017 Jkt 241001 PO 00000 Frm 00044 Fmt 4703 Sfmt 4703 E:\FR\FM\14FEN1.SGM 14FEN1](https://thefederalregister.org/doc/82_FR_10619/page/16.svg)
Document Created: 2017-02-13 23:40:32
Document Modified: 2017-02-13 23:40:32
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by March 16, 2017. A request for a hearing must be filed by April 17, 2017. | |
| Contact | Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-2242, email: [email protected] | |
| FR Citation | 82 FR 10590 |
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