Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 26098-26109 [2018-11843]

Download as PDF 26098 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices Laurence Brewer, Chief Records Officer for the U.S. Government. [FR Doc. 2018–11987 Filed 6–4–18; 8:45 am] BILLING CODE 7515–01–P NUCLEAR REGULATORY COMMISSION [NRC–2018–0105] Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations Nuclear Regulatory Commission. ACTION: Biweekly notice. AGENCY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued, from May 8, 2018, to May 21, 2018. The last biweekly notice was published on May 22, 2018. DATES: Comments must be filed by July 5, 2018. A request for a hearing must be filed by August 6, 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–2018–0105. Address amozie on DSK3GDR082PROD with NOTICES1 SUMMARY: VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 questions about NRC dockets to Jennifer Borges; telephone: 301–287–9127; email: Jennifer.Borges@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. • Mail comments to: May Ma, Office of Administration, Mail Stop: TWFN–7– A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001. For additional direction on obtaining information and submitting comments, see ‘‘Obtaining Information and Submitting Comments’’ in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555–0001; telephone: 301–415–1384, email: Janet.Burkhardt@nrc.gov. SUPPLEMENTARY INFORMATION: and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http:// www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. I. Obtaining Information and Submitting Comments entities, persons, products, and offerings. 7. United States Postal Service, Office of the Regional Director, Atlanta, Georgia (DAA–0028–2018–0001, 6 items, 6 temporary items). Dedication files and site selection files of individual post offices in Florida, Georgia, North Carolina, South Carolina, and Puerto Rico. Includes personnel records and routine organization data. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment A. Obtaining Information Please refer to Docket ID NRC–2018– 0105, 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 publiclyavailable 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–2018–0105. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may obtain publiclyavailable 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 pdr.resource@nrc.gov. 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– 0105, facility name, unit number(s), plant docket number, application date, PO 00000 Frm 00104 Fmt 4703 Sfmt 4703 E:\FR\FM\05JNN1.SGM 05JNN1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices amozie on DSK3GDR082PROD with NOTICES1 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/doccollections/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 VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 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 PO 00000 Frm 00105 Fmt 4703 Sfmt 4703 26099 imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner’s interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice August 6, 2018. 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 nonparty 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 EFiling process requires participants to submit and serve all adjudicatory E:\FR\FM\05JNN1.SGM 05JNN1 amozie on DSK3GDR082PROD with NOTICES1 26100 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 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 hearing.docket@nrc.gov, 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 EFiling 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, VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 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/esubmittals.html, by email to MSHD.Resource@nrc.gov, or by a tollfree 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 PO 00000 Frm 00106 Fmt 4703 Sfmt 4703 you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC’s PDR. For additional direction on accessing information related to this document, see the ‘‘Obtaining Information and Submitting Comments’’ section of this document. Duke Energy Carolinas, LLC, Docket Nos. 50–369 and 50–370, McGuire Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North Carolina Date of amendment request: December 8, 2017. A publicly-available version is in ADAMS under Accession No. ML17352A404. Description of amendment request: The amendments would modify the MNS, Unit Nos. 1 and 2 Updated Final Safety Analysis Report (UFSAR) to describe the methodology and results of the analyses performed to evaluate the protection of the plant’s structures, systems, and components from tornadogenerated missiles. 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 to the MNS UFSAR constitutes a license amendment to incorporate use of a Nuclear Regulatory Commission (NRC) approved probabilistic methodology to assess the need for additional positive (physical) tornado missile protection of specific features at the MNS site. The UFSAR changes will reflect use of the Electric Power Research Institute (EPRI) Topical Report ‘‘Tornado Missile Risk E:\FR\FM\05JNN1.SGM 05JNN1 amozie on DSK3GDR082PROD with NOTICES1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices Evaluation Methodology’’ (EPRI NP–2005), Volumes I and II. As noted in the NRC Safety Evaluation Report on this topic dated October 26, 1983, the current licensing criteria governing tornado missile protection are contained in NUREG–0800, Sections 3.5.1.4 and 3.5.2. These criteria generally specify that safety-related systems, structures and components be provided positive tornado missile protection (barriers) from the maximum credible tornado threat. However, NUREG–0800 includes acceptance criteria permitting relaxation of the above deterministic guidance, if it can be demonstrated that the probability of damage to unprotected essential safety-related features is sufficiently small. As permitted in NUREG–0800 sections, the combined probability will be maintained below an allowable level, i.e., an acceptance criterion threshold, which reflects an extremely low probability of occurrence. The approach assumes that if the sum of the individual probabilities calculated for tornado missiles striking and damaging portions of important systems, structures or components is greater than or equal to 1 × 10¥6 per year per unit, then installation of unique missile barriers would be needed to lower the total cumulative probability below the acceptance criterion of 1 × 10¥6 per year per unit. With respect to the probability of occurrence or the consequences of an accident previously evaluated in the UFSAR, the possibility of a tornado reaching the site and causing damage to plant structures, systems and components is considered in the MNS UFSAR. The change being proposed does not affect the probability that the natural phenomenon (a tornado) will reach the plant, but from a licensing basis perspective, the change does affect the probability that missiles generated by the winds of the tornado might strike and damage certain plant structures, systems and components. There are a limited number of safety-related components that could theoretically be struck and damaged by tornadogenerated missiles. The probability of tornado-generated missile strikes on important to safety structures, systems and components is what was analyzed using the probabilistic methods discussed above. The combined probability of damage will be maintained below an extremely low acceptance criterion to ensure overall plant safety. The proposed change is not considered to constitute a significant increase in the probability of occurrence or the consequences of an accident, due to the extremely low probability of damage due to tornado-generated missiles and thus an extremely low probability of a radiological release. The results of the analysis documented in this [license amendment request (LAR)] are below the acceptance criterion of 1 × 10¥6 per year per unit. 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? VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 Response: No. The proposed changes to the MNS UFSAR incorporate use of a NRC approved probabilistic methodology to assess the need for additional positive (physical) tornado missile protection for specific features. This will not change the design function or operation of any structure, system or component. This proposed change does not involve any plant modifications. There are no new credible failure mechanisms, malfunctions or accident initiators not considered in the design and licensing bases for MNS. The proposed change involves an already established tornado design basis event and the tornado event is explicitly considered in the MNS UFSAR. 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 the margin of safety? Response: No. The existing licensing basis for MNS for protecting safety-related, safe shutdown equipment from tornado generated missiles is to provide positive missile barriers for all safety-related structures, systems and components. The proposed change recognizes that there is an extremely low probability, below an established acceptance limit, that a limited subset of the safetyrelated, safe shutdown structures, systems and components could be struck and consequently damaged. The change from requiring protection of all safety-related, safety shutdown structures, systems and components from tornadogenerated missiles, to only a subset of equipment, is not considered to constitute a significant decrease in the margin of safety due to that extremely low probability of occurrence of tornado-generated missile strikes and consequential damage. 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: Kate B. Nolan, Deputy General Counsel, Duke Energy Carolinas, LLC, 550 South Tryon Street—DEC45A, Charlotte, NC 28202– 1802. NRC Branch Chief: Michael T. Markley. Duke Energy Progress, LLC, Docket No. 50–261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of amendment request: April 5, 2018. A publicly-available version is in ADAMS under Accession No. ML18099A130. PO 00000 Frm 00107 Fmt 4703 Sfmt 4703 26101 Description of amendment request: The proposed amendment would revise the licensing basis, by the addition of a license condition, to allow for the implementation of the provisions of 10 CFR 50.69, ‘‘Risk-informed categorization and treatment of structures, systems, and components [SSCs] for nuclear power reactors.’’ 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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions. 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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed. 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 will permit the use of a risk-informed categorization process to E:\FR\FM\05JNN1.SGM 05JNN1 26102 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results. Therefore, the proposed change does not involve a significant reduction in a margin of safety. amozie on DSK3GDR082PROD with NOTICES1 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. Entergy Operations, Inc., Docket No. 50– 313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Date of amendment request: March 12, 2018, as supplemented by letter dated April 26, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML18071A319 and ML18117A493, respectively. Description of amendment request: The amendment would revise the Arkansas Nuclear One, Unit No. 1 Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, ‘‘Relocate Surveillance Frequencies to Licensee Control— RITSTF [Risk-informed TSTF] Initiative 5b.’’ Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.5, ‘‘Programs and Manuals.’’ 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 any accident previously evaluated? Response: No. VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 The proposed change relocates the specified frequencies for periodic surveillance requirements (SRs) to licensee control under a new Surveillance Frequency Control Program [SFCP]. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications (TSs) for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. 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 the margin of safety? Response: No. The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Entergy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI [Nuclear Energy Institute] 04–10, Rev. 1 in accordance with the TS SFCP. NEI 04–10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177. 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 PO 00000 Frm 00108 Fmt 4703 Sfmt 4703 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L–ENT–WDC, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc., Docket No. 50– 368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas Date of amendment request: February 6, 2018, as supplemented by letter dated March 26, 2018. Publicly-available versions are in ADAMS under Accession Nos. ML18038B354, and ML18085A816, respectively. Description of amendment request: The amendment would revise the Arkansas Nuclear One, Unit No. 2 Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the adoption of Technical Specifications Task Force (TSTF)-425, Revision 3, ‘‘Relocate Surveillance Frequencies to Licensee Control— RITSTF [Risk-Informed TSTF] Initiative 5b.’’ The amendment would also add a new program, the Surveillance Frequency Control Program, to TS Section 6.0, ‘‘Administrative Controls.’’ 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 any accident previously evaluated? Response: No. The proposed change relocates the specified frequencies for periodic Surveillance Requirements (SRs) to licensee control under a new Surveillance Frequency Control Program (SFCP). Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the TSs for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. E:\FR\FM\05JNN1.SGM 05JNN1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. 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 the margin of safety? Response: No. The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Entergy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI [Nuclear Energy Institute] 04–10, Rev. 1, in accordance with the TS SFCP. NEI 04–10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. amozie on DSK3GDR082PROD with NOTICES1 The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L–ENT–WDC, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative Energy, A Mississippi Electric Cooperative; and Entergy Mississippi, Inc., Docket No. 50– 416, Grand Gulf Nuclear Station, Unit No. 1, Claiborne County, Mississippi Date of amendment request: April 10, 2018. A publicly-available version is in VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 ADAMS under Accession No. ML18100B304. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF–542, Revision 2, ‘‘Reactor Pressure Vessel Water Inventory Control.’’ The proposed change would replace existing TS requirements related to ‘‘operations with a potential for draining the reactor vessel’’ (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated. The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event. The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode PO 00000 Frm 00109 Fmt 4703 Sfmt 4703 26103 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed. The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to E:\FR\FM\05JNN1.SGM 05JNN1 26104 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin. 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: Anna Vinson Jones, Senior Counsel/Legal Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. amozie on DSK3GDR082PROD with NOTICES1 Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative Energy, A Mississippi Electric Cooperative; and Entergy Mississippi, Inc., Docket No. 50– 416, Grand Gulf Nuclear Station, Unit No. 1 (GGNS), Claiborne County, Mississippi Date of amendment request: April 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18117A514. Description of amendment request: The proposed amendment would revise the Emergency Plan to adopt the Nuclear Energy Institute’s (NEI’s) revised Emergency Action Level (EAL) scheme described in NEI 99–01, Revision 6, ‘‘Development of Emergency Action Levels for Non-Passive Reactors’’ (ADAMS Accession No. ML110240324), which has been endorsed by the NRC (ADAMS Accession No. ML12346A463). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 Response: No. The proposed changes to the GGNS EALs do not involve any physical changes to plant equipment or systems and do not alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors and do not alter design assumptions, plant configuration, or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems or components (SSCs) to perform intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. Because EALs are not accident initiators and no physical changes are made to the plant, no new causal mechanisms are introduced. Therefore, the changes do not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. Margin of safety is associated with 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 impact operation of the plant and no accident analyses are affected by the proposed changes. The changes do not affect the Technical Specifications or the method of operating the plant. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. PO 00000 Frm 00110 Fmt 4703 Sfmt 4703 Attorney for licensee: Anna Vinson Jones, Senior Counsel/Legal Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Exelon Generation Company, LLC, Docket Nos. STN 50–456 and STN 50– 457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois, and Docket Nos. STN 50–454 and STN 50–455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Date of amendment request: April 2, 2018. A publicly-available version is in ADAMS under Accession No. ML18092B081. Description of amendment request: The proposed amendments would revise Technical Specification 3.2.3 to require that the axial flux difference be maintained within the limits specified in the core operating limits report during MODE 1 with reactor thermal power greater or equal to 50 percent. An associated change would also be made to the NOTE modifying surveillance 3.2.3.1. 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 amendment requires that the AFD [axial flux difference] be maintained within the limits specified in the COLR [core operating limits report] at-all-times during MODE 1 when reactor power is ≥50% RTP [reactor thermal power]. This requirement will ensure that all FRD [fuel rod design] performance criteria remain satisfied during ANS [American Nuclear Society] Condition II events (i.e., Faults of Moderate Frequency); thus, ensuring the integrity of the fuel rod cladding. It is noted that maintaining AFD within the COLR limits at-all-times when ≥50% RTP is the normal operating practice as specified in plant procedures. The proposed change will have no impact on accident initiators or precursors; does not alter accident analysis assumptions; does not involve any physical plant modifications that would alter the design or configuration of the facility, or the manner in which the plant is maintained; and does not impact the probability of operator error. The proposed amendment will not impact the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an accident. All accident analysis acceptance criteria will continue to be met as the proposed change will not affect E:\FR\FM\05JNN1.SGM 05JNN1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices amozie on DSK3GDR082PROD with NOTICES1 the source term, containment isolation function, or radiological release assumptions for any accident previously evaluated. Based on the above discussion, the proposed amendment 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 formalizes the existing operating practice of maintaining the AFD within the limits specified in the COLR at-all-times during MODE 1 when reactor power is ≥ 50% RTP. This change ensures that all FRD performance criteria remain satisfied during ANS Condition II events. The ANS Condition II events have all been previously evaluated in the Updated Final Safety Analysis Report. The proposed change does not involve a design change or other changes that would impact safety-related SSCs from performing their specified safety functions. The proposed change does not result in the creation of any new accident precursors; does not result in changes to any existing accident scenarios; and does not introduce any operational changes or mechanisms that would create the possibility of a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change to maintain the AFD within the limits specified in the COLR atall-times during MODE 1 when reactor power is ≥ 50% RTP ensures that all FRD performance criteria remain satisfied during ANS Condition II events; and thus, will maintain the existing margin of safety related to FRD performance criteria and ensure the integrity of the fuel rod cladding. The AFD limits specified in the COLR have been established in accordance with the analysis approach described in NRC-approved Westinghouse Topical Reports. In addition, this change will have no impact on the margin of safety associated with other reactor core safety parameters such as fuel hot channel factors, core power tilt ratios, loss of coolant accident peak cladding temperature and peak local power density. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. FirstEnergy Nuclear Operating Company, Docket No. 50–412, Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania Date of amendment request: March 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18087A293. Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.5.2.d, ‘‘Provisions for SG [Steam Generator] Tube Inspection,’’ and TS 5.5.5.2.f, ‘‘Provisions for SG Tube Repair Methods.’’ More specifically, TSs 5.5.5.2.d.5 and 5.5.5.2.f.3 would be simplified and clarified, respectively, without changing the intent of the specifications. Specification 5.5.5.2.f.3 would also be amended by changing the number of fuel cycles that Westinghouse Electric Company, LLC leak-limiting Alloy 800 sleeves may remain in operation. 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. Proposed amendment of Technical Specification 5.5.5.2.d.5 to simplify the description of the required inspection region, and Technical Specification 5.5.5.2.f.3 to clarify that this specification is only applicable to sleeves installed in the steam generator tubesheet and change the number of fuel cycles that an Alloy 800 steam generator tubesheet sleeve may remain in service from five to eight fuel cycles of operation, does not affect structures, systems or components of the plant, plant operations, design functions or analyses that verify the capability of structures, systems or components to perform a design function. The proposed amendment does not increase the likelihood of steam generator tube sleeve leakage. The proposed amendment of Technical Specification 5.5.5.2.d.5 to simplify the description of the required inspection region, makes it clear that the steam generator parent tube is to be inspected in the areas where the joints will be established prior to installation of the sleeve, regardless of the sleeve location. This proposed amendment does not change the intent of the specification. The proposed amendment of TS 5.5.5.2.f.3 includes two changes. The first change would add the words ‘‘installed in the hotleg or cold-leg tubesheet region’’ after the words ‘‘An Alloy 800 sleeve’’ to make it clear PO 00000 Frm 00111 Fmt 4703 Sfmt 4703 26105 that the specification only applies to Alloy 800 tube sleeves installed in the steam generator tubesheet. The design of Alloy 800 sleeves installed in steam generator tube locations other than the tubesheet does not include a nickel band. For these sleeves, nondestructive examination methods have been demonstrated to be effective and limits on sleeve operating life are not necessary. This proposed amendment does not change the intent of the specification. The second change to TS 5.5.5.2.f.3, increases the number of fuel cycles Alloy 800 tube sleeves installed in the tubesheet may remain in service. The leak-limiting Alloy 800 sleeves are designed using the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and, therefore, meet the design objectives of the original steam generator tubing. The applied stresses and fatigue usage for the sleeves are bounded by the limits established in the ASME Code. Mechanical testing has shown that the structural strength of sleeves under normal, upset, emergency, and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting (three times normal operating pressure differential) burst margin of NRC Regulatory Guide 1.121, ‘‘Bases for Plugging Degraded PWR Steam Generator Tubes.’’ The leak-limiting Alloy 800 sleeve depthbased structural limit is determined using NRC guidance and the pressure stress equation of ASME Code, Section III with margin added to account for the configuration of long axial cracks. Calculations show that a depth-based limit of 45 percent through-wall degradation is acceptable. However, Technical Specifications 5.5.5.2.c.2 and 5.5.5.2.c.3 provide additional margin by requiring an Alloy 800 sleeved tube to be plugged on detection of any flaw in the sleeve or in the pressure boundary portion of the original tube wall in the sleeve to tube joint. Degradation of the original tube adjacent to the nickel band of an Alloy 800 sleeve installed in the tubesheet, regardless of depth, would not prevent the sleeve from satisfying design requirements. Thus, flaw detection capabilities within the original tube adjacent to the sleeve nickel band are a defense-in-depth measure, and are not necessary in order to justify continued operation of the sleeved tube. Evaluation of repaired steam generator tube testing and analysis indicates that there are no detrimental effects on the leak-limiting Alloy 800 sleeve or sleeved tube assembly from reactor coolant system flow, primary or secondary coolant chemistries, thermal conditions or transients, or pressure conditions that may be experienced at Beaver Valley Power Station, Unit No. 2. Westinghouse is not aware of, and has no knowledge of any reports of parent-tube stress corrosion cracking (SCC) in the sleeve roll joint region for any Westinghouse sleeve design. The proposed increase in the number of fuel cycles Alloy 800 tube sleeves installed in the tubesheet may remain in service has no effect on sleeve operation or capability of the sleeve to perform its design function. The mechanical and leakage tests have confirmed E:\FR\FM\05JNN1.SGM 05JNN1 amozie on DSK3GDR082PROD with NOTICES1 26106 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices that degradation of the parent tube adjacent to the nickel band will not prevent the sleeve from satisfying its design function. Consequences of a hypothetical failure of the leak-limiting Alloy 800 sleeve and tube assembly are bounded by the current main steam line break and steam generator tube rupture accident analyses described in the Beaver Valley Power Station, Unit No. 2 Updated Final Safety Analysis Report. The total number of plugged steam generator tubes (including equivalency associated with installed sleeves) is required to be consistent with accident analysis assumptions. The sleeve and tube assembly leakage during plant operation is required to be within the allowable Technical Specification leakage limits and accident analysis assumptions. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Proposed amendment of Technical Specification 5.5.5.2.d.5 to simplify the description of the required inspection region, and Technical Specification 5.5.5.2.f.3 to clarify that this specification is only applicable to sleeves installed in the steam generator tubesheet do not change the intent of these specifications, and do not affect the design function or operation of the tube sleeves. The proposed amendment of Technical Specification 5.5.5.2.f.3 to change the number of fuel cycles that an Alloy 800 steam generator tubesheet sleeve may remain in service from five to eight fuel cycles of operation, does not affect the design function or operation of the tube sleeves. Since these changes do not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design or licensing bases, the changes do not create the possibility of a new or different kind of accident from any previously evaluated. The leak-limiting Alloy 800 sleeves are designed using the applicable ASME Code, and therefore meet the objectives of the original steam generator tubing. As a result, the functions of the steam generator will not be significantly affected by the installation of the proposed sleeve. Therefore, the only credible failure modes for the sleeve and tube are to leak or rupture, which has already been evaluated. The continued integrity of the installed sleeve and tube assembly is periodically verified as required by the Technical Specifications, and a sleeved tube will be plugged on detection of a flaw in the sleeve or in the pressure boundary portion of the original tube wall in the sleeve to tube joint. The proposed amendment to Technical Specification 5.5.5.2.f.3 increases the number of fuel cycles Alloy 800 tube sleeves installed in the tubesheet may remain in service to eight fuel cycles of operation. Implementation of this proposed amendment has no significant effect on either the configuration of the plant, the manner in which it is operated, or ability of the sleeve to perform its design function. VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Proposed amendment of Technical Specification 5.5.5.2.d.5 to simplify the description of the required inspection region, and Technical Specification 5.5.5.2.f.3 to clarify that this specification is only applicable to sleeves installed in the steam generator tubesheet, do not change the intent of these requirements or reduce the margin of safety. The proposed amendment to Technical Specification 5.5.5.2.f.3 to change the number of fuel cycles that an Alloy 800 steam generator tubesheet sleeve may remain in service from five to eight fuel cycles of operation, does not affect a design basis or safety limit (that is, the controlling numerical value for a parameter established in the Updated Final Safety Analysis Report or the license) or reduce the margin of safety. The proposed amendment to Technical Specification 5.5.5.2.f.3 increases the number of fuel cycles Alloy 800 tube sleeves installed in the tubesheet may remain in service to eight fuel cycles of operation. Implementation of this proposed amendment would not affect a design basis or safety limit or reduce the margin of safety. The repair of degraded steam generator tubes with leaklimiting Alloy 800 sleeves restores the structural integrity of the degraded tube under normal operating and postulated accident conditions. Minimum reactor coolant system flow rate from the cumulative effect of repaired (sleeved) and plugged tubes will be greater than the flow rate limit established in the Technical Specification limiting condition for operation 3.4.1. The design safety factors utilized for the sleeves are consistent with the safety factors in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code used in the original steam generator design. Tubes with sleeves are subject to the same safety factors as the original tubes, which are described in the performance criteria for steam generator tube integrity in the existing Technical Specifications. The sleeve and portions of the installed sleeve and tube assembly that represent the reactor coolant pressure boundary will be monitored, and a sleeved tube will be plugged if a flaw is detected in the sleeve or in the pressure boundary portion of the original tube wall in the leaklimiting sleeve and tube assembly. Use of the previously-identified design criteria and design verification testing ensures that the margin of safety is not significantly different from the original steam generator tubes. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. PO 00000 Frm 00112 Fmt 4703 Sfmt 4703 Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: James Danna. PSEG Nuclear LLC, Docket No. 50–354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: March 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18087A095. Description of amendment request: The amendment would revise Technical Specification (TS) 3/4.8.1, ‘‘AC [Alternating Current] Sources— Operating’’; specifically, ACTION b concerning one inoperable emergency diesel generator (EDG). The proposed change would remove the Salem Nuclear Generating Station, Unit No. 3 (Salem Unit 3), gas turbine generator and replace it with portable diesel generators. 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 removes the requirement for the Salem Unit 3 gas turbine generator (GTG) and replaces it with the supplemental power source during the existing extended allowable outage time for the A or B EDG. The emergency diesel generators are safety related components which provide backup electrical power supply to the onsite Safeguards Distribution System. The emergency diesel generators are not accident initiators; the EDGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power. (During normal operation, the proposed portable diesel generators will not be connected to the plant.) The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed change is consistent with safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the E:\FR\FM\05JNN1.SGM 05JNN1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 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 removes the requirement for the Salem Unit 3 gas turbine generator (GTG) and replaces it with the supplemental power source during the existing extended allowable outage time for the A or B EDG. The proposed change does not alter or involve any design basis accident initiators. Equipment will be operated in the same configuration and manner that is currently allowed and designed for. 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 does not alter the permanent plant design, including instrument set points, nor does it change the assumptions contained in the safety analyses. The proposed change does not impact the redundancy or availability requirements of offsite power supplies or change the ability of the plant to cope with station blackout [(SBO)] events. The EDGs continue to meet their design requirements; there is no reduction in capability or change in design configuration. The EDG response to LOOP [loss of offsite power], LOCA [loss-of-coolant accident], SBO, or fire is not changed by this proposed amendment; there is no change to the EDG operating parameters. The remaining operable emergency diesel generators are adequate to supply electrical power to the onsite Safeguards Distribution System. The proposed change does not alter a design basis or safety limit; therefore it does not significantly reduce the margin of safety. The EDGs will continue to operate per the existing design and regulatory requirements. Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety. amozie on DSK3GDR082PROD with NOTICES1 The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC–N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: James G. Danna. Tennessee Valley Authority, Docket Nos. 50–327 and 50–328, Sequoyah Nuclear Plant, Unit Nos. 1 and 2 (SQN), Hamilton County, Tennessee Date of amendment request: March 9, 2018, as supplemented by letter dated April 11, 2018. Publicly-available versions are in ADAMS under VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 Accession Nos. ML18071A349 and ML18102B430, respectively. Description of amendment request: The amendments would make changes to the SQN Essential Raw Cooling Water (ERCW) Motor Control Centers (MCCs) and revise the Updated Final Safety Analysis Report (UFSAR). 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 change does not alter the safety function of any structure, system, or component, does not modify the manner in which the plant is operated, and does not alter equipment out-of-service time. In addition, this request does not degrade the ability of the ERCW to perform its intended safety function. Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve any physical changes to plant safety related structure, system or component or alter the modes of plant operation in a manner that is outside the bounds of the system design analyses. The proposed change to complete the design change for the removal of mechanical interlock device from the feeder breakers and tie breakers for the ERCW MCCs and to revise the ERCW System Description in Section 9.2.2.2 of the SQN UFSAR to describe the normal and alternate power sources for the ERCW system does not create the possibility for an accident or malfunction of a different type than any evaluated previously in SQN’s UFSAR. The proposal does not alter the way any safety related structure, system or component functions and does not modify the manner in which the plant is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to remove the mechanical interlock device from the feeder breakers and tie breakers for ERCW MCCs 1B–B and 2B–B and to revise the ERCW System Description in Section 9.2.2.2 of the SQN UFSAR to describe the normal and alternate power sources for the ERCW system does not reduce the margin of safety because ERCW will continue to perform its safety function. The design features provided by the mechanical interlock device are not described in the SQN UFSAR, are not PO 00000 Frm 00113 Fmt 4703 Sfmt 4703 26107 credited in the SQN accident analysis and do not provide any additional safety margin. The results of accident analyses remain unchanged by this request. 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. Vistra Operations Company LLC, Docket Nos. 50–445 and 50–446, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: March 29, 2018. A publicly-available version is in ADAMS under Accession No. ML18102A516. Description of amendment request: The amendments would revise Technical Specification 3.3.2, ‘‘Engineered Safety Feature Actuation System (ESFAS) Instrumentation,’’ to change the applicability of when the automatic auxiliary feedwater actuation due to the trip of all main feedwater pumps is required to be operable at Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The design basis events which impose auxiliary feedwater safety function requirements are loss of all AC [alternating current] power to plant auxiliaries, loss of normal feedwater, steam generator fault in either the feedwater or steam lines, and small break loss of coolant accidents. These design basis event evaluations assume actuation of auxiliary feedwater due to station blackout, low-low steam generator level or a safety injection signal. The anticipatory auxiliary feedwater automatic start signals from the main feedwater pumps are not credited in any design basis accidents and are, therefore, not part of the primary success path for postulated accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion 3. Modifying MODE 2 Applicability for this function will not impact any previously evaluated design basis accidents. E:\FR\FM\05JNN1.SGM 05JNN1 26108 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. This technical specification change allows for an operational allowance during MODE 2 while placing main feedwater pumps in service. This change involves an anticipatory auxiliary feedwater automatic start function that is not credited in the accident analysis. Since this change only affects the conditions at which this automatic start function needs to be operable and does not affect the function that actuates auxiliary feedwater due to loss of offsite power, low-low steam generator level or a safety injection signal, it will not be an initiator to a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. This technical [s]pecification change involves the automatic start of the auxiliary feedwater pumps due to trip of both main feedwater pumps, which is not an assumed start signal for design basis events. This change does not modify any values or limits involved in a safety related function or accident analysis. Therefore, the proposed change does not involve a significant reduction in a margin of safety. amozie on DSK3GDR082PROD with NOTICES1 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: Timothy P. Matthews, Esq., Morgan, Lewis, and Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004. NRC Branch Chief: Robert J. Pascarelli. 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 VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 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. FirstEnergy Nuclear Operating Company, Docket No. 50–440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: June 8, 2017. Brief description of amendment: The amendment revised technical specifications (TSs) to reflect previously approved changes made as part of the alternative source term initiative. The amendment revised the surveillance requirements for the control room emergency recirculation and annulus exhaust gas treatment systems, which are consistent with Technical Specification Task Force (TSTF) Traveler TSTF–522, ‘‘Revise Ventilation System Surveillance Requirement to Operate for 10 Hours per Month.’’ The amendment also deleted two TS sections related to the fuel handling building and fuel handling building ventilation exhaust system and increased the allowable secondary containment leakage. Lastly, the amendment revised the TS Table of Contents to reflect administrative changes to the titles of TS sections. Date of issuance: May 16, 2018. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. PO 00000 Frm 00114 Fmt 4703 Sfmt 4703 Amendment No.: 180. A publiclyavailable version is in ADAMS under Accession No. ML18110A133; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF– 58: The amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: August 1, 2017 (82 FR 35841). The supplemental letter dated January 30, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff’s original proposed no significant hazards consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated May 16, 2018. No significant hazards consideration comments received: No. NextEra Energy Duane Arnold, LLC, Docket No. 50–331, Duane Arnold Energy Center (DAEC), Linn County, Iowa Date of amendment request: March 24, 2017. Brief description of amendment: The amendment revised the DAEC Technical Specification (TS) Table 3.3.2.1–1, ‘‘Control Rod Block Instrumentation,’’ by relocating certain cycle-specific Minimum Critical Power Ratio values to the DAEC Core Operating Limits Report. The amendment also added a requirement to DAEC TS 5.6.5, ‘‘Core Operating Limits Report.’’ Date of issuance: March 7, 2018. Effective date: As of the date of its issuance and shall be implemented by September 27, 2018. (Note: This Notice of Issuance corrects the ‘‘Effective date’’ of Amendment No. 303 originally noticed in the Federal Register on March 27, 2018 (83 FR 13153). Amendment No.: 303. A publiclyavailable version is in ADAMS under Accession No. ML18011A059; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Amendment No. 303 was corrected by letter dated May 7, 2018 (ADAMS Accession No. ML18081A074). Renewed Facility Operating License No. DPR–49: The amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: May 23, 2017 (82 FR 23627). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 7, 2018. E:\FR\FM\05JNN1.SGM 05JNN1 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 29th day of May, 2018. For the Nuclear Regulatory Commission. Gregory F. Suber, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2018–11843 Filed 6–4–18; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Meeting of the Advisory Committee on Reactor Safeguards (ACRS) Subcommittee on APR1400 The ACRS Subcommittee on APR1400 will hold a meeting on June 5, 2018, at 11545 Rockville Pike, Room T–2B1, Rockville, Maryland 20852. The meeting will be open to public attendance with the exception of portions that may be closed to protect information that is proprietary pursuant to 5 U.S.C. 552b(c)(4). The agenda for the subject meeting shall be as follows: amozie on DSK3GDR082PROD with NOTICES1 Tuesday, June 5, 2018, 8:30 a.m. Until 5:00 p.m. The Subcommittee will review the APR1400 Design Control Document and Safety Evaluation Report with No Open Items, Chapter 17 (Quality Assurance & Reliability Assurance), Chapter 19.1 (Probabilistic Risk Assessment), and Chapter 19.2 (Severe Accident Evaluation). The Subcommittee will hear presentations by and hold discussions with the NRC staff and Korea Hydro & Nuclear Power Company regarding this matter. The Subcommittee will gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the Full Committee. Members of the public desiring to provide oral statements and/or written comments should notify the Designated Federal Official (DFO), Christopher Brown (Telephone 301–415–7111 or Email: Christopher.Brown@nrc.gov) five days prior to the meeting, if possible, so that appropriate arrangements can be made. Thirty-five hard copies of each presentation or handout should be provided to the DFO thirty minutes before the meeting. In addition, one electronic copy of each presentation should be emailed to the DFO one day before the meeting. If an electronic copy cannot be provided within this timeframe, presenters should provide the DFO with a CD containing each VerDate Sep<11>2014 20:19 Jun 04, 2018 Jkt 241001 presentation at least thirty minutes before the meeting. Electronic recordings will be permitted only during those portions of the meeting that are open to the public. Detailed procedures for the conduct of and participation in ACRS meetings were published in the Federal Register on October 4, 2017 (82 FR 46312). The bridgeline number for this meeting is 866–822–3032, passcode 8272423#. Detailed meeting agendas and meeting transcripts are available on the NRC website at http://www.nrc.gov/readingrm/doc-collections/acrs. Information regarding topics to be discussed, changes to the agenda, whether the meeting has been canceled or rescheduled, and the time allotted to present oral statements can be obtained from the website cited above or by contacting the identified DFO. Moreover, in view of the possibility that the schedule for ACRS meetings may be adjusted by the Chairman as necessary to facilitate the conduct of the meeting, persons planning to attend should check with these references if such rescheduling would result in a major inconvenience. If attending this meeting, please enter through the One White Flint North Building, 11555 Rockville Pike, Rockville, Maryland 20852. After registering with Security, please contact Ms. Kendra Freeland (Telephone 301– 415–6207) to be escorted to the meeting room. Dated: May 23, 2018. Mark L. Banks, Chief, Technical Support Branch, Advisory Committee on Reactor Safeguards. [FR Doc. 2018–12022 Filed 6–4–18; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Nuscale; Notice of Meeting The ACRS Subcommittee on NuScale will hold a meeting on June 6, 2018, at 11545 Rockville Pike, Room T–2B1, Rockville, Maryland 20852. The meeting will be open to public attendance. The agenda for the subject meeting shall be as follows: Wednesday, June 6, 2018, 8:30 a.m. Until 12:00 p.m. The Subcommittee will review the staff’s SER with open items for Chapter 8, ‘‘Electrical Systems,’’ of the NuScale PO 00000 Frm 00115 Fmt 4703 Sfmt 4703 26109 design certification application. The Subcommittee will hear presentations by and hold discussions with the NRC staff and other interested persons regarding this matter. The Subcommittee will gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the Full Committee. Members of the public desiring to provide oral statements and/or written comments should notify the Designated Federal Official (DFO), Michael Snodderly (Telephone 301–415–2241 or Email: Michael.Snodderly@nrc.gov) five days prior to the meeting, if possible, so that appropriate arrangements can be made. Thirty-five hard copies of each presentation or handout should be provided to the DFO thirty minutes before the meeting. In addition, one electronic copy of each presentation should be emailed to the DFO one day before the meeting. If an electronic copy cannot be provided within this timeframe, presenters should provide the DFO with a CD containing each presentation at least thirty minutes before the meeting. Electronic recordings will be permitted only during those portions of the meeting that are open to the public. Detailed procedures for the conduct of and participation in ACRS meetings were published in the Federal Register on October 4, 2017 (82 FR 46312). The bridgeline number for this meeting is 866–822–3032, passcode 8272423#. Detailed meeting agendas and meeting transcripts are available on the NRC website at http://www.nrc.gov/readingrm/doc-collections/acrs. Information regarding topics to be discussed, changes to the agenda, whether the meeting has been canceled or rescheduled, and the time allotted to present oral statements can be obtained from the website cited above or by contacting the identified DFO. Moreover, in view of the possibility that the schedule for ACRS meetings may be adjusted by the Chairman as necessary to facilitate the conduct of the meeting, persons planning to attend should check with these references if such rescheduling would result in a major inconvenience. If attending this meeting, please enter through the One White Flint North building, 11555 Rockville Pike, Rockville, Maryland. After registering with Security, please contact Mr. Theron Brown (Telephone 301–415– 6702 or 301–415–8066) to be escorted to the meeting room. E:\FR\FM\05JNN1.SGM 05JNN1

Agencies

[Federal Register Volume 83, Number 108 (Tuesday, June 5, 2018)]
[Notices]
[Pages 26098-26109]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-11843]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2018-0105]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

-----------------------------------------------------------------------

SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from May 8, 2018, to May 21, 2018. The last 
biweekly notice was published on May 22, 2018.

DATES: Comments must be filed by July 5, 2018. A request for a hearing 
must be filed by August 6, 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-2018-0105. Address 
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: May Ma, Office of Administration, Mail 
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1384, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0105, 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-2018-0105.
     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-2018-0105, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 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

[[Page 26099]]

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 the date of publication of this notice August 6, 
2018. 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

[[Page 26100]]

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 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: December 8, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17352A404.
    Description of amendment request: The amendments would modify the 
MNS, Unit Nos. 1 and 2 Updated Final Safety Analysis Report (UFSAR) to 
describe the methodology and results of the analyses performed to 
evaluate the protection of the plant's structures, systems, and 
components from tornado-generated missiles.
    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 to the MNS UFSAR constitutes a license 
amendment to incorporate use of a Nuclear Regulatory Commission 
(NRC) approved probabilistic methodology to assess the need for 
additional positive (physical) tornado missile protection of 
specific features at the MNS site. The UFSAR changes will reflect 
use of the Electric Power Research Institute (EPRI) Topical Report 
``Tornado Missile Risk

[[Page 26101]]

Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted 
in the NRC Safety Evaluation Report on this topic dated October 26, 
1983, the current licensing criteria governing tornado missile 
protection are contained in NUREG-0800, Sections 3.5.1.4 and 3.5.2. 
These criteria generally specify that safety-related systems, 
structures and components be provided positive tornado missile 
protection (barriers) from the maximum credible tornado threat. 
However, NUREG-0800 includes acceptance criteria permitting 
relaxation of the above deterministic guidance, if it can be 
demonstrated that the probability of damage to unprotected essential 
safety-related features is sufficiently small.
    As permitted in NUREG-0800 sections, the combined probability 
will be maintained below an allowable level, i.e., an acceptance 
criterion threshold, which reflects an extremely low probability of 
occurrence. The approach assumes that if the sum of the individual 
probabilities calculated for tornado missiles striking and damaging 
portions of important systems, structures or components is greater 
than or equal to 1 x 10-6 per year per unit, then 
installation of unique missile barriers would be needed to lower the 
total cumulative probability below the acceptance criterion of 1 x 
10-6 per year per unit.
    With respect to the probability of occurrence or the 
consequences of an accident previously evaluated in the UFSAR, the 
possibility of a tornado reaching the site and causing damage to 
plant structures, systems and components is considered in the MNS 
UFSAR.
    The change being proposed does not affect the probability that 
the natural phenomenon (a tornado) will reach the plant, but from a 
licensing basis perspective, the change does affect the probability 
that missiles generated by the winds of the tornado might strike and 
damage certain plant structures, systems and components. There are a 
limited number of safety-related components that could theoretically 
be struck and damaged by tornadogenerated missiles. The probability 
of tornado-generated missile strikes on important to safety 
structures, systems and components is what was analyzed using the 
probabilistic methods discussed above. The combined probability of 
damage will be maintained below an extremely low acceptance 
criterion to ensure overall plant safety. The proposed change is not 
considered to constitute a significant increase in the probability 
of occurrence or the consequences of an accident, due to the 
extremely low probability of damage due to tornado-generated 
missiles and thus an extremely low probability of a radiological 
release.
    The results of the analysis documented in this [license 
amendment request (LAR)] are below the acceptance criterion of 1 x 
10-6 per year per unit. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the MNS UFSAR incorporate use of a NRC 
approved probabilistic methodology to assess the need for additional 
positive (physical) tornado missile protection for specific 
features. This will not change the design function or operation of 
any structure, system or component. This proposed change does not 
involve any plant modifications. There are no new credible failure 
mechanisms, malfunctions or accident initiators not considered in 
the design and licensing bases for MNS. The proposed change involves 
an already established tornado design basis event and the tornado 
event is explicitly considered in the MNS UFSAR.
    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 the margin of safety?
    Response: No.
    The existing licensing basis for MNS for protecting safety-
related, safe shutdown equipment from tornado generated missiles is 
to provide positive missile barriers for all safety-related 
structures, systems and components. The proposed change recognizes 
that there is an extremely low probability, below an established 
acceptance limit, that a limited subset of the safety-related, safe 
shutdown structures, systems and components could be struck and 
consequently damaged. The change from requiring protection of all 
safety-related, safety shutdown structures, systems and components 
from tornadogenerated missiles, to only a subset of equipment, is 
not considered to constitute a significant decrease in the margin of 
safety due to that extremely low probability of occurrence of 
tornado-generated missile strikes and consequential damage.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: April 5, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18099A130.
    Description of amendment request: The proposed amendment would 
revise the licensing basis, by the addition of a license condition, to 
allow for the implementation of the provisions of 10 CFR 50.69, ``Risk-
informed categorization and treatment of structures, systems, and 
components [SSCs] for nuclear power reactors.''
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The process used to evaluate SSCs 
for changes to NRC special treatment requirements and the use of 
alternative requirements ensures the ability of the SSCs to perform 
their design function. The potential change to special treatment 
requirements does not change the design and operation of the SSCs. 
As a result, the proposed change does not significantly affect any 
initiators to accidents previously evaluated or the ability to 
mitigate any accidents previously evaluated. The consequences of the 
accidents previously evaluated are not affected because the 
mitigation functions performed by the SSCs assumed in the safety 
analysis are not being modified. The SSCs required to safely shut 
down the reactor and maintain it in a safe shutdown condition 
following an accident will continue to perform their design 
functions.
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not change 
the functional requirements, configuration, or method of operation 
of any SSC. Under the proposed change, no additional plant equipment 
will be installed.
    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 will permit the use of a risk-informed 
categorization process to

[[Page 26102]]

modify the scope of SSCs subject to NRC special treatment 
requirements and to implement alternative treatments per the 
regulations. The proposed change does not affect any Safety Limits 
or operating parameters used to establish the safety margin. The 
safety margins included in analyses of accidents are not affected by 
the proposed change. The regulation requires that there be no 
significant effect on plant risk due to any change to the special 
treatment requirements for SSCs and that the SSCs continue to be 
capable of performing their design basis functions, as well as to 
perform any beyond design basis functions consistent with the 
categorization process and results.
    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.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 12, 2018, as supplemented by 
letter dated April 26, 2018. Publicly-available versions are in ADAMS 
under Accession Nos. ML18071A319 and ML18117A493, respectively.
    Description of amendment request: The amendment would revise the 
Arkansas Nuclear One, Unit No. 1 Technical Specifications (TSs) by 
relocating specific surveillance frequencies to a licensee-controlled 
program with the adoption of Technical Specification Task Force (TSTF)-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF [Risk-informed TSTF] Initiative 5b.'' Additionally, the 
change would add a new program, the Surveillance Frequency Control 
Program, to TS Section 5.5, ``Programs and Manuals.''
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements (SRs) to licensee control under a 
new Surveillance Frequency Control Program [SFCP]. Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the technical specifications (TSs) for which the 
surveillance frequencies are relocated are still required to be 
operable, meet the acceptance criteria for the SRs, and be capable 
of performing any mitigation function assumed in the accident 
analysis. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Entergy 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 
1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology 
provides reasonable acceptance guidelines and methods for evaluating 
the risk increase of proposed changes to surveillance frequencies 
consistent with Regulatory Guide 1.177.
    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: Anna Vinson Jones, Senior Counsel, Entergy 
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L-ENT-WDC, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: February 6, 2018, as supplemented by 
letter dated March 26, 2018. Publicly-available versions are in ADAMS 
under Accession Nos. ML18038B354, and ML18085A816, respectively.
    Description of amendment request: The amendment would revise the 
Arkansas Nuclear One, Unit No. 2 Technical Specifications (TSs) by 
relocating specific surveillance frequencies to a licensee-controlled 
program with the adoption of Technical Specifications Task Force 
(TSTF)-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.'' The amendment 
would also add a new program, the Surveillance Frequency Control 
Program, to TS Section 6.0, ``Administrative Controls.''
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic Surveillance Requirements (SRs) to licensee control under a 
new Surveillance Frequency Control Program (SFCP). Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the TSs for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the SRs, and be capable of performing any mitigation 
function assumed in the accident analysis. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 26103]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Entergy 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 
1, in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology 
provides reasonable acceptance guidelines and methods for evaluating 
the risk increase of proposed changes to surveillance frequencies 
consistent with Regulatory Guide 1.177.
    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: Anna Vinson Jones, Senior Counsel, Entergy 
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L-ENT-WDC, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative 
Energy, A Mississippi Electric Cooperative; and Entergy Mississippi, 
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit No. 1, 
Claiborne County, Mississippi

    Date of amendment request: April 10, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18100B304.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to adopt Technical 
Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, 
``Reactor Pressure Vessel Water Inventory Control.'' The proposed 
change would replace existing TS requirements related to ``operations 
with a potential for draining the reactor vessel'' (OPDRVs) with new 
requirements on reactor pressure vessel (RPV) water inventory control 
(WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires 
reactor vessel water level to be greater than the top of active 
irradiated fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed change reduces the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed change reduces the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed change reduces or eliminates some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed change will not alter the design 
function of the equipment involved. Under the proposed change, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements.
    The event of concern under the current requirements and the 
proposed change is an unexpected draining event. The proposed change 
does not create new failure mechanisms, malfunctions, or accident 
initiators that would cause a draining event or a new or different 
kind of accident not previously evaluated or included in the design 
and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.1.3. New requirements 
are added to

[[Page 26104]]

determine the limiting time in which the RPV water inventory could 
drain to the top of the fuel in the reactor vessel should an 
unexpected draining event occur. Plant configurations that could 
result in lowering the RPV water level to the TAF within one hour 
are now prohibited. New escalating compensatory measures based on 
the limiting drain time replace the current controls. The proposed 
TS establish a safety margin by providing defense-in-depth to ensure 
that the Safety Limit is protected and to protect the public health 
and safety. While some less restrictive requirements are proposed 
for plant configurations with long calculated drain times, the 
overall effect of the change is to improve plant safety and to add 
safety margin.
    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: Anna Vinson Jones, Senior Counsel/Legal 
Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 
200 East, Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: April 27, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18117A514.
    Description of amendment request: The proposed amendment would 
revise the Emergency Plan to adopt the Nuclear Energy Institute's 
(NEI's) revised Emergency Action Level (EAL) scheme described in NEI 
99-01, Revision 6, ``Development of Emergency Action Levels for Non-
Passive Reactors'' (ADAMS Accession No. ML110240324), which has been 
endorsed by the NRC (ADAMS Accession No. ML12346A463).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the GGNS EALs do not involve any 
physical changes to plant equipment or systems and do not alter the 
assumptions of any accident analyses. The proposed changes do not 
adversely affect accident initiators or precursors and do not alter 
design assumptions, plant configuration, or the manner in which the 
plant is operated and maintained. The proposed changes do not 
adversely affect the ability of structures, systems or components 
(SSCs) to perform intended safety functions in mitigating the 
consequences of an initiating event within the assumed acceptance 
limits.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The changes do not challenge the integrity or performance of any 
safety-related systems. No plant equipment is installed or removed, 
and the changes do not alter the design, physical configuration, or 
method of operation of any plant SSC. Because EALs are not accident 
initiators and no physical changes are made to the plant, no new 
causal mechanisms are introduced.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from an accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with 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 impact 
operation of the plant and no accident analyses are affected by the 
proposed changes. The changes do not affect the Technical 
Specifications or the method of operating the plant. Additionally, 
the proposed changes will not relax any criteria used to establish 
safety limits and will not relax any safety system settings. The 
safety analysis acceptance criteria are not affected by these 
changes. The proposed changes will not result in plant operation in 
a configuration outside the design basis. The proposed changes do 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Therefore, the changes do not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anna Vinson Jones, Senior Counsel/Legal 
Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 
200 East, Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois, and Docket 
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: April 2, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18092B081.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.2.3 to require that the axial flux 
difference be maintained within the limits specified in the core 
operating limits report during MODE 1 with reactor thermal power 
greater or equal to 50 percent. An associated change would also be made 
to the NOTE modifying surveillance 3.2.3.1.
    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 amendment requires that the AFD [axial flux 
difference] be maintained within the limits specified in the COLR 
[core operating limits report] at-all-times during MODE 1 when 
reactor power is >=50% RTP [reactor thermal power]. This requirement 
will ensure that all FRD [fuel rod design] performance criteria 
remain satisfied during ANS [American Nuclear Society] Condition II 
events (i.e., Faults of Moderate Frequency); thus, ensuring the 
integrity of the fuel rod cladding. It is noted that maintaining AFD 
within the COLR limits at-all-times when >=50% RTP is the normal 
operating practice as specified in plant procedures.
    The proposed change will have no impact on accident initiators 
or precursors; does not alter accident analysis assumptions; does 
not involve any physical plant modifications that would alter the 
design or configuration of the facility, or the manner in which the 
plant is maintained; and does not impact the probability of operator 
error.
    The proposed amendment will not impact the ability of 
structures, systems, and components (SSCs) from performing their 
intended functions to mitigate the consequences of an accident. All 
accident analysis acceptance criteria will continue to be met as the 
proposed change will not affect

[[Page 26105]]

the source term, containment isolation function, or radiological 
release assumptions for any accident previously evaluated.
    Based on the above discussion, the proposed amendment 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 formalizes the existing operating practice 
of maintaining the AFD within the limits specified in the COLR at-
all-times during MODE 1 when reactor power is >= 50% RTP. This 
change ensures that all FRD performance criteria remain satisfied 
during ANS Condition II events. The ANS Condition II events have all 
been previously evaluated in the Updated Final Safety Analysis 
Report.
    The proposed change does not involve a design change or other 
changes that would impact safety-related SSCs from performing their 
specified safety functions.
    The proposed change does not result in the creation of any new 
accident precursors; does not result in changes to any existing 
accident scenarios; and does not introduce any operational changes 
or mechanisms that would create the possibility of a new or 
different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to maintain the AFD within the limits 
specified in the COLR at-all-times during MODE 1 when reactor power 
is >= 50% RTP ensures that all FRD performance criteria remain 
satisfied during ANS Condition II events; and thus, will maintain 
the existing margin of safety related to FRD performance criteria 
and ensure the integrity of the fuel rod cladding. The AFD limits 
specified in the COLR have been established in accordance with the 
analysis approach described in NRC-approved Westinghouse Topical 
Reports.
    In addition, this change will have no impact on the margin of 
safety associated with other reactor core safety parameters such as 
fuel hot channel factors, core power tilt ratios, loss of coolant 
accident peak cladding temperature and peak local power density.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: David J. Wrona.

FirstEnergy Nuclear Operating Company, Docket No. 50-412, Beaver Valley 
Power Station, Unit No. 2, Beaver County, Pennsylvania

    Date of amendment request: March 28, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18087A293.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.5.2.d, ``Provisions for SG [Steam 
Generator] Tube Inspection,'' and TS 5.5.5.2.f, ``Provisions for SG 
Tube Repair Methods.'' More specifically, TSs 5.5.5.2.d.5 and 
5.5.5.2.f.3 would be simplified and clarified, respectively, without 
changing the intent of the specifications. Specification 5.5.5.2.f.3 
would also be amended by changing the number of fuel cycles that 
Westinghouse Electric Company, LLC leak-limiting Alloy 800 sleeves may 
remain in operation.
    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.
    Proposed amendment of Technical Specification 5.5.5.2.d.5 to 
simplify the description of the required inspection region, and 
Technical Specification 5.5.5.2.f.3 to clarify that this 
specification is only applicable to sleeves installed in the steam 
generator tubesheet and change the number of fuel cycles that an 
Alloy 800 steam generator tubesheet sleeve may remain in service 
from five to eight fuel cycles of operation, does not affect 
structures, systems or components of the plant, plant operations, 
design functions or analyses that verify the capability of 
structures, systems or components to perform a design function. The 
proposed amendment does not increase the likelihood of steam 
generator tube sleeve leakage.
    The proposed amendment of Technical Specification 5.5.5.2.d.5 to 
simplify the description of the required inspection region, makes it 
clear that the steam generator parent tube is to be inspected in the 
areas where the joints will be established prior to installation of 
the sleeve, regardless of the sleeve location. This proposed 
amendment does not change the intent of the specification.
    The proposed amendment of TS 5.5.5.2.f.3 includes two changes. 
The first change would add the words ``installed in the hot-leg or 
cold-leg tubesheet region'' after the words ``An Alloy 800 sleeve'' 
to make it clear that the specification only applies to Alloy 800 
tube sleeves installed in the steam generator tubesheet. The design 
of Alloy 800 sleeves installed in steam generator tube locations 
other than the tubesheet does not include a nickel band. For these 
sleeves, nondestructive examination methods have been demonstrated 
to be effective and limits on sleeve operating life are not 
necessary. This proposed amendment does not change the intent of the 
specification.
    The second change to TS 5.5.5.2.f.3, increases the number of 
fuel cycles Alloy 800 tube sleeves installed in the tubesheet may 
remain in service. The leak-limiting Alloy 800 sleeves are designed 
using the applicable American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code and, therefore, meet the design 
objectives of the original steam generator tubing. The applied 
stresses and fatigue usage for the sleeves are bounded by the limits 
established in the ASME Code. Mechanical testing has shown that the 
structural strength of sleeves under normal, upset, emergency, and 
faulted conditions provides margin to the acceptance limits. These 
acceptance limits bound the most limiting (three times normal 
operating pressure differential) burst margin of NRC Regulatory 
Guide 1.121, ``Bases for Plugging Degraded PWR Steam Generator 
Tubes.''
    The leak-limiting Alloy 800 sleeve depth-based structural limit 
is determined using NRC guidance and the pressure stress equation of 
ASME Code, Section III with margin added to account for the 
configuration of long axial cracks. Calculations show that a depth-
based limit of 45 percent through-wall degradation is acceptable. 
However, Technical Specifications 5.5.5.2.c.2 and 5.5.5.2.c.3 
provide additional margin by requiring an Alloy 800 sleeved tube to 
be plugged on detection of any flaw in the sleeve or in the pressure 
boundary portion of the original tube wall in the sleeve to tube 
joint. Degradation of the original tube adjacent to the nickel band 
of an Alloy 800 sleeve installed in the tubesheet, regardless of 
depth, would not prevent the sleeve from satisfying design 
requirements. Thus, flaw detection capabilities within the original 
tube adjacent to the sleeve nickel band are a defense-in-depth 
measure, and are not necessary in order to justify continued 
operation of the sleeved tube.
    Evaluation of repaired steam generator tube testing and analysis 
indicates that there are no detrimental effects on the leak-limiting 
Alloy 800 sleeve or sleeved tube assembly from reactor coolant 
system flow, primary or secondary coolant chemistries, thermal 
conditions or transients, or pressure conditions that may be 
experienced at Beaver Valley Power Station, Unit No. 2. Westinghouse 
is not aware of, and has no knowledge of any reports of parent-tube 
stress corrosion cracking (SCC) in the sleeve roll joint region for 
any Westinghouse sleeve design.
    The proposed increase in the number of fuel cycles Alloy 800 
tube sleeves installed in the tubesheet may remain in service has no 
effect on sleeve operation or capability of the sleeve to perform 
its design function. The mechanical and leakage tests have confirmed

[[Page 26106]]

that degradation of the parent tube adjacent to the nickel band will 
not prevent the sleeve from satisfying its design function.
    Consequences of a hypothetical failure of the leak-limiting 
Alloy 800 sleeve and tube assembly are bounded by the current main 
steam line break and steam generator tube rupture accident analyses 
described in the Beaver Valley Power Station, Unit No. 2 Updated 
Final Safety Analysis Report. The total number of plugged steam 
generator tubes (including equivalency associated with installed 
sleeves) is required to be consistent with accident analysis 
assumptions. The sleeve and tube assembly leakage during plant 
operation is required to be within the allowable Technical 
Specification leakage limits and accident analysis assumptions.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Proposed amendment of Technical Specification 5.5.5.2.d.5 to 
simplify the description of the required inspection region, and 
Technical Specification 5.5.5.2.f.3 to clarify that this 
specification is only applicable to sleeves installed in the steam 
generator tubesheet do not change the intent of these 
specifications, and do not affect the design function or operation 
of the tube sleeves. The proposed amendment of Technical 
Specification 5.5.5.2.f.3 to change the number of fuel cycles that 
an Alloy 800 steam generator tubesheet sleeve may remain in service 
from five to eight fuel cycles of operation, does not affect the 
design function or operation of the tube sleeves. Since these 
changes do not create any credible new failure mechanisms, 
malfunctions, or accident initiators not considered in the design or 
licensing bases, the changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The leak-limiting Alloy 800 sleeves are designed using the 
applicable ASME Code, and therefore meet the objectives of the 
original steam generator tubing. As a result, the functions of the 
steam generator will not be significantly affected by the 
installation of the proposed sleeve. Therefore, the only credible 
failure modes for the sleeve and tube are to leak or rupture, which 
has already been evaluated. The continued integrity of the installed 
sleeve and tube assembly is periodically verified as required by the 
Technical Specifications, and a sleeved tube will be plugged on 
detection of a flaw in the sleeve or in the pressure boundary 
portion of the original tube wall in the sleeve to tube joint.
    The proposed amendment to Technical Specification 5.5.5.2.f.3 
increases the number of fuel cycles Alloy 800 tube sleeves installed 
in the tubesheet may remain in service to eight fuel cycles of 
operation. Implementation of this proposed amendment has no 
significant effect on either the configuration of the plant, the 
manner in which it is operated, or ability of the sleeve to perform 
its design function.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Proposed amendment of Technical Specification 5.5.5.2.d.5 to 
simplify the description of the required inspection region, and 
Technical Specification 5.5.5.2.f.3 to clarify that this 
specification is only applicable to sleeves installed in the steam 
generator tubesheet, do not change the intent of these requirements 
or reduce the margin of safety. The proposed amendment to Technical 
Specification 5.5.5.2.f.3 to change the number of fuel cycles that 
an Alloy 800 steam generator tubesheet sleeve may remain in service 
from five to eight fuel cycles of operation, does not affect a 
design basis or safety limit (that is, the controlling numerical 
value for a parameter established in the Updated Final Safety 
Analysis Report or the license) or reduce the margin of safety.
    The proposed amendment to Technical Specification 5.5.5.2.f.3 
increases the number of fuel cycles Alloy 800 tube sleeves installed 
in the tubesheet may remain in service to eight fuel cycles of 
operation. Implementation of this proposed amendment would not 
affect a design basis or safety limit or reduce the margin of 
safety. The repair of degraded steam generator tubes with leak-
limiting Alloy 800 sleeves restores the structural integrity of the 
degraded tube under normal operating and postulated accident 
conditions. Minimum reactor coolant system flow rate from the 
cumulative effect of repaired (sleeved) and plugged tubes will be 
greater than the flow rate limit established in the Technical 
Specification limiting condition for operation 3.4.1. The design 
safety factors utilized for the sleeves are consistent with the 
safety factors in the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code used in the original steam generator 
design. Tubes with sleeves are subject to the same safety factors as 
the original tubes, which are described in the performance criteria 
for steam generator tube integrity in the existing Technical 
Specifications. The sleeve and portions of the installed sleeve and 
tube assembly that represent the reactor coolant pressure boundary 
will be monitored, and a sleeved tube will be plugged if a flaw is 
detected in the sleeve or in the pressure boundary portion of the 
original tube wall in the leak-limiting sleeve and tube assembly. 
Use of the previously-identified design criteria and design 
verification testing ensures that the margin of safety is not 
significantly different from the original steam generator tubes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: James Danna.

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

    Date of amendment request: March 28, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18087A095.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3/4.8.1, ``AC [Alternating Current] 
Sources--Operating''; specifically, ACTION b concerning one inoperable 
emergency diesel generator (EDG). The proposed change would remove the 
Salem Nuclear Generating Station, Unit No. 3 (Salem Unit 3), gas 
turbine generator and replace it with portable diesel generators.
    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 removes the requirement for the Salem Unit 3 
gas turbine generator (GTG) and replaces it with the supplemental 
power source during the existing extended allowable outage time for 
the A or B EDG. The emergency diesel generators are safety related 
components which provide backup electrical power supply to the 
onsite Safeguards Distribution System. The emergency diesel 
generators are not accident initiators; the EDGs are designed to 
mitigate the consequences of previously evaluated accidents 
including a loss of offsite power. (During normal operation, the 
proposed portable diesel generators will not be connected to the 
plant.)
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. The proposed change is consistent with safety analysis 
assumptions and resultant consequences.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 26107]]

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 removes the requirement for the Salem Unit 3 
gas turbine generator (GTG) and replaces it with the supplemental 
power source during the existing extended allowable outage time for 
the A or B EDG. The proposed change does not alter or involve any 
design basis accident initiators. Equipment will be operated in the 
same configuration and manner that is currently allowed and designed 
for.
    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 does not alter the permanent plant design, 
including instrument set points, nor does it change the assumptions 
contained in the safety analyses. The proposed change does not 
impact the redundancy or availability requirements of offsite power 
supplies or change the ability of the plant to cope with station 
blackout [(SBO)] events.
    The EDGs continue to meet their design requirements; there is no 
reduction in capability or change in design configuration. The EDG 
response to LOOP [loss of offsite power], LOCA [loss-of-coolant 
accident], SBO, or fire is not changed by this proposed amendment; 
there is no change to the EDG operating parameters. The remaining 
operable emergency diesel generators are adequate to supply 
electrical power to the onsite Safeguards Distribution System. The 
proposed change does not alter a design basis or safety limit; 
therefore it does not significantly reduce the margin of safety. The 
EDGs will continue to operate per the existing design and regulatory 
requirements.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Unit Nos. 1 and 2 (SQN), Hamilton County, Tennessee

    Date of amendment request: March 9, 2018, as supplemented by letter 
dated April 11, 2018. Publicly-available versions are in ADAMS under 
Accession Nos. ML18071A349 and ML18102B430, respectively.
    Description of amendment request: The amendments would make changes 
to the SQN Essential Raw Cooling Water (ERCW) Motor Control Centers 
(MCCs) and revise the Updated Final Safety Analysis Report (UFSAR).
    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 change does not alter the safety function of any 
structure, system, or component, does not modify the manner in which 
the plant is operated, and does not alter equipment out-of-service 
time. In addition, this request does not degrade the ability of the 
ERCW to perform its intended safety function. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve any physical changes to 
plant safety related structure, system or component or alter the 
modes of plant operation in a manner that is outside the bounds of 
the system design analyses. The proposed change to complete the 
design change for the removal of mechanical interlock device from 
the feeder breakers and tie breakers for the ERCW MCCs and to revise 
the ERCW System Description in Section 9.2.2.2 of the SQN UFSAR to 
describe the normal and alternate power sources for the ERCW system 
does not create the possibility for an accident or malfunction of a 
different type than any evaluated previously in SQN's UFSAR. The 
proposal does not alter the way any safety related structure, system 
or component functions and does not modify the manner in which the 
plant is operated. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to remove the mechanical interlock device 
from the feeder breakers and tie breakers for ERCW MCCs 1B-B and 2B-
B and to revise the ERCW System Description in Section 9.2.2.2 of 
the SQN UFSAR to describe the normal and alternate power sources for 
the ERCW system does not reduce the margin of safety because ERCW 
will continue to perform its safety function. The design features 
provided by the mechanical interlock device are not described in the 
SQN UFSAR, are not credited in the SQN accident analysis and do not 
provide any additional safety margin. The results of accident 
analyses remain unchanged by this request. 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.

Vistra Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche 
Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: March 29, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18102A516.
    Description of amendment request: The amendments would revise 
Technical Specification 3.3.2, ``Engineered Safety Feature Actuation 
System (ESFAS) Instrumentation,'' to change the applicability of when 
the automatic auxiliary feedwater actuation due to the trip of all main 
feedwater pumps is required to be operable at Comanche Peak Nuclear 
Power Plant, Unit Nos. 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The design basis events which impose auxiliary feedwater safety 
function requirements are loss of all AC [alternating current] power 
to plant auxiliaries, loss of normal feedwater, steam generator 
fault in either the feedwater or steam lines, and small break loss 
of coolant accidents. These design basis event evaluations assume 
actuation of auxiliary feedwater due to station blackout, low-low 
steam generator level or a safety injection signal. The anticipatory 
auxiliary feedwater automatic start signals from the main feedwater 
pumps are not credited in any design basis accidents and are, 
therefore, not part of the primary success path for postulated 
accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion 
3. Modifying MODE 2 Applicability for this function will not impact 
any previously evaluated design basis accidents.

[[Page 26108]]

    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This technical specification change allows for an operational 
allowance during MODE 2 while placing main feedwater pumps in 
service. This change involves an anticipatory auxiliary feedwater 
automatic start function that is not credited in the accident 
analysis. Since this change only affects the conditions at which 
this automatic start function needs to be operable and does not 
affect the function that actuates auxiliary feedwater due to loss of 
offsite power, low-low steam generator level or a safety injection 
signal, it will not be an initiator to a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This technical [s]pecification change involves the automatic 
start of the auxiliary feedwater pumps due to trip of both main 
feedwater pumps, which is not an assumed start signal for design 
basis events. This change does not modify any values or limits 
involved in a safety related function or accident analysis.
    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: Timothy P. Matthews, Esq., Morgan, Lewis, 
and Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
    NRC Branch Chief: Robert J. Pascarelli.

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.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: June 8, 2017.
    Brief description of amendment: The amendment revised technical 
specifications (TSs) to reflect previously approved changes made as 
part of the alternative source term initiative. The amendment revised 
the surveillance requirements for the control room emergency 
recirculation and annulus exhaust gas treatment systems, which are 
consistent with Technical Specification Task Force (TSTF) Traveler 
TSTF-522, ``Revise Ventilation System Surveillance Requirement to 
Operate for 10 Hours per Month.'' The amendment also deleted two TS 
sections related to the fuel handling building and fuel handling 
building ventilation exhaust system and increased the allowable 
secondary containment leakage. Lastly, the amendment revised the TS 
Table of Contents to reflect administrative changes to the titles of TS 
sections.
    Date of issuance: May 16, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 180. A publicly-available version is in ADAMS under 
Accession No. ML18110A133; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: The amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: August 1, 2017 (82 FR 
35841). The supplemental letter dated January 30, 2018, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 16, 2018.
    No significant hazards consideration comments received: No.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: March 24, 2017.
    Brief description of amendment: The amendment revised the DAEC 
Technical Specification (TS) Table 3.3.2.1-1, ``Control Rod Block 
Instrumentation,'' by relocating certain cycle-specific Minimum 
Critical Power Ratio values to the DAEC Core Operating Limits Report. 
The amendment also added a requirement to DAEC TS 5.6.5, ``Core 
Operating Limits Report.''
    Date of issuance: March 7, 2018.
    Effective date: As of the date of its issuance and shall be 
implemented by September 27, 2018. (Note: This Notice of Issuance 
corrects the ``Effective date'' of Amendment No. 303 originally noticed 
in the Federal Register on March 27, 2018 (83 FR 13153).
    Amendment No.: 303. A publicly-available version is in ADAMS under 
Accession No. ML18011A059; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment. Amendment 
No. 303 was corrected by letter dated May 7, 2018 (ADAMS Accession No. 
ML18081A074).
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23627).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2018.

[[Page 26109]]

    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 29th day of May, 2018.

    For the Nuclear Regulatory Commission.
Gregory F. Suber,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2018-11843 Filed 6-4-18; 8:45 am]
 BILLING CODE 7590-01-P