83 FR 15412 - 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 69 (April 10, 2018)
Federal Register Volume 83, Issue 69 (April 10, 2018)
| Page Range | 15412-15420 | |
| FR Document | 2018-06668 |
[Federal Register Volume 83, Number 69 (Tuesday, April 10, 2018)] [Notices] [Pages 15412-15420] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2018-06668] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2018-0064] 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 March 13, 2018, to March 26, 2018. The last biweekly notice was published on March 27, 2018. DATES: Comments must be filed by May 10, 2018. A request for a hearing must be filed by June 11, 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-2018-0064. Address questions about NRC dockets to Jennifer Borges; telephone: 301-287- 9127; email: [email protected]. For technical questions, contact the individual(s) 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: Kay Goldstein, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2018-0064, 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-2018-0064. 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-2018-0064, 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. [[Page 15413]] 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 Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's 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 [[Page 15414]] 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 [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 [[Page 15415]] proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP), Darlington County, South Carolina Date of amendment request: February 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18038B289. Description of amendment request: The amendment would revise Technical Specification (TS) Section 3.4.3, ``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' to reduce the applicability terms from 50 effective full power years (EFPY) to 46.3 EFPY in Figures 3.4.3-1 and 3.4.3-2, as a result of the removal of part length fuel assemblies (PLSAs) and the migration to 24-month fuel cycles. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. The proposed change does not involve physical changes to the plant or alter the reactor coolant system (RCS) pressure boundary (i.e., there are no changes in operating pressure, materials or seismic loading). The P/T limit curves and Adjusted Reference Temperature (ART) values will remain as-is. Only the term to which the limit curves applies is effected by the proposed change. The P/T limit curves in TS 3.4.3 with an applicability term of 46.3 EFPY provide continued assurance that the fracture toughness of the reactor pressure vessel (RPV) is consistent with analysis assumptions and NRC regulations. The methodology used to develop the existing P/T limit curves provides assurance that the probability of a rapidly propagating failure will be minimized. The P/T limit curves, with the applicability term reduced to a proposed 46.3 EFPY, will continue to prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. The proposed change does not affect the design or assumed accident performance of any structure, system or component, or introduce any new modes of system operation or failure modes. Compliance with the proposed P/T curves (same as the existing P/T curves with the applicability term reduced to 46.3 EFPY) will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises TS 3.4.3 to reflect that Figures 3.4.3-1 and 3.4.3-2 (P/T limit curves) are applicable up to 46.3 EFPY instead of 50 EFPY with the removal of PLSAs and migration to 24-month fuel cycles. HBRSEP adheres to applicable NRC regulations (i.e., 10 CFR 50, Appendices G and H) and NRC-approved methodologies (i.e., Regulatory Guides 1.99 and 1.190) with respect to the P/T limit curves in TS 3.4.3 in order to provide an adequate margin of safety to the conditions at which brittle fracture may occur. The P/ T limit curves, with the applicability term reduced to 46.3 EFPY, continue to provide assurance that the established P/T limits are not exceeded. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street, DEC45A, Charlotte NC 28202. NRC Acting Branch Chief: Brian W. Tindell. Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of amendment request: February 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18039A123. Description of amendment request: LSCS Technical Specifications (TS) 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' currently requires performance of Surveillance Requirement (SR) 3.6.1.3.8 on each excess flow check valve (EFCV) during each refueling outage. The proposed amendments would revise the number of EFCVs tested by TS SR 3.6.1.3.8 from ``each'' to a ``representative sample.'' The representative sample is based on approximately 20 percent of the reactor instrumentation line EFCVs such that each EFCV will be tested at least once every 10 years (nominal). Therefore, approximately 20 percent of the EFCVs will be tested every operating cycle. The reduced testing associated with the proposed change will result in an increase in the availability of the associated instrumentation during outages and will result in dose savings. 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 analyzed? Response: No. The EFCVs at LSCS, Unit 1 and Unit 2, are designed so that they will not close accidently during normal operations, will close if a rupture of the instrument line is indicated downstream of the valve, can be reopened when appropriate, and have their status indicated in the control room. This proposed change relaxes the number of EFCVs tested for TS SR 3.6.1.3.8 from ``each'' to a ``representative sample'' in accordance with the SFCP [Surveillance Frequency Control Program]. There are no physical plant modifications associated with this change. Industry and LSCS operating experience demonstrate a high reliability of these valves. Neither EFCVs nor their failures are capable of initiating previously evaluated accidents; therefore, there can be no increase in the probability of occurrence of an accident regarding this proposed change. The LSCS Updated Final Safety Analysis Report (UFSAR) demonstrates, consistent with BWROG [Boiling Water Reactor Owners Group] topical report NEDO-32977-A, that the failure of an EFCV has very low [[Page 15416]] consequences. The LSCS UFSAR evaluates a circumferential rupture of an instrument line that is connected to the primary coolant system. The evaluation credits the 0.25-inch diameter flow-restricting orifice installed in the line with limiting flow following the instrumentation line break and does not credit the EFCV with actuating to limit leakage. The dose consequences of the instrument line break are determined using the calculated mass of coolant released over approximately a five-hour period. The reactor was assumed to be operating at design power conditions prior to the break. The Standby Gas Treatment System (SGTS) and secondary containment are not impaired by the event. The evaluation concludes that the consequences of the event are well within 10 CFR 100 limits. Thus, the failure of an EFCV, though not expected as a result of the proposed change, does not affect the dose consequences of an instrument line break. Based on the above, it is concluded that the proposed change to the EFCV surveillance requirement does not involve a significant increase in the probability or consequences of an accident previously analyzed. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. This proposed change allows a reduced number of EFCVs to be tested in accordance with the SFCP [Surveillance Frequency Control Program]. The proposed change would revise SR 3.6.1.3.8 to verify that a ``representative sample'' (i.e., approximately 20 percent) of reactor instrumentation line EFCVs are tested, in accordance with the SFCP, such that each EFCV will be tested at least once every 10 years (nominal). No other changes in the requirements are being proposed. Industry and LSCS-specific operating experience demonstrates the high degree of reliability of the EFCVs and the low consequences of an EFCV failure. The potential failure of an EFCV to isolate by the proposed reduction in test frequency is bounded by the previous evaluation of an instrument line rupture. This change will not alter the operation or process variables, structures, systems, or components as described in the safety analysis. Thus, a new or different kind of accident will not be created from implementation of the proposed change. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes do not involve a significant reduction in the margin of safety. The LSCS UFSAR evaluates a circumferential rupture of an instrument line that is connected to the primary coolant system. The evaluation credits the 0.25-inch diameter flow- restricting orifice installed in the line with limiting flow following the instrumentation line break and does not credit the EFCV with actuating to limit leakage. The dose consequences of the instrument line break are determined using the calculated mass of coolant released over approximately a five-hour period. The reactor was assumed to be operating at design power conditions prior to the break. The SGTS [Standby Gas Treatment System] and secondary containment are not impaired by the event. The evaluation concludes that the consequences of the event are well within 10 CFR 100 limits. Thus, the failure of an EFCV, though not expected as a result of the proposed change, does not affect the dose consequences of an instrument line break. 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, (EGC) Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of amendment request: February 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18058A257. Description of amendment request: The proposed amendments would revise LSCS Technical Specifications (TS) 3.4, Reactor Coolant System (RCS), Section 3.4.4, ``Safety/Relief Valves (S/RVs).'' Specifically, EGC proposes a new safety function lift setpoint lower tolerance for the S/RVs as delineated in Surveillance Requirement 3.4.4.1. The proposed change will revise the lower setpoint tolerances from -3 percent (%) to -5%. This change is limited to the lower tolerances and does not affect the upper tolerances; therefore, the upper tolerance will remain at +3% of the safety function lift setpoint. In addition, this change only applies to the as-found tolerance and not to the as-left tolerance, which will remain unchanged at 1% of the safety lift setpoint. The as-found tolerances are used for determining operability and to increase sample sizes for S/RV testing should the tolerance be exceeded. There will be no revision to the actual setpoints of the valves installed in the plant due to this change. This proposed change will preclude the submittal of previously- reportable licensee event reports (LERs) to the NRC due to setpoint drift in the low (conservative) direction. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This change has no influence on the probability or consequences of any accident previously evaluated. The lower setpoint tolerance change does not affect the operation of the valves and it does not change the as-left setpoint tolerance. The change only affects the lower tolerance for valve opening and does not change the upper tolerance, which is the limit that protects from overpressurization. The proposed amendments do not involve physical changes to the valves, nor do they change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components. The proposed amendments do not change any other behavior or operation of any safety/relief valves (S/RVs), and, therefore, has no significant impact on reactor operation. They also have no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the Updated Final Safety Analysis Report (UFSAR). Based on the above, it is concluded that the proposed change to the S/RV surveillance requirement does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the S/RV safety lower setpoint tolerance from -3% to -5% only affects the criteria to determine when an as- found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance. The proposed lower setpoint tolerance change does not adversely affect the operation of any safety-related components or equipment. The proposed amendments do not involve physical changes to the S/ RVs, nor do they change the safety function of the S/RVs. The proposed amendments do not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed, and installed equipment is not being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure [[Page 15417]] the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed amendments involve a significant reduction in a margin of safety? Response: No. The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the S/RV setpoint upper setpoint tolerance. The TS setpoints for the S/RVs are not changed. The as-left setpoint tolerances are not changed by this proposed change and remain at 1% of the safety lift setpoint. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not significantly impact the condition or performance of structures, systems, and components relied upon for accident mitigation. Therefore, this 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50- 457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50- 455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249, Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: January 31, 2018. A publicly-available version is in ADAMS under Package Accession No. ML18053A159. Description of amendment request: The amendments would revise the emergency response organization (ERO) positions identified in the emergency plan for each site. 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to the [site] Emergency Plan do not increase the probability or consequences of an accident. The proposed changes do not impact the function of plant Structures, Systems, or Components (SSCs). The proposed changes do not affect accident initiators or accident precursors, nor do the changes alter design assumptions. The proposed changes do not alter or prevent the ability of the onsite ERO to perform their intended functions to mitigate the consequences of an accident or event. The proposed changes remove ERO positions no longer credited or considered necessary in support of Emergency Plan implementation. Therefore, the proposed changes to the [site] Emergency Plan do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes have no impact on the design, function, or operation of any plant SSCs. The proposed changes do not affect plant equipment or accident analyses. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed changes do not introduce failure modes that could result in a new accident, and the proposed changes do not alter assumptions made in the safety analysis. The proposed changes remove ERO positions no longer credited or considered necessary in support of Emergency Plan implementation. Therefore, the proposed changes to the [site] Emergency Plan do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. Margins of safety are unaffected by the proposed changes to the ERO staffing. The proposed changes are associated with the [site] Emergency Plan staffing and do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by these proposed changes. The proposed changes to the Emergency Plan will continue to provide the necessary onsite ERO response staff. Therefore, the proposed changes to the [site] Emergency Plan do 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. Florida Power & Light Company, Docket Nos. 50-250 and 251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of application for amendment: June 28, 2017, as supplemented by letter dated February 28, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML17180A447 and ML18075A023, respectively. Description of amendment request: The amendments would modify the Technical Specifications (TSs) by [[Page 15418]] relocating to licensee-controlled documents select acceptance criteria specified in TS surveillance requirements (SRs) credited for satisfying Inservice Testing (IST) Program and Inservice Inspection Program requirements; deleting the SRs for the American Society of Mechanical Engineers Code Class 1, 2, and 3 components; replacing references to the Surveillance Frequency Control Program (SFCP) with reference to the Turkey Point IST Program where appropriate; establishing a Reactor Coolant Pump (RCP) Flywheel Inspection Program; and related editorial changes. Additionally, the amendments would delete a redundant SR for Accumulator check valve testing and add a footnote to the SR for Pressure Isolation Valve (PIV) testing. The license amendment request was originally noticed in the Federal Register on August 29, 2017 (82 FR 41069). The notice is being reissued in its entirety to include the revised scope, description of the amendment request, and proposed no significant hazards consideration determination. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes provide assurance that inservice testing will be performed in the manner and within the timeframes established by 10 CFR 50.55(a). The deletion of SR 4.0.5 and the deletion of IST acceptance criteria from SR 4.5.2.c and SR 4.6.2.1.b neither affect the conduct nor the periodicity of the inservice testing. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative in nature and can neither initiate nor affect the outcome of any accident previously evaluated. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot affect the likelihood or the outcome of accident previously evaluated. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and related editorial changes are administrative changes and cannot affect the likelihood or the outcome of any accident previously evaluated. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not affect the likelihood or the outcome of any accident previously evaluated. Therefore, facility operation in accordance with the proposed changes would 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 deletion of IST acceptance criteria from the TS does not affect the manner in which any SSC [structure, system, or component] is maintained or operated and does not introduce new SSCs or new methods for maintaining existing plant SSCs. Inservice testing will continue in the manner and periodicity specified in the IST program such that no new or different kind of accident can result. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative changes and cannot introduce new or different kinds of accidents. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot be an initiator of a new or different kind of accident. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and the other editorial changes are administrative changes and cannot introduce new or different kinds of accidents. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not introduce new or different kinds of accidents. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings and do not adversely impact plant operating margins or the reliability of equipment credited in safety analyses. The proposed changes provides assurance that inservice inspection and inservice testing will be performed in the manner and within the timeframes established by 10 CFR 50.55(a). The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes with no impact on the margin of safety currently described the Updated Final Safety Analysis Report. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and the other editorial changes are administrative changes with no impact on nuclear safety. In addition, deletion of a redundant Accumulator check valve SR 4.5.1.1.d, and the addition of a footnote to TS SR 4.4.6.2.2.d to avoid PIV repetitive loop testing do not affect any safety analyses assumptions, safety limits, or limiting safety system settings. Therefore, operation of the facility in accordance with the proposed changes will not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Debbie Hendel, Managing Attorney--Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB, Juno Beach, FL 33408-0420. NRC Acting Branch Chief: Brian W. Tindell. Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant (BFN), Unit 1, Limestone County, Alabama Date of amendment request: March 16, 2018. A publicly-available version is in ADAMS under Accession No. ML18080A171. Description of amendment request: The amendment would revise License Condition 2.C(18)(a)3 for Unit 1 that requires the submittal of a revised BFN Unit 1 replacement steam dryer (RSD) analysis utilizing the BFN Unit 3 on-dryer strain gauge based end-to-end bias and uncertainties at extended power conditions ``at least 90 days prior to the start of the BFN Unit 1 EPU [extended power uprate] outage.'' Specifically, the amendment reduces the time from 90 days to 15 days before the BFN Unit 1 EPU outage for the submittal of the revised analysis of the RSD. 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 consequence of an accident previously evaluated? Response: No. The proposed license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the Browns Ferry Nuclear Plant (BFN) Unit 1 replacement steam dryer (RSD), performed using an NRC-approved methodology benchmarked on the BFN Unit 3 RSD, must be submitted to the NRC for [[Page 15419]] information. There is no required review or approval of the revised analysis needed to satisfy the license condition. The proposed change is an administrative change to the period before the outage and does not impact any system, structure or component in such a way as to affect the probability or consequences of an accident previously evaluated. The proposed amendment is purely administrative and has no technical or safety aspects. 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 license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the BFN Unit 1 RSD must be submitted to the NRC for information. The proposed amendment is purely administrative and has no technical or safety aspects. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed license amendment reduces the length of time, from 90 days to 15 days, prior to the outage by which a revised analysis of the BFN Unit 1 RSD must be submitted to the NRC for information. The proposed amendment is purely administrative and has no technical or safety aspects. 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Acting Branch Chief: Brian W. Tindell. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment, as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: July 31, 2017. Brief description of amendment: The amendment revised the license to authorize the description of the emergency response organization requalification training frequency defined in the Emergency Plan to be changed from ``annually'' to ``once per calendar year not to exceed 18 months between training sessions.'' Date of issuance: March 26, 2018. Effective date: As of the date of its issuance and shall be implemented within 90 days. Amendment No.: 318. A publicly-available version is in ADAMS under Accession No. ML17289A175; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-59: The amendment revised the Renewed Facility Operating License and Emergency Plan. Date of initial notice in Federal Register: September 26, 2017 (82 FR 44854). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2018. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Date of amendment request: May 1, 2017, as supplemented by letters dated November 15 and December 20, 2017. Brief description of amendment: The amendment replaced existing technical specification 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. Safety Limit 2.1.1.3 requires reactor pressure vessel water level to be greater than the top of active irradiated fuel. The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF- 542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control.'' Date of issuance: March 22, 2018. Effective date: As of the date of issuance and shall be implemented prior to entering Mode 4 during the next refueling outage, C1R18, currently planned for April 2018. Amendment No.: 216. A publicly-available version is in ADAMS under Accession No. ML18043A505; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-62: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 5, 2017 (82 FR 31096). The supplement letters dated November 15, 2017, and December 20, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 22, 2018. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of amendment request: November 3, 2017. Brief description of amendment: The amendment changed the safety limit [[Page 15420]] minimum critical power ratio numeric values for Operating Cycle 17. Specifically, the amendment increased the numeric values of the safety limit minimum critical power ratio for Nine Mile Point Nuclear Station, Unit 2, from >=1.15 to >=1.17 for two recirculation loop operation, and from >=1.15 to >=1.17 for single recirculation loop operation. Date of issuance: March 16, 2018. Effective date: As of the date of issuance and shall be implemented prior to startup from the next refueling outage. Amendment No.: 167. A publicly-available version is in ADAMS under Accession No. ML18060A016; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 6, 2018 (83 FR 5280). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2018. No significant hazards consideration comments received: No. Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: January 31, 2018. Brief description of amendments: The amendments revised the Emergency Plan for St. Lucie to adopt the fire-related notification of unusual event requirement of the Nuclear Energy Institute 99-01, Revision 6, Emergency Action Level scheme. Date of issuance: March 26, 2018. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment Nos.: 244 and 195. A publicly-available version is in ADAMS under Accession No. ML18046A712; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-67 and NPF-16: The amendments revised the Renewed Facility Operating Licenses. Date of initial notice in Federal Register: February 14, 2018 (83 FR 6621). This notice provided an opportunity to request a hearing by April 15, 2018, but indicated that if the Commission makes a final no significant hazards consideration determination, any such hearing would take place after issuance of the amendments. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 2018. No significant hazards consideration comments received: No. Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: April 9, 2017, as supplemented by letter dated October 4, 2017. Brief description of amendments: The amendments revised the Technical Specifications (TSs) to remove various reporting requirements. Specifically, the amendments removed the requirements to prepare the Startup Report, the Annual Report, and various special reports. In addition, the amendments revised the TSs to remove the completion time for restoring spent fuel pool water level, to address inoperability of one of the two parallel flow paths in the residual heal removal or safely injection headers for the Emergency Core Cooling Systems, and to make other administrative changes, including updating plant staff and responsibilities. Date of issuance: March 19, 2018. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 279 (Unit No. 3) and 274 (Unit No. 4). A publicly- available version is in ADAMS under Accession No. ML18019A078; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: June 19, 2017 (82 FR 27889). The supplemental letter dated October 4, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in an SE dated March 19, 2018. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: August 31, 2017. Description of amendment: The amendments authorized changes to the VEGP Units 3 and 4 Combined Operating License (COL) page 7 and COL Appendix A, Technical Specifications, to make necessary changes so that there will be adequate detection of reactor coolant system and main steam line leakage at all times and that the associated limits account for instrumentation sensitivities not accounted for in the current VEGP Technical Specification 3.4.9. Date of issuance: March 12, 2018. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 115 (Unit 3) and 114 (Unit 4). Publicly-available versions are in ADAMS Package Accession No. ML18036A782, which includes the Safety Evaluation that references documents related to these amendments. Facility Combined License Nos. NPF-91 and NPF-92: Amendments revised the Facility Combined Licenses. Date of initial notice in Federal Register: October 10, 2017 (82 FR 47032). The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated March 12, 2018. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 28th day of March 2018. For the Nuclear Regulatory Commission. Tara Inverso, Acting Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2018-06668 Filed 4-9-18; 8:45 am] BILLING CODE 7590-01-P
![15412 Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices
medical users of nuclear materials. For the Nuclear Regulatory Commission. For additional direction on obtaining
Licenses are issued for, among other David Cullison, information and submitting comments,
things, the possession, use, processing, NRC Clearance Officer, Office of the Chief see ‘‘Obtaining Information and
handling, and importing and exporting Information Officer. Submitting Comments’’ in the
of nuclear materials, and for the [FR Doc. 2018–07257 Filed 4–9–18; 8:45 am] SUPPLEMENTARY INFORMATION section of
operation of nuclear reactors. BILLING CODE 7590–01–P this document.
7. The estimated number of annual FOR FURTHER INFORMATION CONTACT: Kay
responses: 43,530 (11,739 for reporting Goldstein, U.S. Nuclear Regulatory
[1,677 NRC licensees and 10,062 NUCLEAR REGULATORY Commission, Washington DC 20555–
Agreement State licensees], 21,018 for COMMISSION 0001; telephone: 301–415–1506, email:
recordkeeping [3,003 NRC licensees and [NRC–2018–0064] kay.goldstein@nrc.gov.
18,015 Agreement State licensees], and SUPPLEMENTARY INFORMATION:
10,773 for third-party disclosures [1,539 Biweekly Notice; Applications and
NRC licensees and 9,234 Agreement Amendments to Facility Operating I. Obtaining Information and
State licensees]). Licenses and Combined Licenses Submitting Comments
8. The estimated number of annual Involving No Significant Hazards A. Obtaining Information
respondents: 21,018 (3,003 NRC Considerations
Please refer to Docket ID NRC–2018–
licensees and 18,015 Agreement State AGENCY: Nuclear Regulatory 0064, facility name, unit number(s),
licensees). Commission. plant docket number, application date,
9. The estimated number of hours ACTION: Biweekly notice. and subject when contacting the NRC
needed annually to comply with the about the availability of information for
information collection requirement or SUMMARY: Pursuant to Section 189a.(2) this action. You may obtain publicly-
request: 640,776 hours (91,545 hours for of the Atomic Energy Act of 1954, as available information related to this
NRC licensees and 549,231 hours for amended (the Act), the U.S. Nuclear action by any of the following methods:
Agreement State licensees). Regulatory Commission (NRC) is • Federal Rulemaking Website: Go to
10. Abstract: 10 CFR part 20 publishing this regular biweekly notice. http://www.regulations.gov and search
establishes standards for protection The Act requires the Commission to for Docket ID NRC–2018–0064.
against ionizing radiation resulting from publish notice of any amendments • NRC’s Agencywide Documents
activities conducted under licenses issued, or proposed to be issued, and Access and Management System
issued by the NRC and by Agreement grants the Commission the authority to (ADAMS): You may obtain publicly-
States. These standards require the issue and make immediately effective available documents online in the
establishment of radiation protection any amendment to an operating license ADAMS Public Documents collection at
programs, maintenance of radiation or combined license, as applicable, http://www.nrc.gov/reading-rm/
protection programs, maintenance of upon a determination by the adams.html. To begin the search, select
radiation records recording of radiation Commission that such amendment ‘‘ADAMS Public Documents’’ and then
received by workers, reporting of involves no significant hazards select ‘‘Begin Web-based ADAMS
incidents which could cause exposure consideration, notwithstanding the Search.’’ For problems with ADAMS,
to radiation, submittal of an annual pendency before the Commission of a please contact the NRC’s Public
report to NRC and to Agreement States request for a hearing from any person. Document Room (PDR) reference staff at
of the results of individual monitoring, This biweekly notice includes all 1–800–397–4209, 301–415–4737, or by
and submittal of license termination notices of amendments issued, or email to pdr.resource@nrc.gov. The
information. These mandatory proposed to be issued, from March 13, ADAMS accession number for each
requirements are needed to protect 2018, to March 26, 2018. The last document referenced (if it is available in
occupationally exposed individuals biweekly notice was published on ADAMS) is provided the first time that
from undue risks of excessive exposure March 27, 2018. it is mentioned in this document.
to ionizing radiation and to protect the DATES: Comments must be filed by May • NRC’s PDR: You may examine and
health and safety of the public. 10, 2018. A request for a hearing must purchase copies of public documents at
be filed by June 11, 2018. the NRC’s PDR, Room O1–F21, One
III. Specific Requests for Comments ADDRESSES: You may submit comments White Flint North, 11555 Rockville
The NRC is seeking comments that by any of the following methods (unless Pike, Rockville, Maryland 20852.
address the following questions: this document describes a different
method for submitting comments on a B. Submitting Comments
1. Is the proposed collection of
information necessary for the NRC to specific subject): Please include Docket ID NRC–2018–
properly perform its functions? Does the • Federal Rulemaking Website: Go to 0064, facility name, unit number(s),
information have practical utility? http://www.regulations.gov and search plant docket number, application date,
for Docket ID NRC–2018–0064. Address and subject in your comment
2. Is the estimate of the burden of the
questions about NRC dockets to Jennifer submission.
information collection accurate? The NRC cautions you not to include
Borges; telephone: 301–287–9127;
3. Is there a way to enhance the email: Jennifer.Borges@nrc.gov. For identifying or contact information that
quality, utility, and clarity of the technical questions, contact the you do not want to be publicly
information to be collected? individual(s) listed in the FOR FURTHER disclosed in your comment submission.
daltland on DSKBBV9HB2PROD with NOTICES
4. How can the burden of the INFORMATION CONTACT section of this The NRC will post all comment
information collection on respondents document. submissions at http://
be minimized, including the use of • Mail Comments to: May Ma, Office www.regulations.gov, as well as enter
automated collection techniques or of Administration, Mail Stop: TWFN–7– the comment submissions into ADAMS.
other forms of information technology? A60M, U.S. Nuclear Regulatory The NRC does not routinely edit
Dated at Rockville, Maryland, this 5th day Commission, Washington, DC 20555– comment submissions to remove
of April 2018. 0001. identifying or contact information.
VerDate Sep<11>2014 16:56 Apr 09, 2018 Jkt 244001 PO 00000 Frm 00059 Fmt 4703 Sfmt 4703 E:\FR\FM\10APN1.SGM 10APN1](https://thefederalregister.org/doc/83_FR_15481/page/1.svg)
![Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices 15415
proximity to a facility or site. With P/T limit curves provides assurance that the NRC Acting Branch Chief: Brian W.
respect to copyrighted works, except for probability of a rapidly propagating failure Tindell.
limited excerpts that serve the purpose will be minimized. The P/T limit curves,
with the applicability term reduced to a Exelon Generation Company, LLC,
of the adjudicatory filings and would
proposed 46.3 EFPY, will continue to Docket Nos. 50–373 and 50–374, LaSalle
constitute a Fair Use application, prohibit operation in regions where it is County Station (LSCS), Units 1 and 2,
participants are requested not to include possible for brittle fracture of reactor vessel LaSalle County, Illinois
copyrighted materials in their materials to occur, thereby assuring that the
submission. integrity of the RCS pressure boundary is Date of amendment request: February
For further details with respect to maintained. 7, 2018. A publicly-available version is
these license amendment applications, Therefore, the proposed change does not in ADAMS under Accession No.
see the application for amendment involve a significant increase in the ML18039A123.
which is available for public inspection probability or consequences of an accident Description of amendment request:
in ADAMS and at the NRC’s PDR. For previously evaluated. LSCS Technical Specifications (TS)
2. Does the proposed change create the 3.6.1.3, ‘‘Primary Containment Isolation
additional direction on accessing
possibility of a new or different kind of
information related to this document, Valves (PCIVs),’’ currently requires
accident from any accident previously
see the ‘‘Obtaining Information and evaluated? performance of Surveillance
Submitting Comments’’ section of this Response: No. Requirement (SR) 3.6.1.3.8 on each
document. The proposed change revises TS 3.4.3 to excess flow check valve (EFCV) during
reflect that Figures 3.4.3–1 and 3.4.3–2 (P/T each refueling outage. The proposed
Duke Energy Progress, LLC, Docket No. limit curves) are applicable up to 46.3 EFPY amendments would revise the number
50–261, H. B. Robinson Steam Electric instead of 50 EFPY with the removal of of EFCVs tested by TS SR 3.6.1.3.8 from
Plant, Unit No. 2 (HBRSEP), Darlington PLSAs and migration to 24-month fuel ‘‘each’’ to a ‘‘representative sample.’’
County, South Carolina cycles. The proposed change does not affect The representative sample is based on
Date of amendment request: February the design or assumed accident performance
approximately 20 percent of the reactor
of any structure, system or component, or
7, 2018. A publicly-available version is instrumentation line EFCVs such that
introduce any new modes of system
in ADAMS under Accession No. operation or failure modes. Compliance with each EFCV will be tested at least once
ML18038B289. the proposed P/T curves (same as the existing every 10 years (nominal). Therefore,
Description of amendment request: P/T curves with the applicability term approximately 20 percent of the EFCVs
The amendment would revise Technical reduced to 46.3 EFPY) will provide sufficient will be tested every operating cycle.
Specification (TS) Section 3.4.3, ‘‘RCS protection against brittle fracture of reactor The reduced testing associated with
[Reactor Coolant System] Pressure and vessel materials to assure that the RCS the proposed change will result in an
Temperature (P/T) Limits,’’ to reduce pressure boundary performs as previously increase in the availability of the
the applicability terms from 50 effective evaluated.
associated instrumentation during
full power years (EFPY) to 46.3 EFPY in Therefore, the proposed change does not
create the possibility of a new or different outages and will result in dose savings.
Figures 3.4.3–1 and 3.4.3–2, as a result kind of accident from any accident Basis for proposed no significant
of the removal of part length fuel previously evaluated. hazards consideration determination:
assemblies (PLSAs) and the migration to 3. Does the proposed change involve a As required by 10 CFR 50.91(a), the
24-month fuel cycles. significant reduction in a margin of safety? licensee has provided its analysis of the
Basis for proposed no significant Response: No. issue of no significant hazards
hazards consideration determination: The proposed change revises TS 3.4.3 to consideration, which is presented
As required by 10 CFR 50.91(a), the reflect that Figures 3.4.3–1 and 3.4.3–2 (P/T below:
licensee has provided its analysis of the limit curves) are applicable up to 46.3 EFPY
instead of 50 EFPY with the removal of 1. Does the proposed change involve a
issue of no significant hazards significant increase in the probability or
consideration, which is presented PLSAs and migration to 24-month fuel
cycles. HBRSEP adheres to applicable NRC consequences of an accident previously
below: regulations (i.e., 10 CFR 50, Appendices G analyzed?
1. Does the proposed change involve a and H) and NRC-approved methodologies Response: No.
significant increase in the probability or (i.e., Regulatory Guides 1.99 and 1.190) with The EFCVs at LSCS, Unit 1 and Unit 2, are
consequences of an accident previously respect to the P/T limit curves in TS 3.4.3 in designed so that they will not close
evaluated? order to provide an adequate margin of safety accidently during normal operations, will
Response: No. to the conditions at which brittle fracture close if a rupture of the instrument line is
The proposed change revises TS 3.4.3 to may occur. The P/T limit curves, with the indicated downstream of the valve, can be
reflect that Figures 3.4.3–1 and 3.4.3–2 (P/T applicability term reduced to 46.3 EFPY, reopened when appropriate, and have their
limit curves) are applicable up to 46.3 EFPY continue to provide assurance that the status indicated in the control room. This
instead of 50 EFPY with the removal of established P/T limits are not exceeded. proposed change relaxes the number of
PLSAs and migration to 24-month fuel Therefore, the proposed change does not EFCVs tested for TS SR 3.6.1.3.8 from ‘‘each’’
cycles. The proposed change does not involve a significant reduction in a margin of to a ‘‘representative sample’’ in accordance
involve physical changes to the plant or alter safety. with the SFCP [Surveillance Frequency
the reactor coolant system (RCS) pressure Control Program]. There are no physical
boundary (i.e., there are no changes in The NRC staff has reviewed the plant modifications associated with this
operating pressure, materials or seismic licensee’s analysis and, based on this change. Industry and LSCS operating
loading). The P/T limit curves and Adjusted review, it appears that the three experience demonstrate a high reliability of
Reference Temperature (ART) values will standards of 10 CFR 50.92(c) are these valves. Neither EFCVs nor their failures
remain as-is. Only the term to which the satisfied. Therefore, the NRC staff are capable of initiating previously evaluated
daltland on DSKBBV9HB2PROD with NOTICES
limit curves applies is effected by the proposes to determine that the accidents; therefore, there can be no increase
proposed change. The P/T limit curves in TS in the probability of occurrence of an
amendment request involves no
3.4.3 with an applicability term of 46.3 EFPY accident regarding this proposed change.
provide continued assurance that the fracture significant hazards consideration. The LSCS Updated Final Safety Analysis
toughness of the reactor pressure vessel Attorney for licensee: Kathryn B. Report (UFSAR) demonstrates, consistent
(RPV) is consistent with analysis Nolan, Deputy General Counsel, Duke with BWROG [Boiling Water Reactor Owners
assumptions and NRC regulations. The Energy Corporation, 550 South Tryon Group] topical report NEDO–32977–A, that
methodology used to develop the existing Street, DEC45A, Charlotte NC 28202. the failure of an EFCV has very low
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![15416 Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices
consequences. The LSCS UFSAR evaluates a the calculated mass of coolant released over due to setpoint drift in the low
circumferential rupture of an instrument line approximately a five-hour period. The reactor (conservative) direction.
that is connected to the primary coolant was assumed to be operating at design power Basis for proposed no significant
system. The evaluation credits the 0.25-inch conditions prior to the break. The SGTS hazards consideration determination:
diameter flow-restricting orifice installed in [Standby Gas Treatment System] and
the line with limiting flow following the
As required by 10 CFR 50.91(a), the
secondary containment are not impaired by
instrumentation line break and does not the event. The evaluation concludes that the licensee has provided its analysis of the
credit the EFCV with actuating to limit consequences of the event are well within 10 issue of no significant hazards
leakage. The dose consequences of the CFR 100 limits. Thus, the failure of an EFCV, consideration, which is presented
instrument line break are determined using though not expected as a result of the below:
the calculated mass of coolant released over proposed change, does not affect the dose 1. Do the proposed amendments involve a
approximately a five-hour period. The reactor consequences of an instrument line break. significant increase in the probability or
was assumed to be operating at design power Therefore, the proposed changes do not consequences of an accident previously
conditions prior to the break. The Standby involve a significant reduction in a margin of evaluated?
Gas Treatment System (SGTS) and secondary safety. Response: No.
containment are not impaired by the event.
The NRC staff has reviewed the This change has no influence on the
The evaluation concludes that the
probability or consequences of any accident
consequences of the event are well within 10 licensee’s analysis and, based on this
previously evaluated. The lower setpoint
CFR 100 limits. Thus, the failure of an EFCV, review, it appears that the three tolerance change does not affect the
though not expected as a result of the standards of 10 CFR 50.92(c) are operation of the valves and it does not
proposed change, does not affect the dose satisfied. Therefore, the NRC staff change the as-left setpoint tolerance. The
consequences of an instrument line break. proposes to determine that the change only affects the lower tolerance for
Based on the above, it is concluded that the
proposed change to the EFCV surveillance
amendment request involves no valve opening and does not change the upper
significant hazards consideration. tolerance, which is the limit that protects
requirement does not involve a significant from overpressurization.
increase in the probability or consequences Attorney for licensee: Tamra Domeyer,
The proposed amendments do not involve
of an accident previously analyzed. Associate General Counsel, Exelon physical changes to the valves, nor do they
2. Does the proposed change create the Generation Company, LLC, 4300 change the safety function of the valves. The
possibility of a new or different kind of Winfield Road, Warrenville, IL 60555. proposed TS revision involves no significant
accident from any accident previously NRC Branch Chief: David J. Wrona. changes to the operation of any systems or
evaluated? components in normal or accident operating
Response: No. Exelon Generation Company, LLC, conditions and no changes to existing
This proposed change allows a reduced (EGC) Docket Nos. 50–373 and 50–374, structures, systems, or components.
number of EFCVs to be tested in accordance LaSalle County Station (LSCS), Units 1 The proposed amendments do not change
with the SFCP [Surveillance Frequency and 2, LaSalle County, Illinois any other behavior or operation of any safety/
Control Program]. The proposed change relief valves (S/RVs), and, therefore, has no
would revise SR 3.6.1.3.8 to verify that a Date of amendment request: February significant impact on reactor operation. They
‘‘representative sample’’ (i.e., approximately 27, 2018. A publicly-available version is also have no significant impact on response
20 percent) of reactor instrumentation line in ADAMS under Accession No. to any perturbation of reactor operation
EFCVs are tested, in accordance with the ML18058A257. including transients and accidents previously
SFCP, such that each EFCV will be tested at Description of amendment request: analyzed in the Updated Final Safety
least once every 10 years (nominal). No other Analysis Report (UFSAR).
changes in the requirements are being
The proposed amendments would
revise LSCS Technical Specifications Based on the above, it is concluded that the
proposed. Industry and LSCS-specific proposed change to the S/RV surveillance
operating experience demonstrates the high (TS) 3.4, Reactor Coolant System (RCS), requirement does not involve a significant
degree of reliability of the EFCVs and the low Section 3.4.4, ‘‘Safety/Relief Valves increase in the probability or consequences
consequences of an EFCV failure. The (S/RVs).’’ of an accident previously evaluated.
potential failure of an EFCV to isolate by the Specifically, EGC proposes a new 2. Do the proposed amendments create the
proposed reduction in test frequency is safety function lift setpoint lower possibility of a new or different kind of
bounded by the previous evaluation of an tolerance for the S/RVs as delineated in accident from any accident previously
instrument line rupture. This change will not evaluated?
Surveillance Requirement 3.4.4.1. The
alter the operation or process variables, Response: No.
structures, systems, or components as proposed change will revise the lower
The proposed change to the S/RV safety
described in the safety analysis. Thus, a new setpoint tolerances from ¥3 percent (%) lower setpoint tolerance from ¥3% to ¥5%
or different kind of accident will not be to ¥5%. only affects the criteria to determine when an
created from implementation of the proposed This change is limited to the lower as-found S/RV test is considered to be
change. tolerances and does not affect the upper acceptable. This change does not affect the
Therefore, the proposed changes do not tolerances; therefore, the upper criteria for the upper setpoint tolerance.
create the possibility of a new or different tolerance will remain at +3% of the The proposed lower setpoint tolerance
kind of accident from any accident safety function lift setpoint. In addition, change does not adversely affect the
previously evaluated. operation of any safety-related components
3. Does the proposed change involve a
this change only applies to the as-found or equipment. The proposed amendments do
significant reduction in a margin of safety? tolerance and not to the as-left not involve physical changes to the S/RVs,
Response: No. tolerance, which will remain unchanged nor do they change the safety function of the
The proposed changes do not involve a at ±1% of the safety lift setpoint. The as- S/RVs. The proposed amendments do not
significant reduction in the margin of safety. found tolerances are used for require any physical change or alteration of
The LSCS UFSAR evaluates a circumferential determining operability and to increase any existing plant equipment. No new or
rupture of an instrument line that is sample sizes for S/RV testing should the different equipment is being installed, and
daltland on DSKBBV9HB2PROD with NOTICES
connected to the primary coolant system. The tolerance be exceeded. There will be no installed equipment is not being operated in
evaluation credits the 0.25-inch diameter revision to the actual setpoints of the a new or different manner. There is no
flow-restricting orifice installed in the line alteration to the parameters within which the
with limiting flow following the
valves installed in the plant due to this plant is normally operated. This change does
instrumentation line break and does not change. not alter the manner in which equipment
credit the EFCV with actuating to limit This proposed change will preclude operation is initiated, nor will the functional
leakage. The dose consequences of the the submittal of previously-reportable demands on credited equipment be changed.
instrument line break are determined using licensee event reports (LERs) to the NRC No alterations in the procedures that ensure
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![Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices 15417
the plant remains within analyzed limits are Exelon Generation Company, LLC, and the proposed changes do not alter
being proposed, and no changes are being Docket Nos. 50–010, 50–237, and 50– assumptions made in the safety analysis. The
made to the procedures relied upon to 249, Dresden Nuclear Power Station, proposed changes remove ERO positions no
respond to an off-normal event as described longer credited or considered necessary in
Units 1, 2, and 3, Grundy County, support of Emergency Plan implementation.
in the UFSAR. As such, no new failure Illinois Therefore, the proposed changes to the
modes are being introduced. The change does
Exelon Generation Company, LLC, [site] Emergency Plan do not create the
not alter assumptions made in the safety possibility of a new or different kind of
analysis and licensing basis. Docket Nos. 50–373 and 50–374, LaSalle
County Station, Units 1 and 2, LaSalle accident from any accident previously
Therefore, the proposed changes do not evaluated.
create the possibility of a new or different County, Illinois 3. Does the proposed amendment involve
kind of accident from any accident Exelon Generation Company, LLC, a significant reduction in a margin of safety?
previously evaluated. Docket Nos. 50–254 and 50–265, Quad Response: No.
3. Do the proposed amendments involve a Cities Nuclear Power Station, Units 1 Margin of safety is associated with
significant reduction in a margin of safety? confidence in the ability of the fission
and 2, Rock Island County, Illinois
Response: No. product barriers (i.e., fuel cladding, reactor
Date of amendment request: January coolant system pressure boundary, and
The proposed lower setpoint tolerance
change only affects the criteria to determine
31, 2018. A publicly-available version is containment structure) to limit the level of
in ADAMS under Package Accession radiation dose to the public.
when an as-found S/RV test is considered to The proposed changes do not adversely
be acceptable. This change does not affect the
No. ML18053A159.
Description of amendment request: affect existing plant safety margins or the
criteria for the S/RV setpoint upper setpoint reliability of the equipment assumed to
tolerance. The TS setpoints for the S/RVs are
The amendments would revise the
operate in the safety analyses. There are no
not changed. The as-left setpoint tolerances
emergency response organization (ERO) changes being made to safety analysis
are not changed by this proposed change and positions identified in the emergency assumptions, safety limits, or limiting safety
remain at ±1% of the safety lift setpoint. plan for each site. system settings that would adversely affect
The margin of safety is established through
Basis for proposed no significant plant safety as a result of the proposed
the design of the plant structures, systems, hazards consideration determination: changes. Margins of safety are unaffected by
and components, the parameters within As required by 10 CFR 50.91(a), the the proposed changes to the ERO staffing.
licensee has provided its analysis of the The proposed changes are associated with
which the plant is operated, and the
issue of no significant hazards the [site] Emergency Plan staffing and do not
establishment of the setpoints for the impact operation of the plant or its response
actuation of equipment relied upon to consideration for each site, which is
to transients or accidents. The proposed
respond to an event. The proposed change presented below:
changes do not affect the Technical
does not significantly impact the condition or 1. Does the proposed amendment involve Specifications. The proposed changes do not
performance of structures, systems, and a significant increase in the probability or involve a change in the method of plant
components relied upon for accident consequences of an accident previously operation, and no accident analyses will be
mitigation. evaluated? affected by the proposed changes. Safety
Therefore, this proposed change does not Response: No. analysis acceptance criteria are not affected
involve a significant reduction in a margin of The proposed changes to the [site] by these proposed changes. The proposed
Emergency Plan do not increase the changes to the Emergency Plan will continue
safety.
probability or consequences of an accident. to provide the necessary onsite ERO response
The NRC staff has reviewed the The proposed changes do not impact the staff.
function of plant Structures, Systems, or Therefore, the proposed changes to the
licensee’s analysis and, based on this Components (SSCs). The proposed changes [site] Emergency Plan do not involve a
review, it appears that the three do not affect accident initiators or accident significant reduction in a margin of safety.
standards of 10 CFR 50.92(c) are precursors, nor do the changes alter design
satisfied. Therefore, the NRC staff assumptions. The proposed changes do not The NRC staff has reviewed the
alter or prevent the ability of the onsite ERO licensee’s analysis for each site and,
proposes to determine that the
to perform their intended functions to based on this review, it appears that the
amendment request involves no
mitigate the consequences of an accident or three standards of 10 CFR 50.92(c) are
significant hazards consideration. event. The proposed changes remove ERO satisfied. Therefore, the NRC staff
Attorney for licensee: Tamra Domeyer, positions no longer credited or considered proposes to determine that the
Associate General Counsel, Exelon necessary in support of Emergency Plan requested amendments involve no
Generation Company, LLC, 4300 implementation. significant hazards consideration.
Therefore, the proposed changes to the Attorney for licensee: Tamra Domeyer,
Winfield Road, Warrenville, IL 60555.
[site] Emergency Plan do not involve a
NRC Branch Chief: David J. Wrona. Associate General Counsel, Exelon
significant increase in the probability or
consequences of an accident previously Generation Company, LLC, 4300
Exelon Generation Company, LLC, evaluated. Winfield Road, Warrenville, IL 60555.
Docket Nos. STN 50–456 and STN 50– 2. Does the proposed amendment create NRC Branch Chief: David J. Wrona.
457, Braidwood Station, Units 1 and 2, the possibility of a new or different kind of Florida Power & Light Company, Docket
Will County, Illinois accident from any accident previously
Nos. 50–250 and 251, Turkey Point
evaluated?
Exelon Generation Company, LLC, Response: No. Nuclear Generating Unit Nos. 3 and 4,
Docket Nos. STN 50–454 and STN 50– The proposed changes have no impact on Miami-Dade County, Florida
455, Byron Station, Unit Nos. 1 and 2, the design, function, or operation of any Date of application for amendment:
Ogle County, Illinois plant SSCs. The proposed changes do not June 28, 2017, as supplemented by letter
daltland on DSKBBV9HB2PROD with NOTICES
affect plant equipment or accident analyses. dated February 28, 2018. Publicly-
Exelon Generation Company, LLC, The proposed changes do not involve a
available versions are in ADAMS under
Docket No. 50–461, Clinton Power physical alteration of the plant (i.e., no new
or different type of equipment will be Accession Nos. ML17180A447 and
Station, Unit No. 1, DeWitt County,
installed), a change in the method of plant ML18075A023, respectively.
Illinois Description of amendment request:
operation, or new operator actions. The
proposed changes do not introduce failure The amendments would modify the
modes that could result in a new accident, Technical Specifications (TSs) by
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![15418 Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices
relocating to licensee-controlled to avoid PIV repetitive loop testing do not nuclear safety. In addition, deletion of a
documents select acceptance criteria affect the likelihood or the outcome of any redundant Accumulator check valve SR
specified in TS surveillance accident previously evaluated. 4.5.1.1.d, and the addition of a footnote to TS
Therefore, facility operation in accordance SR 4.4.6.2.2.d to avoid PIV repetitive loop
requirements (SRs) credited for with the proposed changes would not testing do not affect any safety analyses
satisfying Inservice Testing (IST) involve a significant increase in the assumptions, safety limits, or limiting safety
Program and Inservice Inspection probability or consequences of an accident system settings.
Program requirements; deleting the SRs previously evaluated. Therefore, operation of the facility in
for the American Society of Mechanical 2. Does the proposed amendment create accordance with the proposed changes will
Engineers Code Class 1, 2, and 3 the possibility of a new or different kind of not involve a significant reduction in the
components; replacing references to the accident from any accident previously margin of safety.
Surveillance Frequency Control evaluated?
Response: No. The NRC staff has reviewed the
Program (SFCP) with reference to the licensee’s analysis and, based on this
The deletion of IST acceptance criteria
Turkey Point IST Program where from the TS does not affect the manner in review, it appears that the three
appropriate; establishing a Reactor which any SSC [structure, system, or standards of 10 CFR 50.92(c) are
Coolant Pump (RCP) Flywheel component] is maintained or operated and satisfied. Therefore, the NRC staff
Inspection Program; and related does not introduce new SSCs or new proposes to determine that the
editorial changes. Additionally, the methods for maintaining existing plant SSCs. amendment request involves no
amendments would delete a redundant Inservice testing will continue in the manner
and periodicity specified in the IST program
significant hazards consideration.
SR for Accumulator check valve testing Attorney for licensee: Debbie Hendel,
such that no new or different kind of
and add a footnote to the SR for Managing Attorney—Nuclear, Florida
accident can result. The addition of
Pressure Isolation Valve (PIV) testing. references to the IST Program in SR(s) where Power & Light Company, 700 Universe
The license amendment request was applicable and the deletion of references to Blvd. MS LAW/JB, Juno Beach, FL
originally noticed in the Federal the SFCP in SR testing credited by the IST 33408–0420.
Register on August 29, 2017 (82 FR Program are administrative changes and NRC Acting Branch Chief: Brian W.
41069). The notice is being reissued in cannot introduce new or different kinds of Tindell.
its entirety to include the revised scope, accidents. The deletion of SR 4.0.5 and the
description of the amendment request, relocation of the RCP flywheel inspection Tennessee Valley Authority, Docket No.
and proposed no significant hazards requirements within the TS are 50–259, Browns Ferry Nuclear Plant
consideration determination. administrative changes and cannot be an (BFN), Unit 1, Limestone County,
Basis for proposed no significant initiator of a new or different kind of Alabama
accident. Deletion of the SR 4.4.6.2.2.c
hazards consideration determination: requirement regarding returning PIV(s) to Date of amendment request: March
As required by 10 CFR 50.91(a), the service following maintenance, repair or 16, 2018. A publicly-available version is
licensee has provided its analysis of the replacement, deletion of a SR 4.5.1.1.d in ADAMS under Accession No.
issue of no significant hazards footnote previously applicable during Unit 3 ML18080A171.
consideration, which is presented Cycle 26, and the other editorial changes are Description of amendment request:
below: administrative changes and cannot introduce The amendment would revise License
new or different kinds of accidents. In
1. Does the proposed amendment involve Condition 2.C(18)(a)3 for Unit 1 that
addition, deletion of a redundant
a significant increase in the probability or requires the submittal of a revised BFN
consequences of an accident previously Accumulator check valve SR 4.5.1.1.d, and
the addition of a footnote to TS SR 4.4.6.2.2.d Unit 1 replacement steam dryer (RSD)
evaluated? analysis utilizing the BFN Unit 3 on-
Response: No. to avoid PIV repetitive loop testing do not
The proposed changes provide assurance introduce new or different kinds of accidents. dryer strain gauge based end-to-end bias
that inservice testing will be performed in the Therefore, the proposed changes do not and uncertainties at extended power
manner and within the timeframes create the possibility of a new or different conditions ‘‘at least 90 days prior to the
established by 10 CFR 50.55(a). The deletion kind of accident from any previously start of the BFN Unit 1 EPU [extended
of SR 4.0.5 and the deletion of IST evaluated. power uprate] outage.’’ Specifically, the
acceptance criteria from SR 4.5.2.c and SR 3. Does the proposed amendment involve
amendment reduces the time from 90
4.6.2.1.b neither affect the conduct nor the a significant reduction in a margin of safety?
Response: No. days to 15 days before the BFN Unit 1
periodicity of the inservice testing. The EPU outage for the submittal of the
addition of references to the IST Program in The proposed changes do not involve
SR(s) where applicable and the deletion of changes to any safety analyses assumptions, revised analysis of the RSD.
references to the SFCP in SR testing credited safety limits, or limiting safety system Basis for proposed no significant
by the IST Program are administrative in settings and do not adversely impact plant hazards consideration determination:
nature and can neither initiate nor affect the operating margins or the reliability of As required by 10 CFR 50.91(a), the
outcome of any accident previously equipment credited in safety analyses. The licensee has provided its analysis of the
evaluated. The deletion of SR 4.0.5 and the proposed changes provides assurance that issue of no significant hazards
relocation of the RCP flywheel inspection inservice inspection and inservice testing
consideration, which is presented
requirements within the TS are will be performed in the manner and within
the timeframes established by 10 CFR below.
administrative changes and cannot affect the
likelihood or the outcome of accident 50.55(a). The deletion of SR 4.0.5 and the 1. Does the proposed amendment involve
previously evaluated. Deletion of the SR relocation of the RCP flywheel inspection a significant increase in the probability or
4.4.6.2.2.c requirement regarding returning requirements within the TS are consequence of an accident previously
PIV(s) to service following maintenance, administrative changes with no impact on evaluated?
repair or replacement, deletion of a SR the margin of safety currently described the Response: No.
daltland on DSKBBV9HB2PROD with NOTICES
4.5.1.1.d footnote previously applicable Updated Final Safety Analysis Report. The proposed license amendment reduces
during Unit 3 Cycle 26, and related editorial Deletion of the SR 4.4.6.2.2.c requirement the length of time, from 90 days to 15 days,
changes are administrative changes and regarding returning PIV(s) to service prior to the outage by which a revised
cannot affect the likelihood or the outcome following maintenance, repair or analysis of the Browns Ferry Nuclear Plant
of any accident previously evaluated. In replacement, deletion of a SR 4.5.1.1.d (BFN) Unit 1 replacement steam dryer (RSD),
addition, deletion of a redundant footnote previously applicable during Unit 3 performed using an NRC-approved
Accumulator check valve SR 4.5.1.1.d, and Cycle 26, and the other editorial changes are methodology benchmarked on the BFN Unit
the addition of a footnote to TS SR 4.4.6.2.2.d administrative changes with no impact on 3 RSD, must be submitted to the NRC for
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![15420 Federal Register / Vol. 83, No. 69 / Tuesday, April 10, 2018 / Notices
minimum critical power ratio numeric The Commission’s related evaluation Description of amendment: The
values for Operating Cycle 17. of the amendments is contained in a amendments authorized changes to the
Specifically, the amendment increased Safety Evaluation dated March 26, 2018. VEGP Units 3 and 4 Combined
the numeric values of the safety limit No significant hazards consideration Operating License (COL) page 7 and
minimum critical power ratio for Nine comments received: No. COL Appendix A, Technical
Mile Point Nuclear Station, Unit 2, from Florida Power & Light Company, Docket Specifications, to make necessary
≥1.15 to ≥1.17 for two recirculation loop Nos. 50–250 and 50–251, Turkey Point changes so that there will be adequate
operation, and from ≥1.15 to ≥1.17 for Nuclear Generating Unit Nos. 3 and 4, detection of reactor coolant system and
single recirculation loop operation. Miami-Dade County, Florida main steam line leakage at all times and
Date of issuance: March 16, 2018. that the associated limits account for
Effective date: As of the date of Date of amendment request: April 9, instrumentation sensitivities not
issuance and shall be implemented 2017, as supplemented by letter dated accounted for in the current VEGP
prior to startup from the next refueling October 4, 2017. Technical Specification 3.4.9.
outage. Brief description of amendments: The Date of issuance: March 12, 2018.
Amendment No.: 167. A publicly- amendments revised the Technical Effective date: As of the date of
available version is in ADAMS under Specifications (TSs) to remove various issuance and shall be implemented
Accession No. ML18060A016; reporting requirements. Specifically, the within 30 days of issuance.
documents related to this amendment amendments removed the requirements Amendment No.: 115 (Unit 3) and 114
are listed in the Safety Evaluation to prepare the Startup Report, the (Unit 4). Publicly-available versions are
enclosed with the amendment. Annual Report, and various special in ADAMS Package Accession No.
Renewed Facility Operating License reports. In addition, the amendments ML18036A782, which includes the
No. NPF–69: Amendment revised the revised the TSs to remove the Safety Evaluation that references
Renewed Facility Operating License and completion time for restoring spent fuel documents related to these
Technical Specifications. pool water level, to address amendments.
Date of initial notice in Federal inoperability of one of the two parallel Facility Combined License Nos. NPF–
Register: February 6, 2018 (83 FR flow paths in the residual heal removal 91 and NPF–92: Amendments revised
5280). or safely injection headers for the the Facility Combined Licenses.
The Commission’s related evaluation Emergency Core Cooling Systems, and Date of initial notice in Federal
of the amendment is contained in a to make other administrative changes, Register: October 10, 2017 (82 FR
Safety Evaluation dated March 16, 2018. including updating plant staff and 47032).
No significant hazards consideration responsibilities. The Commission’s related evaluation
comments received: No. Date of issuance: March 19, 2018. of the amendment is contained in the
Effective date: As of the date of Safety Evaluation dated March 12, 2018.
Florida Power & Light Company, et al., issuance and shall be implemented
Docket Nos. 50–335 and 50–389, St. No significant hazards consideration
within 90 days of issuance. comments received: No.
Lucie Plant, Unit Nos. 1 and 2, St. Lucie Amendment Nos.: 279 (Unit No. 3)
County, Florida and 274 (Unit No. 4). A publicly- Dated at Rockville, Maryland, this 28th day
Date of amendment request: January available version is in ADAMS under of March 2018.
31, 2018. Accession No. ML18019A078; For the Nuclear Regulatory Commission.
Brief description of amendments: The documents related to these amendments Tara Inverso,
amendments revised the Emergency are listed in the Safety Evaluation (SE) Acting Deputy Director, Division of Operating
Plan for St. Lucie to adopt the fire- enclosed with the amendments. Reactor Licensing, Office of Nuclear Reactor
related notification of unusual event Renewed Facility Operating License Regulation.
requirement of the Nuclear Energy Nos. DPR–31 and DPR–41: Amendments [FR Doc. 2018–06668 Filed 4–9–18; 8:45 am]
Institute 99–01, Revision 6, Emergency revised the Renewed Facility Operating BILLING CODE 7590–01–P
Action Level scheme. Licenses and TSs.
Date of issuance: March 26, 2018. Date of initial notice in Federal
Effective date: As of the date of Register: June 19, 2017 (82 FR 27889). NUCLEAR REGULATORY
issuance and shall be implemented The supplemental letter dated October COMMISSION
within 90 days. 4, 2017, provided additional [Docket No. 72–10; NRC–2018–0057]
Amendment Nos.: 244 and 195. A information that clarified the
publicly-available version is in ADAMS application, did not expand the scope of Northern States Power Company—
under Accession No. ML18046A712; the application as originally noticed, Minnesota; Prairie Island Nuclear
documents related to these amendments and did not change the NRC staff’s Generating Plant; Independent Spent
are listed in the Safety Evaluation original proposed no significant hazards Fuel Storage Installation; Correct
enclosed with the amendments. consideration determination as Inspection Intervals Acceptance
Renewed Facility Operating License published in the Federal Register. Criteria
Nos. DPR–67 and NPF–16: The The Commission’s related evaluation
amendments revised the Renewed of the amendments is contained in an AGENCY: Nuclear Regulatory
Facility Operating Licenses. SE dated March 19, 2018. Commission.
Date of initial notice in Federal No significant hazards consideration ACTION: License amendment application;
Register: February 14, 2018 (83 FR comments received: No. issuance.
daltland on DSKBBV9HB2PROD with NOTICES
6621). This notice provided an
opportunity to request a hearing by Southern Nuclear Operating Company, SUMMARY: The U.S. Nuclear Regulatory
April 15, 2018, but indicated that if the Docket Nos. 52–025 and 52–026, Vogtle Commission (NRC) reconciled an error
Commission makes a final no significant Electric Generating Plant (VEGP), Units in the Northern States Power
hazards consideration determination, 3 and 4, Burke County, Georgia Company—Minnesota (NSPM) Renewed
any such hearing would take place after Date of amendment request: August License No. SNM–2506. Under this
issuance of the amendments. 31, 2017. license, NSPM is authorized to receive,
VerDate Sep<11>2014 16:56 Apr 09, 2018 Jkt 244001 PO 00000 Frm 00067 Fmt 4703 Sfmt 4703 E:\FR\FM\10APN1.SGM 10APN1](https://thefederalregister.org/doc/83_FR_15481/page/9.svg)
Document Created: 2018-04-09 23:51:40
Document Modified: 2018-04-09 23:51:40
| 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 May 10, 2018. A request for a hearing must be filed by June 11, 2018. | |
| Contact | Kay Goldstein, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected] | |
| FR Citation | 83 FR 15412 |
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