83 FR 28456 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 83, Issue 118 (June 19, 2018)
Federal Register Volume 83, Issue 118 (June 19, 2018)
| Page Range | 28456-28467 | |
| FR Document | 2018-12506 |
[Federal Register Volume 83, Number 118 (Tuesday, June 19, 2018)] [Notices] [Pages 28456-28467] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2018-12506] [[Page 28456]] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2018-0114] 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 May 22, 2018, to June 4, 2018. The last biweekly notice was published on June 5, 2018. DATES: Comments must be filed by July 19, 2018. A request for a hearing must be filed by August 20, 2018. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114. Address questions about NRC dockets to Jennifer Borges; telephone: 301-287- 9127; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: May Ma, Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1384, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2018-0114, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods: Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114. NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [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-2018-0114, facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in section 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. [[Page 28457]] A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (First Floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued. As required by 10 CFR 2.309(d), the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest. In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures. Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [[Page 28458]] [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 website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system. A person filing electronically using the NRC's adjudicatory E- Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the ``Contact Us'' link located on the NRC's public website at http://www.nrc.gov/site-help/e-submittals.html, by email to [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 all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina Date of amendment request: January 23, 2018. A publicly-available version is in ADAMS under Accession No. ML18023A896. Description of amendment request: The proposed amendments would revise Technical Specification 3.6.4.1, ``Secondary Containment,'' Surveillance Requirement (SR) 3.6.4.1.2, for Brunswick Steam Electric Plant, Units 1 and 2. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, ``Revise Secondary Containment Surveillance Requirements'' (ADAMS Accession No. ML16277A226). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change addresses conditions during which Secondary Containment SR 3.6.4.1.2 is not met. The Secondary Containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed change is no different than the consequences of an accident while utilizing the existing eight hour Completion Time for an inoperable Secondary Containment. As a result, the consequences of an accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not alter the protection system design, create new failure [[Page 28459]] modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change addresses conditions during which Secondary Containment SR 3.6.4.1.2 is not met. The allowance for both an inner and outer Secondary Containment door to be open simultaneously for entry and exit does not affect the safety function of the Secondary Containment as the doors are promptly closed after entry or exit, thereby restoring the Secondary Containment boundary. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202. NRC Acting Branch Chief: Brian W. Tindell. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina Date of amendment request: January 23, 2018. A publicly-available version is in ADAMS under Accession No. ML18023A899. Description of amendment request: The amendments would revise the Technical Specifications to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, ``Extension of Time to Reach Mode 2 in LCO [Limiting Condition for Operation] 3.0.3.'' 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 time frame to take response action in accordance with LCO 3.0.3 is not an initiating condition for any accident previously evaluated. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The small increase in the time allowed to reach Mode 2 would not place the plant in any significantly increased probability of an accident occurring. The unit would already be preparing for a plant shutdown condition because of the 1 hour requirement to initiate shutdown actions. There is no change in the time period to reach Mode 3. The Mode 3 Condition is the point at which the plant reactor core is no longer critical (i.e., Hot Shutdown). Therefore, since there is no change to the time period to reach the Hot Shutdown condition, the small change in the time to reach Mode 2 status does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the allowed time to reach Mode 2 in LCO 3.0.3 does not require any modification to the plant or change equipment operation. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those that have been previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change to the allowed time to reach Mode 2 in LCO 3.0.3 does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, Appendix A will continue to be met. Therefore, the proposed change does not involve any reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202. NRC Acting Branch Chief: Brian W. Tindell. Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit No. 3 (Waterford 3), St. Charles Parish, Louisiana Date of amendment request: March 8, 2018. A publicly-available version is in ADAMS under Accession No. ML18068A705. Description of amendment request: The amendment would update Section 15.4.3.1 of the Updated Final Safety Analysis Report for Waterford 3, which describes the dose consequence of the worst undetectable single fuel assembly misload. The updated analysis would reflect the use of Next Generation Fuel and integrated fuel burnable absorbers. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises the fuel assembly misload event analysis. The analysis of the fuel assembly misload event showed that the total number of failed fuel rods is less than other Waterford 3 Condition 3 events that have already been demonstrated to meet the 10 CFR 50.67 acceptance criteria. For Waterford 3, the Excess Load with Loss of Alternating Current (LOAC) has this same release and fuel failure that has been shown to meet the offsite dose requirements. Since the worst undetectable misload has a fuel failure less than the excess load with LOAC event, the fuel assembly misload event is consistent with the Standard Review Plan 15.4.7 and meets the 10 CFR 50.67 requirements. This change is only analyzing the consequences of the fuel assembly misload event and no changes are being made that would impact the probability of the event occurring. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises the fuel assembly misload event analysis. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing plant operations. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose [[Page 28460]] consequences exceed the consequences of accidents previously analyzed. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises the fuel assembly misload event analysis. The worst undetectable misloads have fuel failure less than the excess load with the Excess Load with Loss of Alternating Current (LOAC) event; the fuel assembly misload event meets the 10 CFR 50.67 criteria and is consistent with the Standard Review Plan Section 15.4.7 guidance. The new analysis shows more adverse consequences than were shown in previous fuel assembly misload event analyses, but remains within the regulatory acceptance limits. Since the event remains within the 10 CFR 50.67 requirements and is bounded by the excess load with LOAC event, this is not a significant reduction in margin. Therefore, this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Unit Nos. 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Unit Nos. 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Unit Nos. 1 and 2, Rock Island County, Illinois Date of amendment request: April 25, 2018. A publicly-available version is in ADAMS under Accession No. ML18116A133. Description of amendment request: The amendments would revise the technical specification (TS) requirements for inoperable snubbers for each facility. The amendments would also make other administrative changes to the TSs. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration for each site, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change allows a delay time before declaring supported Technical Specification (TS) systems inoperable when the associated snubber(s) cannot perform its required safety function. Entrance into Actions or delaying entrance into Actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on the delay time allowed before declaring a TS supported system inoperable and taking its Conditions and Required Actions are no different than the consequences of an accident under the same plant conditions while relying on the existing TS supported system Conditions and Required Actions. Therefore, the consequences of an accident previously evaluated are not significantly increased by this change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change restores an allowance in the pre-Improved Standard Technical Specifications (ISTS) conversion TS that was unintentionally eliminated by the conversion. The pre-ISTS TS were considered to provide an adequate margin of safety for plant operation, as does the post-ISTS conversion TS. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis for each site 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. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: March 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18066A648. Description of amendment request: The proposed amendment would revise Technical Specification 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to follow guidance developed by the Nuclear Energy Institute (NEI) in topical report NEI 94-01, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' Revision 3-A, dated July 2012, with the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008. The proposed license amendment would also revise Technical Specification 5.5.12 by deleting two of the four listed exceptions to program guidelines. 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 test interval extensions do not involve either a physical change to the plant or a change in the way 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. Therefore, the proposed extensions do not involve a significant increase in the probability of an accident previously evaluated. The effect resulting from changing the Type A test frequency to 1 per 15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is [[Page 28461]] 0.0318 person-rem/year. EPRI [Electric Power Research Institute] Report No. 1009325, Revision 2-A, states that a very small population dose is defined as an increase of less than or equal to 1.0 person-rem per year or less than or equal to 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended integrated leak rate test intervals. The results of the risk assessment calculation for the Type A test extension meet these criteria. The risk impact for the integrated leak rate test extension when compared to other severe accident risks is negligible. 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 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 [American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)], Section XI, and Technical Specification requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes two previously granted exceptions to Primary Containment Leakage Rate Testing Program guidelines. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do not affect the probability or consequences of an accident previously evaluated. Therefore, the proposed changes do 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. Containment Type A and Type C testing requirements periodically demonstrate the integrity of the containment and exist to ensure the plant's ability to mitigate the consequences of an accident. These tests do not involve any accident precursors or initiators. The proposed change does not involve a physical modification to the plant (that is, no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment. The proposed amendment also deletes two previously granted exceptions. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do not create the possibility of a new or different kind of accident from any previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed license amendment does not alter the way safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the Technical Specification Primary Containment Leakage 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 Technical Specifications 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 amendment, since they are not affected by implementation of a performance-based containment testing program. This ensures that the margin of safety in the plant safety analysis is maintained. The proposed amendment also deletes two previously granted exceptions. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do 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: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: David J. Wrona. NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant (PBNP), Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: March 30, 2018. A publicly-available version is in ADAMS under Accession No. ML18092A239. Description of amendment request: The amendments would revise Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing Program,'' to require a program in accordance with Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J.'' This proposed change will allow extension of the Type A test interval up to one test in 15 years and extension of the Type C test interval up to 75 months, based on acceptable performance history as defined in NEI 94-01, Revision 3-A. 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 consequences of an accident previously evaluated? Response: No. The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' for development of the PBNP performance-based containment testing program. NEI 94-01 allows, based on risk and performance, an extension of Type A and Type C containment leak test intervals. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses. The findings of the PBNP risk assessment confirm the general findings of previous studies that the risk impact with extending the containment leak rate is small. Per the guidance provided in Regulatory Guide 1.174, an extension of the leak test interval in accordance with NEI 94-01, Revision 3-A results in an estimated change within, the very small change region. Since the change is implementing a performance-based containment testing program, the proposed amendment 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 requirement for containment leakage rate acceptance will not be changed by this amendment. [[Page 28462]] Therefore, the containment will continue to perform its design function as a barrier to fission product releases. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not change the design or operation of structures, systems, or components of the plant. The proposed change would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by this change. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test and local leak rate testing frequency, does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation. The specific requirements and conditions of the TS Containment Leakage 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. This 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 implementation of a performance-based containment testing program. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station (HCGS), Salem County, New Jersey Date of amendment request: April 13, 2018. A publicly-available version is in ADAMS under Accession No. ML18103A218. Description of amendment request: The amendment would revise Technical Specification (TS) 3.8.3.1, ``Distribution--Operating,'' to increase the alternating current (AC) inverters allowed outage time (AOT) from 24 hours to 7 days. The proposed change is based on application of the HCGS probabilistic risk assessment (PRA) in support of a risk-informed extension, and on additional considerations and compensatory actions. 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 TS amendment does not affect the design of the AC inverters, the operational characteristics or function of the inverters, the interfaces between the inverters and other plant systems, or the reliability of the inverters. An inoperable AC inverter is not considered an initiator of an analyzed event. In addition, TS Actions and the associated Allowed Outage Times are not initiators of previously evaluated accidents. Extending the Allowed Outage Time for an inoperable AC inverter would not have a significant impact on the frequency of occurrence of an accident previously evaluated. The proposed amendment will not result in modifications to plant activities associated with inverter maintenance, but rather, provides operational flexibility by allowing additional time to perform inverter troubleshooting, corrective maintenance, and post-maintenance testing on-line. The proposed extension of the Completion Time for an inoperable AC inverter will not significantly affect the capability of the inverters to perform their safety function, which is to ensure an uninterruptible supply of 120-volt AC electrical power to the associated power distribution subsystems. An evaluation, using PRA methods, confirmed that the increase in plant risk associated with implementation of the proposed Allowed Outage Time extension is consistent with the NRC's Safety Goal Policy Statement, as further described in RG [Regulatory Guide] 1.174 and RG 1.177. In addition, a deterministic evaluation concluded that plant defense-in-depth philosophy will be maintained with the proposed Allowed Outage Time extension. There will be no impact on the source term or pathways assumed in accidents previously evaluated. No analysis assumptions will be changed and there will be no adverse effects on onsite or offsite doses as the result of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment does not involve physical alteration of the HCGS. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters with in which the HCGS is operated. There are no setpoints at which protective or mitigating actions are initiated that are affected by this proposed action. The use of the alternate Class 1E power source for the AC distribution panel is consistent with the HCGS plant design. The change does not alter assumptions made in the safety analysis. This proposed action will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alteration is proposed to the procedures that ensure the HCGS remains with in analyzed limits, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change, which would increase the AOT from 24 hours to 7 days for one inoperable inverter, does not exceed or alter a setpoint, design basis or safety limit. 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. [[Page 28463]] Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: James G. Danna. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia Date of amendment request: April 26, 2018. A publicly-available version is in ADAMS under Accession No. ML18116A138. Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix C, with corresponding changes to the associated plant-specific Tier 1 information, and involves associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document (DCD) Tier 2 information). Pursuant to the provisions of 10 CFR 52.63(b)(1), also requested is an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule for the plant-specific DCD departures. The requested amendment proposes changes to COL Appendix C (and plant-specific Tier 1) to reflect a new design of containment sump level sensors that affects the acceptance criterion for the detected containment sump level change test and the associated minimum detectable unidentified leakage rate in plant-specific DCD Tier 2 information. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change is to the containment sump water level instrumentation and its expected [reactor coolant system (RCS)] leakage detection capability. The affected equipment is not safety- related, but the containment sump water level sensors are seismically qualified. The change in containment sump level monitoring instruments has no adverse effect on the ability to detect a 0.5 [gallons per minute (gpm)] leak in containment, and therefore, has no adverse effect on design criteria for leak-before- break. 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. Because the containment sump water level monitoring channels are still capable of detecting a 0.5 gpm leak in containment, the change to the SSC has no effect on plant operations. 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. 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 the containment sump water level instrumentation and its expected RCS leakage detection capability has no adverse effect on the ability to detect a 0.5 gpm leak in containment. The containment sump level instrumentation functions are unchanged and leak-before-break design criteria are not adversely affected. Loss of coolant accidents for a spectrum of pipe sizes and locations are already postulated in UFSAR Chapter 15, Section 15.6. Breaks in the main steam lines inside containment are also analyzed in UFSAR Chapter 15, Section 15.1. Unidentified leakage detection and operator action in response to unidentified leakage are not postulated for any of the design basis accident analyses described in UFSAR Chapter 15. Therefore, the requested amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The described change to the containment sump water level instrumentation and its expected RCS leakage detection capability is proposed to verify that the ability to detect a 0.5 gpm leak in containment is maintained. The proposed change does not alter any safety-related equipment, applicable design codes, code compliance, design function, or safety analysis. By ensuring that the chosen equipment can detect a 0.5 gpm leak in containment with the described accuracy, guidance in Regulatory Guide 1.45, Revision 0, as committed to in the UFSAR, and requirements in the Technical Specifications are met which ensures that leak-before-break design criteria are not adversely affected. 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 hazard consideration Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue, North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer L. Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia Date of amendment request: April 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18117A464. Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document Tier 2 information) and involves related changes to plant- specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. Specifically, the amendment, if approved, would revise the Tier 2 information in the UFSAR and related changes to Tier 1 and the associated COL Appendix C to remove the fire protection system non-safety related containment cable spray and install passive fire stops and radiant energy shields. The changes to Tier 1 require an exemption, which is included in the license amendment request. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes do not affect the operation or reliability of any system, structure or component (SSC) required to maintain a normal power operating condition or to mitigate anticipated transients without safety-related systems. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones. The proposed changes do not affect the operation of equipment whose failure could initiate an accident previously analyzed. The existence or failure of passive fire stops in fire zone 1100 AF 11300B does not affect normal equipment operation. [[Page 28464]] The proposed changes do not adversely affect the reliability or function of an SSC relied upon to mitigate an accident previously analyzed. The existence or failure of passive fire stops in fire zone 1100 AF 11300B will not adversely affect passive core cooling system (PXS) performance during containment recirculation because the passive fire stops are located outside of the zone of influence (ZOI) of postulated high energy line breaks, and the passive fire stops' material-of-construction complies with in-containment refueling water storage tank (IRWST) and containment recirculation screens design criteria for debris generation and transport. The existing active open nozzle cable tray suppression system is not fully automatic, is nonsafety-related, and is not credited in the probabilistic risk assessment (PRA). Therefore, replacing the active open nozzle cable tray suppression system with passive fire stops does not have an impact on PRA calculations and results. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not affect the operation of systems or equipment that could 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 use of passive fire stops is recognized by Regulatory Guide 1.189. The passive fire stops in nonsafety-related open cable trays are more reliable than active systems such as the current open nozzle cable tray suppression system because they require no mechanical or human action to perform their protective function. When protection is required, there is no delay for operator or mechanical response. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones. The existence or failure of passive fire stops in fire zone 1100 AF 11300B will not adversely affect passive core cooling system (PXS) performance during containment recirculation because the passive fire stops are located outside of the zone of influence (ZOI) of postulated high energy line breaks, and their material-of- construction complies with in-containment refueling water storage tank (IRWST) and containment recirculation screens design criteria for debris generation and transport. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not affect existing safety margins. The current open nozzle cable tray suppression system is nonsafety- related. The use of passive fire stops is recognized by Regulatory Guide 1.189. The passive fire stops in nonsafety-related open cable trays are more reliable than active systems such as the current open nozzle cable tray suppression system because they require no mechanical or human action to perform their protective function. When protection is required, there is no delay for operator or mechanical response. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and based on this review it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazard consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue, North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer L. Dixon-Herrity. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia Date of amendment request: April 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18117A464. Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document Tier 2 information) and involves related changes to plant- specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. Specifically, the amendment, if approved, would revise the Tier 2 information in the UFSAR and related changes to Tier 1 and the associated COL Appendix C to remove the fire protection system non-safety related containment cable spray and install passive fire stops and radiant energy shields. The changes to Tier 1 require an exemption, which is included in the license amendment request. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events. The proposed changes do not affect the physical design and operation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) or In-containment Refueling Water Storage Tank (IRWST) as described in the Updated Final Safety Analysis Report (UFSAR). The proposed changes do not affect the probability of inadvertent operation or failure. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected. The proposed changes do not affect the ability of the PRHR HX and IRWST to perform their design functions. The designs of the PRHR HX and IRWST continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the PRHR HX and IRWST to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria. The proposed changes do not affect the prevention and mitigation of other abnormal events (e.g. anticipated operational occurrences, earthquakes, floods and turbine missiles), or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes do not affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety- related or nonsafety related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures. Therefore, the requested amendment does not create the possibility of a new or different type of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes maintain existing safety margins. The proposed changes verify and maintain the capabilities of the PRHR HX and IRWST to perform their design functions. Therefore, the proposed changes [[Page 28465]] satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the requested amendment does not involve a significant reduction in 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 L. Dixon-Herrity. Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2 (Surry), Surry County, Virginia Date of amendment request: March 2, 2018. A publicly-available version is in ADAMS under Accession No. ML18075A021. Description of amendment request: The amendments would revise the Technical Specifications (TSs) consistent with Revision 0 to the Technical Specifications Task Force (TSTF) Standard Technical Specification Change Document TSTF-490, ``Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.'' The proposed amendments would adopt TSTF-490 and make the following associated changes: (1) Adoption of a TS change to replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity, and (2) an update of the Alternative Source Term (AST) analyses for Surry. 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: Criterion 1. The Proposed Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated Reactor coolant specific activity is not an initiator for any accident previously evaluated, and the allowed time period when primary coolant gross activity is not within limits is not an initiator for any accident previously evaluated. In addition, the current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. Updating the Alternative Source Term analyses does not require any changes to any plant structures, systems, or components (SSCs) and therefore does not affect any accident initiators. As a result, the proposed changes do not significantly increase the probability of an accident. The proposed TS change will limit primary coolant noble gases to concentrations consistent with the accident analyses, and the proposed completion time when the limit may be exceeded has no impact on the consequences of any design basis accident since the consequences of an accident during this time period is the same as the consequences of an accident during the existing time periods. The revised assessments of the radiological consequences due to design basis accidents listed in the Surry Updated Final Safety Analysis Report, using the updated AST methodology and proposed assumptions and inputs, conclude that the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room doses are within the limits of 10 CFR 50.67 and within the limits of Regulatory Guide (RG) 1.183. As a result, the consequences of any accident previously evaluated are not significantly increased. Criterion 2. The Proposed Changes Do Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated The proposed TS change in specific activity limits and the updated AST dose consequences analyses do not alter any physical part of the plant, (i.e., no new or different type of equipment will be installed,) nor do they affect any plant operating parameter or create new accident precursors. Therefore, the proposed changes do not create the potential for a new or different kind of accident from any previously calculated. Criterion 3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety. The proposed TS change in specific activity limits is consistent with the assumptions in -the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses. The proposed changes for radiological events related to the computer code used to calculate dose, revised X/Qs for control room and offsite receptors (including the computer code and method used to determine control room X/Qs for SG releases), the computer code used to determine core inventory, the change in FHA [Fuel Handling Accident] gap fraction methodology, and removing the LRA [Locked Rotor Accident] from the radiological design basis have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10 CFR 50.67 and RG 1.183. The proposed changes will not result in plant operation in a configuration outside the analyses or design basis and do not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition. Therefore, the 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219. NRC Branch Chief: Michael T. Markley. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. [[Page 28466]] DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan Date of amendment request: August 14, 2017. Brief description of amendment: The amendment modified Fermi 2 Technical Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' by adopting the format and language of NUREG-1433, ``Standard Technical Specifications for General Electric BWR/4 Plants,'' Revision 4. Date of issuance: May 24, 2018. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No: 208. A publicly-available version is in ADAMS under Accession No. ML18108A022; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-43: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: September 26, 2017 (82 FR 44851). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2018. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina Date of amendment request: June 29, 2017, as supplemented by letters dated January 4, 2018, and January 23, 2018. Brief description of amendments: The amendments adopted Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' for Brunswick Steam Electric Plant, Units 1 and 2. The amendments replaced existing technical specification (TS) requirements associated with ``operations with the potential for draining the reactor vessel,'' with revised TSs providing alternative requirements for reactor pressure vessel water inventory control. These alternative requirements protect Safety Limit 2.1.1.3, which states, ``Reactor vessel water level shall be greater than the top of active irradiated fuel.'' Date of issuance: April 13, 2018. Effective date: As of the date of issuance and shall be implemented prior to the 2019 Unit 2 refueling outage. This Notice of Issuance corrects the effective date of License Amendment No. 283, originally noticed in the Federal Register on May 8, 2018 (83 FR 20865). Amendment Nos.: 283 (Unit 1) and 311 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18039A444; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments. Amendment Nos. 283 and 311 were corrected by letter dated May 23, 2018 (ADAMS Accession No. ML18137A143). Renewed Facility Operating License No. DPR-49: The amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: September 12, 2017 (82 FR 42846). The supplemental letters dated January 4, 2018, and January 23, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety evaluation dated April 13, 2018. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of amendment request: April 3, 2017, as supplemented by letters dated April 3, 2017; May 2, 2017; September 28, 2017; and January 8, 2018. Brief description of amendment: The amendment revised the Technical Specifications (TSs) to extend the required frequency of certain 18- month Surveillance Requirements to 24 months to accommodate a 24-month refueling cycle. In addition, the amendment revised certain programs in TS Section 5.5, ``Programs and Manuals,'' to change 18-month frequencies to 24 months. Date of issuance: May 25, 2018. Effective date: As of the date of issuance and shall be implemented within 120 days from the end of the next refueling outage. Amendment No.: 258. A publicly-available version is in ADAMS under Accession No. ML18115A150; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-23: The amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: July 5, 2017 (82 FR 31092). The supplemental letters dated September 28, 2017, and January 8, 2018, 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 May 25, 2018. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1 (Clinton), DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LaSalle), Unit Nos. 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station (Limerick), Unit Nos. 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit No. 2 (Nine Mile), Oswego County, New York Date of amendment request: November 8, 2017. Brief description of amendments: The amendments revised the technical specification requirements for secondary containment. Date of issuance: May 29, 2018. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: Clinton--218; LaSalle, Units 1 and 2--228 and 214; Limerick, Units 1 and 2--229 and 192; and Nine Mile--169. A publicly- available version is in ADAMS under Accession No. ML18113A045. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-62, NPF-11, NPF-18, NPF-39, NPF-85, and NPF-69: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: December 19, 2017 (82 FR 60227). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 29, 2018. No significant hazards consideration comments received: No. [[Page 28467]] Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: April 20, 2017, as supplemented by letters dated September 14, 2017; February 19, 2018; and May 1, 2018. Brief description of amendments: The amendments revised the Technical Specifications by replacing the existing requirements related to ``operations with a potential for draining the reactor vessel'' with new requirements on Reactor Pressure Vessel Water Inventory Control to protect Safety Limit 2.1.1.3, which requires reactor vessel water level to be greater than the top of active irradiated fuel. Date of issuance: May 31, 2018. Effective date: As of the date of issuance and shall be implemented prior to the commencement of the Unit No. 2 refueling outage (U2R25) in February 2019. Amendment Nos.: Unit 1--290, Unit 2--235. A publicly-available version is in ADAMS under Accession No. ML18123A368; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-57 and NPF-5: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: August 29, 2017 (82 FR 41071). The supplemental letters dated September 14, 2017; February 19, 2018; and May 1, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 31, 2018. No significant hazards consideration comments received: No. Tennessee Valley Authority (TVA) Docket Nos. 50-259, 50-260, 50-296, and 72-052, Browns Ferry Nuclear Plant, Unit Nos. 1, 2, and 3, Limestone County, Alabama TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant, Unit Nos. 1 and 2, Rhea County, Tennessee Date of amendment request: January 4, 2017, as supplemented by letters dated July 7, 2017, and July 27, 2017. (Note: This Notice of Issuance corrects the amendments by adding the supplement dated July 27, 2017, which was inadvertently omitted from the original Federal Register notice (January 16, 2018; 83 FR 2234). Brief description of amendments: The amendments revised TVA Emergency Plans for the above nuclear plants. Specifically, the amendments adopted the NRC-endorsed Radiological Emergency Plan Emergency Action Level schemes developed by the Nuclear Energy Institute (NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors''). Date of issuance: December 22, 2017. Effective date: As of the date of issuance and shall be implemented within 180 days from the date of its issuance, or July 3, 2018, whichever comes later. Amendment Nos.: Browns Ferry Nuclear Plant--303 (Unit 1), 327 (Unit 2), and 287 (Unit 3); Sequoyah Nuclear Plant--339 (Unit 1) and 332 (Unit 2); and Watts Bar Nuclear Plant--118 (Unit 1) and 18 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17289A032; documents related to these amendments are listed in the Safety Evaluations enclosed with the amendments. These amendments were corrected by letter dated May 29, 2018 (ADAMS Accession No. ML18138A452). Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68, DPR-77, and DPR-79, and Facility Operating License Nos, NPF-90 and NPF- 96: The amendments revised the licenses. Date of initial notice in Federal Register: June 19, 2017 (82 FR 27891). The supplemental letters dated July 7, 2017, and July 27, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 22, 2017. No significant hazards consideration comments received: No. Union Electric Company, Docket No. 50-483, Callaway Plant, Unit No. 1, Callaway County, Missouri Date of amendment request: April 6, 2017, as supplemented by letter dated February 5, 2018. Brief description of amendment: The amendment revised the Final Safety Analysis Report to clearly describe conformance with NRC Regulatory Guide 1.106, Revision 1, ``Thermal Overload Protection for Electric Motors on Motor-Operated Valves.'' Date of issuance: May 30, 2018. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 218. A publicly-available version is in ADAMS under Accession No. ML18124A026; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-30: The amendment revised the Final Safety Analysis Report. Date of initial notice in Federal Register: July 18, 2017 (82 FR 32885). The supplemental letter dated February 5, 2018, 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 May 30, 2018. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 6th day of June 2018. For the Nuclear Regulatory Commission. Tara Inverso, Acting Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2018-12506 Filed 6-18-18; 8:45 am] BILLING CODE 7590-01-P
![28456 Federal Register / Vol. 83, No. 118 / Tuesday, June 19, 2018 / Notices
NUCLEAR REGULATORY Reactor Regulation, U.S. Nuclear Your request should state that the NRC
COMMISSION Regulatory Commission, Washington, does not routinely edit comment
DC 20555–0001; telephone: 301–415– submissions to remove such information
[NRC–2018–0114]
1384, email: Janet.Burkhardt@nrc.gov. before making the comment
Biweekly Notice; Applications and SUPPLEMENTARY INFORMATION: submissions available to the public or
Amendments to Facility Operating entering the comment into ADAMS.
I. Obtaining Information and
Licenses and Combined Licenses Submitting Comments II. Notice of Consideration of Issuance
Involving No Significant Hazards of Amendments to Facility Operating
Considerations A. Obtaining Information
Licenses and Combined Licenses and
AGENCY: Nuclear Regulatory Please refer to Docket ID NRC–2018– Proposed No Significant Hazards
Commission. 0114, facility name, unit number(s), Consideration Determination
plant docket number, application date,
ACTION: Biweekly notice.
and subject when contacting the NRC The Commission has made a
SUMMARY: Pursuant to Section 189a.(2) about the availability of information for proposed determination that the
of the Atomic Energy Act of 1954, as this action. You may obtain publicly- following amendment requests involve
amended (the Act), the U.S. Nuclear available information related to this no significant hazards consideration.
Regulatory Commission (NRC) is action by any of the following methods: Under the Commission’s regulations in
publishing this regular biweekly notice. • Federal Rulemaking website: Go to section 50.92 of title 10 of the Code of
The Act requires the Commission to http://www.regulations.gov and search Federal Regulations (10 CFR), this
publish notice of any amendments for Docket ID NRC–2018–0114. means that operation of the facility in
issued, or proposed to be issued, and • NRC’s Agencywide Documents accordance with the proposed
grants the Commission the authority to Access and Management System amendment would not (1) involve a
issue and make immediately effective (ADAMS): You may obtain publicly- significant increase in the probability or
any amendment to an operating license available documents online in the consequences of an accident previously
or combined license, as applicable, ADAMS Public Documents collection at evaluated, or (2) create the possibility of
upon a determination by the http://www.nrc.gov/reading-rm/ a new or different kind of accident from
Commission that such amendment adams.html. To begin the search, select any accident previously evaluated; or
involves no significant hazards ‘‘ADAMS Public Documents’’ and then (3) involve a significant reduction in a
consideration, notwithstanding the select ‘‘Begin Web-based ADAMS margin of safety. The basis for this
pendency before the Commission of a Search.’’ For problems with ADAMS, proposed determination for each
request for a hearing from any person. please contact the NRC’s Public amendment request is shown below.
This biweekly notice includes all Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by The Commission is seeking public
notices of amendments issued, or
email to pdr.resource@nrc.gov. The comments on this proposed
proposed to be issued, from May 22,
ADAMS accession number for each determination. Any comments received
2018, to June 4, 2018. The last biweekly
document referenced (if it is available in within 30 days after the date of
notice was published on June 5, 2018.
ADAMS) is provided the first time that publication of this notice will be
DATES: Comments must be filed by July
it is mentioned in this document considered in making any final
19, 2018. A request for a hearing must
• NRC’s PDR: You may examine and determination.
be filed by August 20, 2018.
purchase copies of public documents at Normally, the Commission will not
ADDRESSES: You may submit comments the NRC’s PDR, Room O1–F21, One
by any of the following methods (unless issue the amendment until the
White Flint North, 11555 Rockville expiration of 60 days after the date of
this document describes a different Pike, Rockville, Maryland 20852.
method for submitting comments on a publication of this notice. The
specific subject): B. Submitting Comments Commission may issue the license
• Federal Rulemaking website: Go to amendment before expiration of the 60-
Please include Docket ID NRC–2018–
http://www.regulations.gov and search day period provided that its final
0114, facility name, unit number(s),
for Docket ID NRC–2018–0114. Address determination is that the amendment
plant docket number, application date,
questions about NRC dockets to Jennifer involves no significant hazards
and subject in your comment
Borges; telephone: 301–287–9127; submission. consideration. In addition, the
email: Jennifer.Borges@nrc.gov. For The NRC cautions you not to include Commission may issue the amendment
technical questions, contact the identifying or contact information that prior to the expiration of the 30-day
individual listed in the FOR FURTHER you do not want to be publicly comment period if circumstances
INFORMATION CONTACT section of this disclosed in your comment submission. change during the 30-day comment
document. The NRC will post all comment period such that failure to act in a
• Mail comments to: May Ma, Office submissions at http:// timely way would result, for example in
of Administration, Mail Stop: TWFN–7– www.regulations.gov as well as enter the derating or shutdown of the facility. If
A60M, U.S. Nuclear Regulatory comment submissions into ADAMS. the Commission takes action prior to the
Commission, Washington, DC 20555– The NRC does not routinely edit expiration of either the comment period
0001. comment submissions to remove or the notice period, it will publish in
For additional direction on obtaining identifying or contact information. the Federal Register a notice of
daltland on DSKBBV9HB2PROD with NOTICES
information and submitting comments, If you are requesting or aggregating issuance. If the Commission makes a
see ‘‘Obtaining Information and comments from other persons for final no significant hazards
Submitting Comments’’ in the submission to the NRC, then you should consideration determination, any
SUPPLEMENTARY INFORMATION section of inform those persons not to include hearing will take place after issuance.
this document. identifying or contact information that The Commission expects that the need
FOR FURTHER INFORMATION CONTACT: they do not want to be publicly to take this action will occur very
Janet Burkhardt, Office of Nuclear disclosed in their comment submission. infrequently.
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![Federal Register / Vol. 83, No. 118 / Tuesday, June 19, 2018 / Notices 28459
modes, or change any modes of operation. initiating condition for any accident NRC Acting Branch Chief: Brian W.
The proposed change does not involve a previously evaluated. The proposed change Tindell.
physical alteration of the plant; and no new does not authorize the addition of any new
or different kind of equipment will be plant equipment or systems, nor does it alter Entergy Operations, Inc., Docket No. 50–
installed. Consequently, there are no new the assumptions of any accident analyses. 382, Waterford Steam Electric Station,
initiators that could result in a new or The small increase in the time allowed to Unit No. 3 (Waterford 3), St. Charles
different kind of accident. reach Mode 2 would not place the plant in Parish, Louisiana
Therefore, the proposed change does not any significantly increased probability of an
create the possibility of a new or different accident occurring. The unit would already Date of amendment request: March 8,
kind of accident from any previously be preparing for a plant shutdown condition 2018. A publicly-available version is in
evaluated. because of the 1 hour requirement to initiate ADAMS under Accession No.
3. Does the proposed amendment involve shutdown actions. There is no change in the ML18068A705.
a significant reduction in a margin of safety? time period to reach Mode 3. The Mode 3 Description of amendment request:
Response: No. Condition is the point at which the plant The amendment would update Section
The proposed change addresses conditions reactor core is no longer critical (i.e., Hot
during which Secondary Containment SR 15.4.3.1 of the Updated Final Safety
Shutdown). Analysis Report for Waterford 3, which
3.6.4.1.2 is not met. The allowance for both Therefore, since there is no change to the
an inner and outer Secondary Containment describes the dose consequence of the
time period to reach the Hot Shutdown
door to be open simultaneously for entry and condition, the small change in the time to worst undetectable single fuel assembly
exit does not affect the safety function of the reach Mode 2 status does not involve a misload. The updated analysis would
Secondary Containment as the doors are significant increase in the probability or reflect the use of Next Generation Fuel
promptly closed after entry or exit, thereby and integrated fuel burnable absorbers.
consequences of an accident previously
restoring the Secondary Containment Basis for proposed no significant
evaluated.
boundary.
Therefore, the proposed change does not
2. Does the proposed change create the hazards consideration determination:
possibility of a new or different kind of As required by 10 CFR 50.91(a), the
involve a significant reduction in a margin of
safety. accident from any accident previously licensee has provided its analysis of the
evaluated? issue of no significant hazards
The NRC staff has reviewed the Response: No.
consideration, which is presented
licensee’s analysis and, based on this The proposed change to the allowed time
to reach Mode 2 in LCO 3.0.3 does not below:
review, it appears that the three
standards of 10 CFR 50.92(c) are require any modification to the plant or 1. Does the proposed change involve a
satisfied. Therefore, the NRC staff change equipment operation. The proposed significant increase in the probability or
change will not introduce failure modes that consequences of an accident previously
proposes to determine that the
could result in a new accident, and the evaluated?
amendment request involves no change does not alter assumptions made in Response: No.
significant hazards consideration. the safety analysis. The proposed change will The proposed change revises the fuel
Attorney for licensee: Kathryn B. not alter the design configuration, or method assembly misload event analysis. The
Nolan, Deputy General Counsel, 550 of operation of plant equipment beyond its analysis of the fuel assembly misload event
South Tryon Street, Mail Code DEC45A, normal functional capabilities. The proposed showed that the total number of failed fuel
Charlotte, NC 28202. change does not create any new credible rods is less than other Waterford 3 Condition
NRC Acting Branch Chief: Brian W. failure mechanisms, malfunctions, or 3 events that have already been demonstrated
Tindell. accident initiators. to meet the 10 CFR 50.67 acceptance criteria.
Therefore, the proposed change does not For Waterford 3, the Excess Load with Loss
Duke Energy Progress, LLC, Docket Nos. create the possibility of a new or different of Alternating Current (LOAC) has this same
50–325 and 50–324, Brunswick Steam kind of accident from those that have been release and fuel failure that has been shown
Electric Plant, Unit Nos. 1 and 2, previously evaluated. to meet the offsite dose requirements. Since
Brunswick County, North Carolina 3. Does the proposed change involve a the worst undetectable misload has a fuel
significant reduction in a margin of safety? failure less than the excess load with LOAC
Date of amendment request: January Response: No. event, the fuel assembly misload event is
23, 2018. A publicly-available version is The proposed change to the allowed time consistent with the Standard Review Plan
in ADAMS under Accession No. to reach Mode 2 in LCO 3.0.3 does not alter 15.4.7 and meets the 10 CFR 50.67
ML18023A899. or exceed a design basis or safety limit. There requirements.
Description of amendment request: is no change being made to safety analysis This change is only analyzing the
The amendments would revise the assumptions or the safety limits that would consequences of the fuel assembly misload
Technical Specifications to adopt adversely affect plant safety as a result of the event and no changes are being made that
proposed change. Margins of safety are would impact the probability of the event
Technical Specifications Task Force occurring.
(TSTF) Traveler TSTF–208, Revision 0, unaffected by the proposed change and the
applicable requirements of 10 CFR Therefore, the proposed change does not
‘‘Extension of Time to Reach Mode 2 in 50.36(c)(2)(ii) and 10 CFR 50, Appendix A involve a significant increase in the
LCO [Limiting Condition for Operation] will continue to be met. probability or consequences of an accident
3.0.3.’’ Therefore, the proposed change does not previously evaluated.
Basis for proposed no significant involve any reduction in a margin of safety. 2. Does the proposed change create the
hazards consideration determination: possibility of a new or different kind of
As required by 10 CFR 50.91(a), the The NRC staff has reviewed the accident from any accident previously
licensee has provided its analysis of the licensee’s analysis and, based on this evaluated?
issue of no significant hazards review, it appears that the three Response: No.
standards of 10 CFR 50.92(c) are The proposed change revises the fuel
consideration, which is presented assembly misload event analysis. The
below: satisfied. Therefore, the NRC staff
daltland on DSKBBV9HB2PROD with NOTICES
proposes to determine that the proposed change does not involve a physical
1. Does the proposed change involve a alteration of the plant (no new or different
amendment request involves no
significant increase in the probability or type of equipment will be installed) or a
consequences of an accident previously
significant hazards consideration. change in the methods governing plant
evaluated? Attorney for licensee: Kathryn B. operations. The proposed change will not
Response: No. Nolan, Deputy General Counsel, 550 introduce new failure modes or effects and
The time frame to take response action in South Tryon Street, M/C DEC45A, will not, in the absence of other unrelated
accordance with LCO 3.0.3 is not an Charlotte, NC 28202. failures, lead to an accident whose
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![Federal Register / Vol. 83, No. 118 / Tuesday, June 19, 2018 / Notices 28461
0.0318 person-rem/year. EPRI [Electric Power configuration, or change the manner in NRC Branch Chief: David J. Wrona.
Research Institute] Report No. 1009325, which the plant is operated or controlled
Revision 2–A, states that a very small beyond the standard functional capabilities NextEra Energy, Point Beach, LLC,
population dose is defined as an increase of of the equipment. Docket Nos. 50–266 and 50–301, Point
less than or equal to 1.0 person-rem per year The proposed amendment also deletes two Beach Nuclear Plant (PBNP), Unit Nos.
or less than or equal to 1 percent of the total previously granted exceptions. The exception 1 and 2, Town of Two Creeks,
population dose, whichever is less restrictive regarding the performance of a Type A test Manitowoc County, Wisconsin
for the risk impact assessment of the no later than a specified date would be
extended integrated leak rate test intervals. deleted as this Type A test has already been Date of amendment request: March
The results of the risk assessment calculation performed. Additionally, the exception to use 30, 2018. A publicly-available version is
for the Type A test extension meet these the corrections to NEI 94–01, Revision 0, in ADAMS under Accession No.
criteria. The risk impact for the integrated would be deleted as those corrections would ML18092A239.
leak rate test extension when compared to no longer be in use. These changes to the Description of amendment request:
other severe accident risks is negligible. exceptions in Technical Specification 5.5.12 The amendments would revise
The integrity of the containment is subject are administrative in nature and do not create Technical Specification (TS) 5.5.15,
to two types of failure mechanisms that can the possibility of a new or different kind of
be categorized as: (1) Activity based, and (2)
‘‘Containment Leakage Rate Testing
accident from any previously evaluated.
time based. Activity based failure Program,’’ to require a program in
Therefore, the proposed change does not
mechanisms are defined as degradation due create the possibility of a new or different accordance with Nuclear Energy
to system and component modifications or kind of accident from any previously Institute (NEI) topical report NEI 94–01,
maintenance. Local leak rate test evaluated. Revision 3–A, ‘‘Industry Guideline for
requirements and administrative controls 3. Does the proposed change involve a Implementing Performance-Based
such as configuration management and significant reduction in a margin of safety? Option of 10 CFR part 50, Appendix J.’’
procedural requirements for system Response: No. This proposed change will allow
restoration ensure that containment integrity The proposed license amendment does not extension of the Type A test interval up
is not degraded by plant modifications or alter the way safety limits, limiting safety
maintenance activities. The design and to one test in 15 years and extension of
system set points, or limiting conditions for
construction requirements of the the Type C test interval up to 75
operation are determined. The specific
containment combined with the containment requirements and conditions of the Technical months, based on acceptable
inspections performed in accordance with Specification Primary Containment Leakage performance history as defined in NEI
[American Society for Mechanical Engineers Rate Testing Program exist to ensure that the 94–01, Revision 3–A.
Boiler and Pressure Vessel Code (ASME degree of containment structural integrity Basis for proposed no significant
Code)], Section XI, and Technical and leak-tightness that is considered in the hazards consideration determination:
Specification requirements serve to provide a plant safety analysis is maintained. The As required by 10 CFR 50.91(a), the
high degree of assurance that the overall containment leak rate limit specified licensee has provided its analysis of the
containment would not degrade in a manner by Technical Specifications is maintained.
that is detectable only by a Type A test.
issue of no significant hazards
The design, operation, testing methods and consideration, which is presented
Based on the above, the proposed test acceptance criteria for Type A, B, and C
interval extensions do not significantly containment leakage tests specified in
below:
increase the consequences of an accident applicable codes and standards would 1. Does the proposed change involve a
previously evaluated. continue to be met, with the acceptance of significant increase in the probability
The proposed amendment also deletes two this proposed amendment, since they are not consequences of an accident previously
previously granted exceptions to Primary affected by implementation of a performance- evaluated?
Containment Leakage Rate Testing Program based containment testing program. This Response: No.
guidelines. The exception regarding the ensures that the margin of safety in the plant The proposed amendment adopts the NRC-
performance of a Type A test no later than safety analysis is maintained. accepted guidelines of NEI 94–01, Revision
a specified date would be deleted as this The proposed amendment also deletes two 3–A, ‘‘Industry Guideline for Implementing
Type A test has already been performed. previously granted exceptions. The exception Performance-Based Option of 10 CFR part 50,
Additionally, the exception to use the regarding the performance of a Type A test Appendix J,’’ for development of the PBNP
corrections to NEI 94–01, Revision 0, would no later than a specified date would be performance-based containment testing
be deleted as those corrections would no deleted as this Type A test has already been program. NEI 94–01 allows, based on risk
longer be in use. These changes to the performed. Additionally, the exception to use and performance, an extension of Type A and
exceptions in Technical Specification 5.5.12 the corrections to NEI 94–01, Revision 0, Type C containment leak test intervals.
are administrative in nature and do not affect would be deleted as those corrections would Implementation of these guidelines continues
the probability or consequences of an no longer be in use. These changes to the to provide adequate assurance that during
accident previously evaluated. exceptions in Technical Specification 5.5.12 design basis accidents, the primary
Therefore, the proposed changes do not are administrative in nature and do not containment and its components will limit
result in a significant increase in the involve a significant reduction in a margin of leakage rates to less than the values assumed
probability or consequences of an accident safety. in the plant safety analyses.
previously evaluated. Therefore, the proposed changes do not The findings of the PBNP risk assessment
2. Does the proposed change create the involve a significant reduction in a margin of confirm the general findings of previous
possibility of a new or different kind of safety. studies that the risk impact with extending
accident from any accident previously the containment leak rate is small. Per the
evaluated? The NRC staff has reviewed the guidance provided in Regulatory Guide
Response: No. licensee’s analysis and, based on this 1.174, an extension of the leak test interval
Containment Type A and Type C testing review, it appears that the three in accordance with NEI 94–01, Revision 3–
requirements periodically demonstrate the standards of 10 CFR 50.92(c) are A results in an estimated change within, the
integrity of the containment and exist to satisfied. Therefore, the NRC staff very small change region.
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ensure the plant’s ability to mitigate the proposes to determine that the Since the change is implementing a
consequences of an accident. These tests do performance-based containment testing
amendment request involves no
not involve any accident precursors or program, the proposed amendment does not
initiators. significant hazards consideration. involve either a physical change to the plant
The proposed change does not involve a Attorney for licensee: David W. or a change in the manner in which the plant
physical modification to the plant (that is, no Jenkins, Attorney, FirstEnergy is operated or controlled. The requirement
new or different type of equipment will be Corporation, Mail Stop A–GO–15, 76 for containment leakage rate acceptance will
installed) nor does it alter the design, South Main Street, Akron, OH 44308. not be changed by this amendment.
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Therefore, the containment will continue to satisfied. Therefore, the NRC staff associated with implementation of the
perform its design function as a barrier to proposes to determine that the proposed Allowed Outage Time extension is
fission product releases. amendment request involves no consistent with the NRC’s Safety Goal Policy
Therefore, the proposed change does not Statement, as further described in RG
significant hazards consideration.
involve a significant increase in the [Regulatory Guide] 1.174 and RG 1.177. In
Attorney for licensee: William Blair, addition, a deterministic evaluation
probability or consequences of an accident
previously evaluated. Managing Attorney—Nuclear, Florida concluded that plant defense-in-depth
2. Does the proposed change create the Power & Light Company, P.O. Box philosophy will be maintained with the
possibility of a new or different kind of 14000, 700 Universe Boulevard, Juno proposed Allowed Outage Time extension.
accident from any accident previously Beach, FL 33408–0420. There will be no impact on the source term
evaluated? NRC Branch Chief: David J. Wrona. or pathways assumed in accidents previously
Response: No. evaluated. No analysis assumptions will be
The proposed change to implement a PSEG Nuclear LLC, Docket No. 50–354, changed and there will be no adverse effects
performance-based containment testing Hope Creek Generating Station (HCGS), on onsite or offsite doses as the result of an
program, associated with integrated leakage Salem County, New Jersey accident.
rate test frequency, does not change the Therefore, the proposed change does not
Date of amendment request: April 13,
design or operation of structures, systems, or involve a significant increase in the
2018. A publicly-available version is in probability or consequences of an accident
components of the plant.
The proposed change would continue to
ADAMS under Accession No. previously evaluated.
ensure containment integrity and would ML18103A218. 2. Does the proposed amendment create
ensure operation within the bounds of Description of amendment request: the possibility of a new or different kind of
existing accident analyses. There are no The amendment would revise Technical accident from any accident previously
accident initiators created or affected by this Specification (TS) 3.8.3.1, evaluated?
change. Therefore, the proposed change will ‘‘Distribution—Operating,’’ to increase Response: No.
not create the possibility of a new or different the alternating current (AC) inverters The proposed amendment does not involve
kind of accident from any accident physical alteration of the HCGS. No new
allowed outage time (AOT) from 24
previously evaluated. equipment is being introduced, and installed
hours to 7 days. The proposed change equipment is not being operated in a new or
Therefore, the proposed change does not is based on application of the HCGS
create the possibility of a new or different different manner. There is no change being
kind of accident from any accident
probabilistic risk assessment (PRA) in made to the parameters with in which the
previously evaluated. support of a risk-informed extension, HCGS is operated. There are no setpoints at
3. Does the proposed change involve a and on additional considerations and which protective or mitigating actions are
significant reduction in a margin of safety? compensatory actions. initiated that are affected by this proposed
Response: No. Basis for proposed no significant action. The use of the alternate Class 1E
Margin of safety is related to confidence in hazards consideration determination: power source for the AC distribution panel
the ability of the fission product barriers (fuel is consistent with the HCGS plant design.
As required by 10 CFR 50.91(a), the The change does not alter assumptions made
cladding, reactor coolant system, and licensee has provided its analysis of the
primary containment) to perform their design in the safety analysis. This proposed action
issue of no significant hazards will not alter the manner in which
functions during and following postulated
consideration, which is presented equipment operation is initiated, nor will the
accidents. The proposed change to
implement a performance-based containment below: functional demands on credited equipment
be changed. No alteration is proposed to the
testing program, associated with integrated 1. Does the proposed amendment involve
procedures that ensure the HCGS remains
leakage rate test and local leak rate testing a significant increase in the probability or
with in analyzed limits, and no change is
frequency, does not affect plant operations, consequences of an accident previously
evaluated? being made to procedures relied upon to
design functions, or any analysis that verifies
Response: No. respond to an off-normal event. As such, no
the capability of a structure, system, or
The proposed TS amendment does not new failure modes are being introduced.
component of the plant to perform a design
affect the design of the AC inverters, the Therefore, the proposed changes do not
function. In addition, this change does not
operational characteristics or function of the create the possibility of a new or different
affect safety limits, limiting safety system
inverters, the interfaces between the inverters kind of accident from any previously
setpoints, or limiting conditions for
and other plant systems, or the reliability of evaluated.
operation.
the inverters. An inoperable AC inverter is 3. Does the proposed amendment involve
The specific requirements and conditions
not considered an initiator of an analyzed a significant reduction in a margin of safety?
of the TS Containment Leakage Rate Testing
event. In addition, TS Actions and the Response: No.
Program exist to ensure that the degree of
associated Allowed Outage Times are not Margin of safety is related to the
containment structural integrity and leak-
initiators of previously evaluated accidents. confidence in the ability of the fission
tightness that is considered in the plant
Extending the Allowed Outage Time for an product barriers to perform their design
safety analysis is maintained. The overall
inoperable AC inverter would not have a functions during and following an accident.
containment leak rate limit specified by TS
significant impact on the frequency of These barriers include the fuel cladding, the
is maintained. This ensures that the margin
occurrence of an accident previously reactor coolant system, and the containment
of safety in the plant safety analysis is
evaluated. The proposed amendment will not system. The proposed change, which would
maintained. The design, operation, testing
result in modifications to plant activities increase the AOT from 24 hours to 7 days for
methods and acceptance criteria for Type A,
associated with inverter maintenance, but one inoperable inverter, does not exceed or
B, and C containment leakage tests specified
rather, provides operational flexibility by alter a setpoint, design basis or safety limit.
in applicable codes and standards would
allowing additional time to perform inverter Therefore, the proposed amendment does
continue to be met with the acceptance of
troubleshooting, corrective maintenance, and not involve a significant reduction in a
this proposed change since these are not
post-maintenance testing on-line. margin of safety.
affected by implementation of a performance-
based containment testing program. The proposed extension of the Completion The NRC staff has reviewed the
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Therefore, the proposed change does not Time for an inoperable AC inverter will not licensee’s analysis and, based on this
involve a significant reduction in a margin of significantly affect the capability of the
review, it appears that the three
safety. inverters to perform their safety function,
which is to ensure an uninterruptible supply standards of 10 CFR 50.92(c) are
The NRC staff has reviewed the of 120-volt AC electrical power to the satisfied. Therefore, the NRC staff
licensee’s analysis and, based on this associated power distribution subsystems. proposes to determine that the
review, it appears that the three An evaluation, using PRA methods, amendment request involves no
standards of 10 CFR 50.92(c) are confirmed that the increase in plant risk significant hazards consideration.
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Attorney for licensee: Jeffrie J. Keenan, detecting a 0.5 gpm leak in containment, the standards of 10 CFR 50.92(c) are
PSEG Nuclear LLC—N21, P.O. Box 236, change to the SSC has no effect on plant satisfied. Therefore, the NRC staff
Hancocks Bridge, NJ 08038. operations. There is no change to plant proposes to determine that the
NRC Branch Chief: James G. Danna. systems or the response of systems to
amendment request involves no
postulated accident conditions. There is no
Southern Nuclear Operating Company, change to the predicted radioactive releases significant hazard consideration
Docket Nos. 52–025 and 52–026, Vogtle due to normal operation or postulated Attorney for licensee: M. Stanford
Electric Generating Plant, Unit Nos. 3 accident conditions. Blanton, Balch & Bingham LLP, 1710
and 4, Burke County, Georgia Therefore, the proposed amendment does Sixth Avenue, North, Birmingham, AL
not involve a significant increase in the 35203–2015.
Date of amendment request: April 26, probability or consequences of an accident NRC Branch Chief: Jennifer L. Dixon-
2018. A publicly-available version is in previously evaluated. Herrity.
ADAMS under Accession No. 2. Does the proposed amendment create
ML18116A138. the possibility of a new or different kind of Southern Nuclear Operating Company,
Description of amendment request: accident from any accident previously Docket Nos. 52–025 and 52–026, Vogtle
The requested amendment proposes evaluated? Electric Generating Plant, Unit Nos. 3
changes to combined license (COL) Response: No. and 4, Burke County, Georgia
The proposed change does not affect the
Appendix C, with corresponding operation of any systems or equipment that Date of amendment request: April 27,
changes to the associated plant-specific may initiate a new or different kind of 2018. A publicly-available version is in
Tier 1 information, and involves accident, or alter any SSC such that a new ADAMS under Accession No.
associated Tier 2 information in the accident initiator or initiating sequence of ML18117A464.
Updated Final Safety Analysis Report events is created. The proposed change to the Description of amendment request:
(UFSAR) (which includes the plant- containment sump water level The requested amendment proposes to
specific Design Control Document instrumentation and its expected RCS
depart from Tier 2 information in the
(DCD) Tier 2 information). Pursuant to leakage detection capability has no adverse
effect on the ability to detect a 0.5 gpm leak Updated Final Safety Analysis Report
the provisions of 10 CFR 52.63(b)(1), (UFSAR) (which includes the plant-
in containment. The containment sump level
also requested is an exemption from instrumentation functions are unchanged and specific Design Control Document Tier
elements of the design as certified in the leak-before-break design criteria are not 2 information) and involves related
10 CFR part 52, appendix D, design adversely affected. changes to plant-specific Tier 1
certification rule for the plant-specific Loss of coolant accidents for a spectrum of information, with corresponding
DCD departures. pipe sizes and locations are already changes to the associated combined
The requested amendment proposes postulated in UFSAR Chapter 15, Section license (COL) Appendix C information.
changes to COL Appendix C (and plant- 15.6. Breaks in the main steam lines inside
Specifically, the amendment, if
specific Tier 1) to reflect a new design containment are also analyzed in UFSAR
Chapter 15, Section 15.1. Unidentified approved, would revise the Tier 2
of containment sump level sensors that information in the UFSAR and related
affects the acceptance criterion for the leakage detection and operator action in
response to unidentified leakage are not changes to Tier 1 and the associated
detected containment sump level postulated for any of the design basis COL Appendix C to remove the fire
change test and the associated minimum accident analyses described in UFSAR protection system non-safety related
detectable unidentified leakage rate in Chapter 15. containment cable spray and install
plant-specific DCD Tier 2 information. Therefore, the requested amendment does passive fire stops and radiant energy
Basis for proposed no significant not create the possibility of a new or different shields. The changes to Tier 1 require an
hazards consideration determination: kind of accident from any accident
exemption, which is included in the
As required by 10 CFR 50.91(a), the previously evaluated.
3. Does the proposed amendment involve license amendment request.
licensee has provided its analysis of the Basis for proposed no significant
issue of no significant hazards a significant reduction in a margin of safety?
Response: No. hazards consideration determination:
consideration, which is presented As required by 10 CFR 50.91(a), the
The described change to the containment
below: sump water level instrumentation and its licensee has provided its analysis of the
1. Does the proposed amendment involve expected RCS leakage detection capability is issue of no significant hazards
a significant increase in the probability or proposed to verify that the ability to detect consideration, which is presented
consequences of an accident previously a 0.5 gpm leak in containment is maintained. below:
evaluated? The proposed change does not alter any
Response: No. safety-related equipment, applicable design 1. Does the proposed amendment involve
The proposed change is to the containment codes, code compliance, design function, or a significant increase in the probability or
sump water level instrumentation and its safety analysis. By ensuring that the chosen consequences of an accident previously
expected [reactor coolant system (RCS)] equipment can detect a 0.5 gpm leak in evaluated?
leakage detection capability. The affected containment with the described accuracy, Response: No.
equipment is not safety-related, but the guidance in Regulatory Guide 1.45, Revision The proposed changes do not affect the
containment sump water level sensors are 0, as committed to in the UFSAR, and operation or reliability of any system,
seismically qualified. The change in requirements in the Technical Specifications structure or component (SSC) required to
containment sump level monitoring are met which ensures that leak-before-break maintain a normal power operating condition
instruments has no adverse effect on the design criteria are not adversely affected. or to mitigate anticipated transients without
ability to detect a 0.5 [gallons per minute Consequently, no safety analysis or design safety-related systems. Testing has
(gpm)] leak in containment, and therefore, basis acceptance limit/criterion is challenged demonstrated that the passive fire stops
has no adverse effect on design criteria for or exceeded by the proposed change, thus the prevent propagation of fires along the length
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leak-before-break. The change does not affect margin of safety is not reduced. of cable trays and prevent the propagation of
the operation of any systems or equipment Therefore, the proposed amendment does cable tray fires to adjacent fire zones. The
that initiate an analyzed accident or alter any not involve a significant reduction in a proposed changes do not affect the operation
structures, systems, and components (SSC) margin of safety. of equipment whose failure could initiate an
accident initiator or initiating sequence of accident previously analyzed. The existence
events. The NRC staff has reviewed the or failure of passive fire stops in fire zone
Because the containment sump water level licensee’s analysis and based on this 1100 AF 11300B does not affect normal
monitoring channels are still capable of review it appears that the three equipment operation.
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satisfy the same design functions in limits is not an initiator for any accident operating condition. Therefore, the changes
accordance with the same codes and previously evaluated. In addition, the current do not involve a significant reduction in a
standards as stated in the UFSAR. These variable limit on primary coolant iodine margin of safety.
changes do not affect any design code, concentration is not an initiator to any
The NRC staff has reviewed the
function, design analysis, safety analysis accident previously evaluated. Updating the
input or result, or design/safety margin. Alternative Source Term analyses does not licensee’s analysis and, based on this
No safety analysis or design basis require any changes to any plant structures, review, it appears that the three
acceptance limit/criterion is challenged or systems, or components (SSCs) and therefore standards of 10 CFR 50.92(c) are
exceeded by the proposed changes, and no does not affect any accident initiators. As a satisfied. Therefore, the NRC staff
margin of safety is reduced. result, the proposed changes do not proposes to determine that the
Therefore, the requested amendment does significantly increase the probability of an amendment request involves no
not involve a significant reduction in a accident. The proposed TS change will limit significant hazards consideration.
margin of safety. primary coolant noble gases to Attorney for licensee: Lillian M.
concentrations consistent with the accident Cuoco, Senior Counsel, Dominion
The NRC staff has reviewed the analyses, and the proposed completion time
licensee’s analysis and, based on this when the limit may be exceeded has no
Resources Services, Inc., 120 Tredegar
review, it appears that the three impact on the consequences of any design St., RS–2, Richmond, VA 23219.
standards of 10 CFR 50.92(c) are basis accident since the consequences of an NRC Branch Chief: Michael T.
satisfied. Therefore, the NRC staff accident during this time period is the same Markley.
proposes to determine that the as the consequences of an accident during III. Notice of Issuance of Amendments
the existing time periods. The revised
amendment request involves no to Facility Operating Licenses and
assessments of the radiological consequences
significant hazards consideration. due to design basis accidents listed in the Combined Licenses
Attorney for licensee: M. Stanford Surry Updated Final Safety Analysis Report, During the period since publication of
Blanton, Balch & Bingham LLP, 1710 using the updated AST methodology and the last biweekly notice, the
Sixth Avenue North, Birmingham, AL proposed assumptions and inputs, conclude Commission has issued the following
35203–2015. that the Exclusion Area Boundary (EAB),
amendments. The Commission has
NRC Branch Chief: Jennifer L. Dixon- Low Population Zone (LPZ), and Control
Room doses are within the limits of 10 CFR determined for each of these
Herrity. amendments that the application
50.67 and within the limits of Regulatory
Virginia Electric and Power Company, Guide (RG) 1.183. As a result, the complies with the standards and
Docket Nos. 50–280 and 50–281, Surry consequences of any accident previously requirements of the Atomic Energy Act
Power Station, Unit Nos. 1 and 2 evaluated are not significantly increased. of 1954, as amended (the Act), and the
(Surry), Surry County, Virginia Criterion 2. The Proposed Changes Do Not Commission’s rules and regulations.
Create the Possibility of a New or Different The Commission has made appropriate
Date of amendment request: March 2, Kind of Accident From Any Accident findings as required by the Act and the
2018. A publicly-available version is in Previously Evaluated Commission’s rules and regulations in
ADAMS under Accession No. The proposed TS change in specific 10 CFR chapter I, which are set forth in
ML18075A021. activity limits and the updated AST dose the license amendment.
Description of amendment request: consequences analyses do not alter any A notice of consideration of issuance
The amendments would revise the physical part of the plant, (i.e., no new or of amendment to facility operating
Technical Specifications (TSs) different type of equipment will be installed,) license or combined license, as
consistent with Revision 0 to the nor do they affect any plant operating
parameter or create new accident precursors.
applicable, proposed no significant
Technical Specifications Task Force hazards consideration determination,
(TSTF) Standard Technical Therefore, the proposed changes do not
create the potential for a new or different and opportunity for a hearing in
Specification Change Document TSTF– connection with these actions, was
kind of accident from any previously
490, ‘‘Deletion of E Bar Definition and calculated. published in the Federal Register as
Revision to RCS Specific Activity Tech indicated.
Criterion 3. The Proposed Change Does Not
Spec.’’ The proposed amendments Involve a Significant Reduction in the Margin Unless otherwise indicated, the
would adopt TSTF–490 and make the of Safety. Commission has determined that these
following associated changes: (1) amendments satisfy the criteria for
The proposed TS change in specific
Adoption of a TS change to replace the activity limits is consistent with the categorical exclusion in accordance
current limits on primary coolant gross assumptions in -the safety analyses and will with 10 CFR 51.22. Therefore, pursuant
specific activity with limits on primary ensure the monitored values protect the to 10 CFR 51.22(b), no environmental
coolant noble gas activity, and (2) an initial assumptions in the safety analyses. impact statement or environmental
update of the Alternative Source Term The proposed changes for radiological events assessment need be prepared for these
(AST) analyses for Surry. related to the computer code used to amendments. If the Commission has
Basis for proposed no significant calculate dose, revised X/Qs for control room
and offsite receptors (including the computer
prepared an environmental assessment
hazards consideration determination: under the special circumstances
code and method used to determine control
As required by 10 CFR 50.91(a), the provision in 10 CFR 51.22(b) and has
room X/Qs for SG releases), the computer
licensee has provided its analysis of the code used to determine core inventory, the made a determination based on that
issue of no significant hazards change in FHA [Fuel Handling Accident] gap assessment, it is so indicated.
consideration, which is presented fraction methodology, and removing the LRA For further details with respect to the
below: [Locked Rotor Accident] from the action see (1) the applications for
radiological design basis have been analyzed amendment, (2) the amendment, and (3)
daltland on DSKBBV9HB2PROD with NOTICES
Criterion 1. The Proposed Changes Do Not and result in acceptable consequences,
Involve a Significant Increase in the the Commission’s related letter, Safety
meeting the criteria as specified in 10 CFR
Probability or Consequences of an Accident Evaluation, and/or Environmental
50.67 and RG 1.183. The proposed changes
Previously Evaluated will not result in plant operation in a Assessment as indicated. All of these
Reactor coolant specific activity is not an configuration outside the analyses or design items can be accessed as described in
initiator for any accident previously basis and do not adversely affect systems that the ‘‘Obtaining Information and
evaluated, and the allowed time period when are required to respond for safe shutdown of Submitting Comments’’ section of this
primary coolant gross activity is not within the plant and to maintain the plant in a safe document.
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![Federal Register / Vol. 83, No. 118 / Tuesday, June 19, 2018 / Notices 28467
Southern Nuclear Operating Company, Tennessee Valley Authority (TVA) consideration determination as
Inc.; Georgia Power Company; Docket Nos. 50–259, 50–260, 50–296, published in the Federal Register.
Oglethorpe Power Corporation; and 72–052, Browns Ferry Nuclear The Commission’s related evaluation
Municipal Electric Authority of Georgia; Plant, Unit Nos. 1, 2, and 3, Limestone of the amendment is contained in a
and City of Dalton, Georgia, Docket Nos. County, Alabama Safety Evaluation dated December 22,
50–321 and 50–366, Edwin I. Hatch TVA Docket Nos. 50–327, 50–328, and 2017.
Nuclear Plant, Unit Nos. 1 and 2, 72–034, Sequoyah Nuclear Plant, Unit No significant hazards consideration
Appling County, Georgia Nos. 1 and 2, Hamilton County, comments received: No.
Tennessee
Date of amendment request: April 20, Union Electric Company, Docket No.
2017, as supplemented by letters dated TVA Docket Nos. 50–390, 50–391, and
72–1048, Watts Bar Nuclear Plant, Unit 50–483, Callaway Plant, Unit No. 1,
September 14, 2017; February 19, 2018; Callaway County, Missouri
and May 1, 2018. Nos. 1 and 2, Rhea County, Tennessee
Brief description of amendments: The Date of amendment request: January Date of amendment request: April 6,
amendments revised the Technical 4, 2017, as supplemented by letters 2017, as supplemented by letter dated
Specifications by replacing the existing dated July 7, 2017, and July 27, 2017. February 5, 2018.
requirements related to ‘‘operations (Note: This Notice of Issuance corrects Brief description of amendment: The
with a potential for draining the reactor the amendments by adding the amendment revised the Final Safety
vessel’’ with new requirements on supplement dated July 27, 2017, which Analysis Report to clearly describe
was inadvertently omitted from the conformance with NRC Regulatory
Reactor Pressure Vessel Water Inventory
original Federal Register notice (January Guide 1.106, Revision 1, ‘‘Thermal
Control to protect Safety Limit 2.1.1.3,
16, 2018; 83 FR 2234). Overload Protection for Electric Motors
which requires reactor vessel water
Brief description of amendments: The on Motor-Operated Valves.’’
level to be greater than the top of active
amendments revised TVA Emergency
irradiated fuel. Date of issuance: May 30, 2018.
Plans for the above nuclear plants.
Date of issuance: May 31, 2018. Specifically, the amendments adopted Effective date: As of the date of
Effective date: As of the date of the NRC-endorsed Radiological issuance and shall be implemented
issuance and shall be implemented Emergency Plan Emergency Action within 90 days from the date of
prior to the commencement of the Unit Level schemes developed by the issuance.
No. 2 refueling outage (U2R25) in Nuclear Energy Institute (NEI 99–01,
Amendment No.: 218. A publicly-
February 2019. Revision 6, ‘‘Development of Emergency
available version is in ADAMS under
Action Levels for Non-Passive
Amendment Nos.: Unit 1—290, Unit Reactors’’). Accession No. ML18124A026;
2—235. A publicly-available version is documents related to this amendment
Date of issuance: December 22, 2017.
in ADAMS under Accession No. are listed in the Safety Evaluation
Effective date: As of the date of enclosed with the amendment.
ML18123A368; documents related to issuance and shall be implemented
these amendments are listed in the within 180 days from the date of its Renewed Facility Operating License
Safety Evaluation enclosed with the issuance, or July 3, 2018, whichever No. NPF–30: The amendment revised
amendments. comes later. the Final Safety Analysis Report.
Renewed Facility Operating License Amendment Nos.: Browns Ferry Date of initial notice in Federal
Nos. DPR–57 and NPF–5: The Nuclear Plant—303 (Unit 1), 327 (Unit Register: July 18, 2017 (82 FR 32885).
amendments revised the Renewed 2), and 287 (Unit 3); Sequoyah Nuclear The supplemental letter dated February
Facility Operating Licenses and Plant—339 (Unit 1) and 332 (Unit 2); 5, 2018, provided additional
Technical Specifications. and Watts Bar Nuclear Plant—118 (Unit information that clarified the
1) and 18 (Unit 2). A publicly-available application, did not expand the scope of
Date of initial notice in Federal version is in ADAMS under Accession
Register: August 29, 2017 (82 FR the application as originally noticed,
No. ML17289A032; documents related and did not change the NRC staff’s
41071). The supplemental letters dated to these amendments are listed in the original proposed no significant hazards
September 14, 2017; February 19, 2018; Safety Evaluations enclosed with the consideration determination as
and May 1, 2018, provided additional amendments. These amendments were published in the Federal Register.
information that clarified the corrected by letter dated May 29, 2018
application, did not expand the scope of (ADAMS Accession No. ML18138A452). The Commission’s related evaluation
the application as originally noticed, Renewed Facility Operating License of the amendment is contained in a
and did not change the NRC staff’s Nos. DPR–33, DPR–52, DPR–68, DPR– Safety Evaluation dated May 30, 2018.
original proposed no significant hazards 77, and DPR–79, and Facility Operating No significant hazards consideration
consideration determination as License Nos, NPF–90 and NPF–96: The comments received: No.
published in the Federal Register. amendments revised the licenses. Dated at Rockville, Maryland, this 6th day
The Commission’s related evaluation Date of initial notice in Federal of June 2018.
of the amendments is contained in a Register: June 19, 2017 (82 FR 27891).
For the Nuclear Regulatory Commission.
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Safety Evaluation dated May 31, 2018. The supplemental letters dated July 7,
2017, and July 27, 2017, provided Tara Inverso,
No significant hazards consideration additional information that clarified the Acting Deputy Director, Division of Operating
comments received: No. application, did not expand the scope of Reactor Licensing, Office of Nuclear Reactor
the application as originally noticed, Regulation.
and did not change the NRC staff’s [FR Doc. 2018–12506 Filed 6–18–18; 8:45 am]
original proposed no significant hazards BILLING CODE 7590–01–P
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Document Created: 2018-07-02 11:21:50
Document Modified: 2018-07-02 11:21:50
| 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 July 19, 2018. A request for a hearing must be filed by August 20, 2018. | |
| Contact | Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1384, email: [email protected] | |
| FR Citation | 83 FR 28456 |
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