83 FR 26098 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 83, Issue 108 (June 5, 2018)
Federal Register Volume 83, Issue 108 (June 5, 2018)
| Page Range | 26098-26109 | |
| FR Document | 2018-11843 |
[Federal Register Volume 83, Number 108 (Tuesday, June 5, 2018)] [Notices] [Pages 26098-26109] From the Federal Register Online [www.thefederalregister.org] [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
![26098 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices
entities, persons, products, and questions about NRC dockets to Jennifer and subject in your comment
offerings. Borges; telephone: 301–287–9127; submission.
7. United States Postal Service, Office email: Jennifer.Borges@nrc.gov. For The NRC cautions you not to include
of the Regional Director, Atlanta, technical questions, contact the identifying or contact information that
Georgia (DAA–0028–2018–0001, 6 individual listed in the FOR FURTHER you do not want to be publicly
items, 6 temporary items). Dedication INFORMATION CONTACT section of this disclosed in your comment submission.
files and site selection files of document. The NRC will post all comment
individual post offices in Florida, • Mail comments to: May Ma, Office submissions at http://
Georgia, North Carolina, South Carolina, of Administration, Mail Stop: TWFN–7– www.regulations.gov as well as enter the
and Puerto Rico. Includes personnel A60M, U.S. Nuclear Regulatory comment submissions into ADAMS.
records and routine organization data. Commission, Washington, DC 20555– The NRC does not routinely edit
0001. comment submissions to remove
Laurence Brewer, For additional direction on obtaining identifying or contact information.
Chief Records Officer for the U.S. information and submitting comments, If you are requesting or aggregating
Government. see ‘‘Obtaining Information and comments from other persons for
[FR Doc. 2018–11987 Filed 6–4–18; 8:45 am] Submitting Comments’’ in the submission to the NRC, then you should
BILLING CODE 7515–01–P SUPPLEMENTARY INFORMATION section of inform those persons not to include
this document. identifying or contact information that
FOR FURTHER INFORMATION CONTACT: they do not want to be publicly
NUCLEAR REGULATORY Janet Burkhardt, Office of Nuclear disclosed in their comment submission.
COMMISSION Reactor Regulation, U.S. Nuclear Your request should state that the NRC
Regulatory Commission, Washington DC does not routinely edit comment
[NRC–2018–0105] submissions to remove such information
20555–0001; telephone: 301–415–1384,
email: Janet.Burkhardt@nrc.gov. before making the comment
Biweekly Notice; Applications and submissions available to the public or
Amendments to Facility Operating SUPPLEMENTARY INFORMATION:
entering the comment into ADAMS.
Licenses and Combined Licenses I. Obtaining Information and
Involving No Significant Hazards II. Notice of Consideration of Issuance
Submitting Comments of Amendments to Facility Operating
Considerations
A. Obtaining Information Licenses and Combined Licenses and
AGENCY: Nuclear Regulatory Proposed No Significant Hazards
Commission. Please refer to Docket ID NRC–2018–
Consideration Determination
0105, facility name, unit number(s),
ACTION: Biweekly notice. The Commission has made a
plant docket number, application date,
SUMMARY: Pursuant to Section 189a.(2) and subject when contacting the NRC proposed determination that the
of the Atomic Energy Act of 1954, as about the availability of information for following amendment requests involve
amended (the Act), the U.S. Nuclear this action. You may obtain publicly- no significant hazards consideration.
Regulatory Commission (NRC) is available information related to this Under the Commission’s regulations in
publishing this regular biweekly notice. action by any of the following methods: § 50.92 of title 10 of the Code of Federal
The Act requires the Commission to • Federal Rulemaking website: Go to Regulations (10 CFR), this means that
publish notice of any amendments http://www.regulations.gov and search operation of the facility in accordance
issued, or proposed to be issued, and for Docket ID NRC–2018–0105. with the proposed amendment would
grants the Commission the authority to • NRC’s Agencywide Documents not (1) involve a significant increase in
issue and make immediately effective Access and Management System the probability or consequences of an
any amendment to an operating license (ADAMS): You may obtain publicly- accident previously evaluated; or (2)
or combined license, as applicable, available documents online in the create the possibility of a new or
upon a determination by the ADAMS Public Documents collection at different kind of accident from any
Commission that such amendment http://www.nrc.gov/reading-rm/ accident previously evaluated; or (3)
involves no significant hazards adams.html. To begin the search, select involve a significant reduction in a
consideration, notwithstanding the ‘‘ADAMS Public Documents’’ and then margin of safety. The basis for this
pendency before the Commission of a select ‘‘Begin Web-based ADAMS proposed determination for each
request for a hearing from any person. Search.’’ For problems with ADAMS, amendment request is shown below.
please contact the NRC’s Public The Commission is seeking public
This biweekly notice includes all
Document Room (PDR) reference staff at comments on this proposed
notices of amendments issued, or
1–800–397–4209, 301–415–4737, or by determination. Any comments received
proposed to be issued, from May 8,
email to pdr.resource@nrc.gov. The within 30 days after the date of
2018, to May 21, 2018. The last
ADAMS accession number for each publication of this notice will be
biweekly notice was published on May
document referenced (if it is available in considered in making any final
22, 2018.
ADAMS) is provided the first time that determination.
DATES: Comments must be filed by July Normally, the Commission will not
it is mentioned in this document.
5, 2018. A request for a hearing must be • NRC’s PDR: You may examine and issue the amendment until the
filed by August 6, 2018. purchase copies of public documents at expiration of 60 days after the date of
ADDRESSES: You may submit comments the NRC’s PDR, Room O1–F21, One publication of this notice. The
amozie on DSK3GDR082PROD with NOTICES1
by any of the following methods (unless White Flint North, 11555 Rockville Commission may issue the license
this document describes a different Pike, Rockville, Maryland 20852. amendment before expiration of the 60-
method for submitting comments on a day period provided that its final
specific subject): B. Submitting Comments determination is that the amendment
• Federal Rulemaking website: Go to Please include Docket ID NRC–2018– involves no significant hazards
http://www.regulations.gov and search 0105, facility name, unit number(s), consideration. In addition, the
for Docket ID NRC–2018–0105. Address plant docket number, application date, Commission may issue the amendment
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![Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 26101
Evaluation Methodology’’ (EPRI NP–2005), Response: No. Description of amendment request:
Volumes I and II. As noted in the NRC Safety The proposed changes to the MNS UFSAR The proposed amendment would revise
Evaluation Report on this topic dated incorporate use of a NRC approved the licensing basis, by the addition of a
October 26, 1983, the current licensing probabilistic methodology to assess the need license condition, to allow for the
criteria governing tornado missile protection for additional positive (physical) tornado
are contained in NUREG–0800, Sections missile protection for specific features. This
implementation of the provisions of 10
3.5.1.4 and 3.5.2. These criteria generally will not change the design function or CFR 50.69, ‘‘Risk-informed
specify that safety-related systems, structures operation of any structure, system or categorization and treatment of
and components be provided positive component. This proposed change does not structures, systems, and components
tornado missile protection (barriers) from the involve any plant modifications. There are no [SSCs] for nuclear power reactors.’’
maximum credible tornado threat. However, new credible failure mechanisms, Basis for proposed no significant
NUREG–0800 includes acceptance criteria malfunctions or accident initiators not hazards consideration determination:
permitting relaxation of the above considered in the design and licensing bases As required by 10 CFR 50.91(a), the
deterministic guidance, if it can be for MNS. The proposed change involves an licensee has provided its analysis of the
demonstrated that the probability of damage already established tornado design basis
to unprotected essential safety-related issue of no significant hazards
event and the tornado event is explicitly
features is sufficiently small. considered in the MNS UFSAR.
consideration, which is presented
As permitted in NUREG–0800 sections, the Therefore, the proposed change does not below:
combined probability will be maintained create the possibility of a new or different 1. Does the proposed change involve a
below an allowable level, i.e., an acceptance kind of accident from any accident significant increase in the probability or
criterion threshold, which reflects an previously evaluated. consequences of an accident previously
extremely low probability of occurrence. The 3. Does the proposed amendment involve evaluated?
approach assumes that if the sum of the a significant reduction in the margin of Response: No.
individual probabilities calculated for safety? The proposed change will permit the use
tornado missiles striking and damaging Response: No. of a risk-informed categorization process to
portions of important systems, structures or The existing licensing basis for MNS for modify the scope of SSCs subject to NRC
components is greater than or equal to 1 × protecting safety-related, safe shutdown special treatment requirements and to
10¥6 per year per unit, then installation of equipment from tornado generated missiles is implement alternative treatments per the
unique missile barriers would be needed to to provide positive missile barriers for all regulations. The process used to evaluate
lower the total cumulative probability below safety-related structures, systems and SSCs for changes to NRC special treatment
the acceptance criterion of 1 × 10¥6 per year components. The proposed change requirements and the use of alternative
per unit. recognizes that there is an extremely low requirements ensures the ability of the SSCs
With respect to the probability of probability, below an established acceptance to perform their design function. The
occurrence or the consequences of an limit, that a limited subset of the safety- potential change to special treatment
accident previously evaluated in the UFSAR, related, safe shutdown structures, systems requirements does not change the design and
the possibility of a tornado reaching the site and components could be struck and operation of the SSCs. As a result, the
and causing damage to plant structures, consequently damaged. The change from proposed change does not significantly affect
systems and components is considered in the requiring protection of all safety-related, any initiators to accidents previously
MNS UFSAR. safety shutdown structures, systems and evaluated or the ability to mitigate any
The change being proposed does not affect components from tornadogenerated missiles, accidents previously evaluated. The
the probability that the natural phenomenon to only a subset of equipment, is not consequences of the accidents previously
(a tornado) will reach the plant, but from a considered to constitute a significant evaluated are not affected because the
licensing basis perspective, the change does mitigation functions performed by the SSCs
decrease in the margin of safety due to that
affect the probability that missiles generated assumed in the safety analysis are not being
extremely low probability of occurrence of
by the winds of the tornado might strike and modified. The SSCs required to safely shut
tornado-generated missile strikes and
damage certain plant structures, systems and down the reactor and maintain it in a safe
consequential damage.
components. There are a limited number of shutdown condition following an accident
Therefore, the proposed change does not will continue to perform their design
safety-related components that could involve a significant reduction in a margin of
theoretically be struck and damaged by functions.
safety. Therefore, the proposed change does not
tornadogenerated missiles. The probability of
tornado-generated missile strikes on The NRC staff has reviewed the involve a significant increase in the
important to safety structures, systems and licensee’s analysis and, based on this probability or consequences of an accident
review, it appears that the three previously evaluated.
components is what was analyzed using the
2. Does the proposed change create the
probabilistic methods discussed above. The standards of 10 CFR 50.92(c) are possibility of a new or different kind of
combined probability of damage will be satisfied. Therefore, the NRC staff accident from any accident previously
maintained below an extremely low proposes to determine that the evaluated?
acceptance criterion to ensure overall plant amendment request involves no Response: No.
safety. The proposed change is not The proposed change will permit the use
significant hazards consideration.
considered to constitute a significant increase of a risk-informed categorization process to
in the probability of occurrence or the
Attorney for licensee: Kate B. Nolan,
Deputy General Counsel, Duke Energy modify the scope of SSCs subject to NRC
consequences of an accident, due to the special treatment requirements and to
extremely low probability of damage due to Carolinas, LLC, 550 South Tryon
implement alternative treatments per the
tornado-generated missiles and thus an Street—DEC45A, Charlotte, NC 28202– regulations. The proposed change does not
extremely low probability of a radiological 1802. change the functional requirements,
release. NRC Branch Chief: Michael T. configuration, or method of operation of any
The results of the analysis documented in Markley. SSC. Under the proposed change, no
this [license amendment request (LAR)] are additional plant equipment will be installed.
below the acceptance criterion of 1 × 10¥6 Duke Energy Progress, LLC, Docket No. Therefore, the proposed change does not
amozie on DSK3GDR082PROD with NOTICES1
per year per unit. Therefore, the proposed 50–261, H. B. Robinson Steam Electric create the possibility of a new or different
change does not involve a significant Plant, Unit No. 2, Darlington County, kind of accident from any accident
increase in the probability or consequences South Carolina previously evaluated.
of an accident previously evaluated. 3. Does the proposed change involve a
2. Does the proposed amendment create Date of amendment request: April 5, significant reduction in a margin of safety?
the possibility of a new or different kind of 2018. A publicly-available version is in Response: No.
accident from any accident previously ADAMS under Accession No. The proposed change will permit the use
evaluated? ML18099A130. of a risk-informed categorization process to
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![26102 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices
modify the scope of SSCs subject to NRC The proposed change relocates the standards of 10 CFR 50.92(c) are
special treatment requirements and to specified frequencies for periodic satisfied. Therefore, the NRC staff
implement alternative treatments per the surveillance requirements (SRs) to licensee proposes to determine that the
regulations. The proposed change does not control under a new Surveillance Frequency
Control Program [SFCP]. Surveillance
amendment request involves no
affect any Safety Limits or operating
parameters used to establish the safety frequencies are not an initiator to any significant hazards consideration.
margin. The safety margins included in accident previously evaluated. As a result, Attorney for licensee: Anna Vinson
analyses of accidents are not affected by the the probability of any accident previously Jones, Senior Counsel, Entergy Services,
proposed change. The regulation requires evaluated is not significantly increased. The Inc., 101 Constitution Avenue NW,
that there be no significant effect on plant systems and components required by the Suite 200 East, L–ENT–WDC,
risk due to any change to the special technical specifications (TSs) for which the Washington, DC 20001.
treatment requirements for SSCs and that the surveillance frequencies are relocated are NRC Branch Chief: Robert J.
SSCs continue to be capable of performing still required to be operable, meet the Pascarelli.
their design basis functions, as well as to acceptance criteria for the SRs, and be
perform any beyond design basis functions capable of performing any mitigation Entergy Operations, Inc., Docket No. 50–
consistent with the categorization process function assumed in the accident analysis. 368, Arkansas Nuclear One, Unit No. 2,
and results. As a result, the consequences of any accident Pope County, Arkansas
Therefore, the proposed change does not previously evaluated are not significantly
involve a significant reduction in a margin of increased. Date of amendment request: February
safety. Therefore, the proposed change does not 6, 2018, as supplemented by letter dated
The NRC staff has reviewed the involve a significant increase in the March 26, 2018. Publicly-available
licensee’s analysis and, based on this probability or consequences of an accident versions are in ADAMS under
previously evaluated. Accession Nos. ML18038B354, and
review, it appears that the three
2. Does the proposed change create the ML18085A816, respectively.
standards of 10 CFR 50.92(c) are possibility of a new or different kind of
satisfied. Therefore, the NRC staff Description of amendment request:
accident from any previously evaluated?
proposes to determine that the Response: No. The amendment would revise the
amendment request involves no No new or different accidents result from Arkansas Nuclear One, Unit No. 2
significant hazards consideration. utilizing the proposed change. The changes Technical Specifications (TSs) by
Attorney for licensee: Kathryn B. do not involve a physical alteration of the relocating specific surveillance
Nolan, Deputy General Counsel, Duke plant (i.e., no new or different type of frequencies to a licensee-controlled
Energy Corporation, 550 South Tryon equipment will be installed) or a change in program with the adoption of Technical
the methods governing normal plant Specifications Task Force (TSTF)-425,
Street, DEC45A, Charlotte NC 28202. operation. In addition, the changes do not
NRC Acting Branch Chief: Brian W. Revision 3, ‘‘Relocate Surveillance
impose any new or different requirements.
Tindell. The changes do not alter assumptions made Frequencies to Licensee Control—
in the safety analysis. The proposed changes RITSTF [Risk-Informed TSTF] Initiative
Entergy Operations, Inc., Docket No. 50– 5b.’’ The amendment would also add a
are consistent with the safety analysis
313, Arkansas Nuclear One, Unit No. 1, new program, the Surveillance
assumptions and current plant operating
Pope County, Arkansas practice. Frequency Control Program, to TS
Date of amendment request: March Therefore, the proposed changes do not Section 6.0, ‘‘Administrative Controls.’’
12, 2018, as supplemented by letter create the possibility of a new or different Basis for proposed no significant
dated April 26, 2018. Publicly-available kind of accident from any accident hazards consideration determination:
previously evaluated.
versions are in ADAMS under As required by 10 CFR 50.91(a), the
3. Does the proposed change involve a
Accession Nos. ML18071A319 and significant reduction in the margin of safety? licensee has provided its analysis of the
ML18117A493, respectively. Response: No. issue of no significant hazards
Description of amendment request: The design, operation, testing methods, consideration, which is presented
The amendment would revise the and acceptance criteria for systems, below:
Arkansas Nuclear One, Unit No. 1 structures, and components (SSCs), specified
1. Does the proposed change involve a
Technical Specifications (TSs) by in applicable codes and standards (or
significant increase in the probability or
relocating specific surveillance alternatives approved for use by the NRC)
consequences of any accident previously
frequencies to a licensee-controlled will continue to be met as described in the
evaluated?
plant licensing basis (including the final
program with the adoption of Technical Response: No.
safety analysis report and bases to TS), since
Specification Task Force (TSTF)-425, The proposed change relocates the
these are not affected by changes to the
Revision 3, ‘‘Relocate Surveillance specified frequencies for periodic
surveillance frequencies. Similarly, there is
Frequencies to Licensee Control— Surveillance Requirements (SRs) to licensee
no impact to safety analysis acceptance
RITSTF [Risk-informed TSTF] Initiative control under a new Surveillance Frequency
criteria as described in the plant licensing
basis. To evaluate a change in the relocated Control Program (SFCP). Surveillance
5b.’’ Additionally, the change would frequencies are not an initiator to any
add a new program, the Surveillance surveillance frequency, Entergy will perform
a probabilistic risk evaluation using the accident previously evaluated. As a result,
Frequency Control Program, to TS the probability of any accident previously
guidance contained in NRC approved NEI
Section 5.5, ‘‘Programs and Manuals.’’ evaluated is not significantly increased. The
[Nuclear Energy Institute] 04–10, Rev. 1 in
Basis for proposed no significant accordance with the TS SFCP. NEI 04–10, systems and components required by the TSs
hazards consideration determination: Rev. 1, methodology provides reasonable for which the surveillance frequencies are
As required by 10 CFR 50.91(a), the acceptance guidelines and methods for relocated are still required to be operable,
licensee has provided its analysis of the evaluating the risk increase of proposed meet the acceptance criteria for the SRs, and
issue of no significant hazards changes to surveillance frequencies be capable of performing any mitigation
amozie on DSK3GDR082PROD with NOTICES1
consideration, which is presented consistent with Regulatory Guide 1.177. function assumed in the accident analysis.
Therefore, the proposed changes do not As a result, the consequences of any accident
below: previously evaluated are not significantly
involve a significant reduction in a margin of
1. Does the proposed change involve a safety. increased.
significant increase in the probability or Therefore, the proposed change does not
consequences of any accident previously The NRC staff has reviewed the involve a significant increase in the
evaluated? licensee’s analysis and, based on this probability or consequences of an accident
Response: No. review, it appears that the three previously evaluated.
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2. Does the proposed change create the ADAMS under Accession No. 5. The change in requirement from two ECCS
possibility of a new or different kind of ML18100B304. subsystems to one ECCS subsystem in Modes
accident from any previously evaluated? Description of amendment request: 4 and 5 does not significantly affect the
Response: No. The proposed amendment would revise consequences of an unexpected draining
No new or different accidents result from event because the proposed Actions ensure
utilizing the proposed change. The changes the Technical Specifications (TSs) to equipment is available within the limiting
do not involve a physical alteration of the adopt Technical Specifications Task drain time that is as capable of mitigating the
plant (i.e., no new or different type of Force (TSTF) Traveler TSTF–542, event as the current requirements. The
equipment will be installed) or a change in Revision 2, ‘‘Reactor Pressure Vessel proposed controls provide escalating
the methods governing normal plant Water Inventory Control.’’ The proposed compensatory measures to be established as
operation. In addition, the changes do not change would replace existing TS calculated drain times decrease, such as
impose any new or different requirements. requirements related to ‘‘operations verification of a second method of water
The changes do not alter assumptions made injection and additional confirmations that
in the safety analysis. The proposed changes
with a potential for draining the reactor
containment and/or filtration would be
are consistent with the safety analysis vessel’’ (OPDRVs) with new available if needed.
assumptions and current plant operating requirements on reactor pressure vessel The proposed change reduces or eliminates
practice. (RPV) water inventory control (WIC) to some requirements that were determined to
Therefore, the proposed changes do not protect Safety Limit 2.1.1.3. Safety Limit be unnecessary to manage the consequences
create the possibility of a new or different 2.1.1.3 requires reactor vessel water of an unexpected draining event, such as
kind of accident from any accident level to be greater than the top of active automatic initiation of an ECCS subsystem
previously evaluated. and control room ventilation. These changes
irradiated fuel.
3. Does the proposed change involve a do not affect the consequences of any
significant reduction in the margin of safety?
Basis for proposed no significant
hazards consideration determination: accident previously evaluated since a
Response: No. draining event in Modes 4 and 5 is not a
The design, operation, testing methods, As required by 10 CFR 50.91(a), the previously evaluated accident and the
and acceptance criteria for systems, licensee has provided its analysis of the requirements are not needed to adequately
structures, and components (SSCs), specified issue of no significant hazards respond to a draining event.
in applicable codes and standards (or consideration, which is presented Therefore, the proposed change does not
alternatives approved for use by the NRC) below: involve a significant increase in the
will continue to be met as described in the probability or consequences of an accident
plant licensing basis (including the final 1. Does the proposed amendment involve
a significant increase in the probability or previously evaluated.
safety analysis report and bases to TS), since 2. Does the proposed amendment create
these are not affected by changes to the consequences of an accident previously
evaluated? the possibility of a new or different kind of
surveillance frequencies. Similarly, there is accident from any accident previously
no impact to safety analysis acceptance Response: No.
The proposed change replaces existing TS evaluated?
criteria as described in the plant licensing
requirements related to OPDRVs with new Response: No.
basis. To evaluate a change in the relocated
requirements on RPV WIC that will protect The proposed change replaces existing TS
surveillance frequency, Entergy will perform
Safety Limit 2.1.1.3. Draining of RPV water requirements related to OPDRVs with new
a probabilistic risk evaluation using the
inventory in Mode 4 (i.e., cold shutdown) requirements on RPV WIC that will protect
guidance contained in NRC approved NEI
and Mode 5 (i.e., refueling) is not an accident Safety Limit 2.1.1.3. The proposed change
[Nuclear Energy Institute] 04–10, Rev. 1, in
previously evaluated and, therefore, will not alter the design function of the
accordance with the TS SFCP. NEI 04–10,
replacing the existing TS controls to prevent equipment involved. Under the proposed
Rev. 1, methodology provides reasonable
or mitigate such an event with a new set of change, some systems that are currently
acceptance guidelines and methods for
evaluating the risk increase of proposed controls has no effect on any accident required to be operable during OPDRVs
changes to surveillance frequencies previously evaluated. RPV water inventory would be required to be available within the
consistent with Regulatory Guide 1.177. control in Mode 4 or Mode 5 is not an limiting drain time or to be in service
Therefore, the proposed changes do not initiator of any accident previously depending on the limiting drain time. Should
involve a significant reduction in a margin of evaluated. The existing OPDRV controls or those systems be unable to be placed into
safety. the proposed RPV WIC controls are not service, the consequences are no different
mitigating actions assumed in any accident than if those systems were unable to perform
The NRC staff has reviewed the previously evaluated. their function under the current TS
licensee’s analysis and, based on this The proposed change reduces the requirements.
review, it appears that the three probability of an unexpected draining event The event of concern under the current
standards of 10 CFR 50.92(c) are (which is not a previously evaluated requirements and the proposed change is an
satisfied. Therefore, the NRC staff accident) by imposing new requirements on unexpected draining event. The proposed
proposes to determine that the the limiting time in which an unexpected change does not create new failure
draining event could result in the reactor mechanisms, malfunctions, or accident
amendment request involves no initiators that would cause a draining event
vessel water level dropping to the top of the
significant hazards consideration. active fuel (TAF). These controls require or a new or different kind of accident not
Attorney for licensee: Anna Vinson cognizance of the plant configuration and previously evaluated or included in the
Jones, Senior Counsel, Entergy Services, control of configurations with unacceptably design and licensing bases.
Inc., 101 Constitution Avenue NW, short drain times. These requirements reduce Therefore, the proposed change does not
Suite 200 East, L–ENT–WDC, the probability of an unexpected draining create the possibility of a new or different
Washington, DC 20001. event. The current TS requirements are only kind of accident from any previously
NRC Branch Chief: Robert J. mitigating actions and impose no evaluated.
Pascarelli. requirements that reduce the probability of 3. Does the proposed amendment involve
an unexpected draining event. a significant reduction in a margin of safety?
Entergy Operations, Inc.; System Energy The proposed change reduces the Response: No.
amozie on DSK3GDR082PROD with NOTICES1
Resources, Inc.; Cooperative Energy, A consequences of an unexpected draining The proposed change replaces existing TS
Mississippi Electric Cooperative; and event (which is not a previously evaluated requirements related to OPDRVs with new
Entergy Mississippi, Inc., Docket No. 50– accident) by requiring an Emergency Core requirements on RPV WIC. The current
416, Grand Gulf Nuclear Station, Unit Cooling System (ECCS) subsystem to be requirements do not have a stated safety basis
No. 1, Claiborne County, Mississippi operable at all times in Modes 4 and 5. The and no margin of safety is established in the
current TS requirements do not require any licensing basis. The safety basis for the new
Date of amendment request: April 10, water injection systems, ECCS or otherwise, requirements is to protect Safety Limit
2018. A publicly-available version is in to be Operable in certain conditions in Mode 2.1.1.3. New requirements are added to
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![26104 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices
determine the limiting time in which the Response: No. Attorney for licensee: Anna Vinson
RPV water inventory could drain to the top The proposed changes to the GGNS EALs Jones, Senior Counsel/Legal
of the fuel in the reactor vessel should an do not involve any physical changes to plant Department, Entergy Services, Inc., 101
unexpected draining event occur. Plant equipment or systems and do not alter the
assumptions of any accident analyses. The
Constitution Avenue NW, Suite 200
configurations that could result in lowering
the RPV water level to the TAF within one proposed changes do not adversely affect East, Washington, DC 20001.
hour are now prohibited. New escalating accident initiators or precursors and do not NRC Branch Chief: Robert J.
compensatory measures based on the limiting alter design assumptions, plant Pascarelli.
drain time replace the current controls. The configuration, or the manner in which the
plant is operated and maintained. The
Exelon Generation Company, LLC,
proposed TS establish a safety margin by
proposed changes do not adversely affect the Docket Nos. STN 50–456 and STN 50–
providing defense-in-depth to ensure that the
Safety Limit is protected and to protect the ability of structures, systems or components 457, Braidwood Station, Unit Nos. 1 and
public health and safety. While some less (SSCs) to perform intended safety functions 2, Will County, Illinois, and Docket Nos.
restrictive requirements are proposed for in mitigating the consequences of an STN 50–454 and STN 50–455, Byron
plant configurations with long calculated initiating event within the assumed Station, Unit Nos. 1 and 2, Ogle County,
drain times, the overall effect of the change acceptance limits. Illinois
is to improve plant safety and to add safety Therefore, the changes do not involve a
significant increase in the probability or Date of amendment request: April 2,
margin.
Therefore, the proposed change does not consequences of an accident previously 2018. A publicly-available version is in
involve a significant reduction in a margin of evaluated. ADAMS under Accession No.
2. Do the proposed changes create the ML18092B081.
safety.
possibility of a new or different kind of Description of amendment request:
The NRC staff has reviewed the accident from any accident previously The proposed amendments would
licensee’s analysis and, based on this evaluated?
Response: No. revise Technical Specification 3.2.3 to
review, it appears that the three require that the axial flux difference be
standards of 10 CFR 50.92(c) are No new accident scenarios, failure
mechanisms, or limiting single failures are maintained within the limits specified
satisfied. Therefore, the NRC staff introduced as a result of the proposed in the core operating limits report
proposes to determine that the changes. The changes do not challenge the during MODE 1 with reactor thermal
amendment request involves no integrity or performance of any safety-related power greater or equal to 50 percent. An
significant hazards consideration. systems. No plant equipment is installed or associated change would also be made
Attorney for licensee: Anna Vinson removed, and the changes do not alter the
design, physical configuration, or method of
to the NOTE modifying surveillance
Jones, Senior Counsel/Legal 3.2.3.1.
Department, Entergy Services, Inc., 101 operation of any plant SSC. Because EALs are
not accident initiators and no physical Basis for proposed no significant
Constitution Avenue NW, Suite 200 changes are made to the plant, no new causal hazards consideration determination:
East, Washington, DC 20001. mechanisms are introduced. As required by 10 CFR 50.91(a), the
NRC Branch Chief: Robert J. Therefore, the changes do not create the licensee has provided its analysis of the
Pascarelli. possibility of a new or different kind of issue of no significant hazards
accident from an accident previously consideration, which is presented
Entergy Operations, Inc.; System Energy evaluated.
Resources, Inc.; Cooperative Energy, A 3. Do the proposed changes involve a
below:
Mississippi Electric Cooperative; and significant reduction in a margin of safety? 1. Does the proposed change involve a
Entergy Mississippi, Inc., Docket No. 50– Response: No. significant increase in the probability or
416, Grand Gulf Nuclear Station, Unit Margin of safety is associated with the consequences of an accident previously
No. 1 (GGNS), Claiborne County, ability of the fission product barriers (i.e., evaluated?
fuel cladding, reactor coolant system Response: No.
Mississippi
pressure boundary, and containment The proposed amendment requires that the
Date of amendment request: April 27, structure) to limit the level of radiation dose AFD [axial flux difference] be maintained
2018. A publicly-available version is in to the public. The proposed changes do not within the limits specified in the COLR [core
ADAMS under Accession No. impact operation of the plant and no accident operating limits report] at-all-times during
ML18117A514. analyses are affected by the proposed MODE 1 when reactor power is ≥50% RTP
changes. The changes do not affect the [reactor thermal power]. This requirement
Description of amendment request: Technical Specifications or the method of will ensure that all FRD [fuel rod design]
The proposed amendment would revise operating the plant. Additionally, the performance criteria remain satisfied during
the Emergency Plan to adopt the proposed changes will not relax any criteria ANS [American Nuclear Society] Condition II
Nuclear Energy Institute’s (NEI’s) used to establish safety limits and will not events (i.e., Faults of Moderate Frequency);
revised Emergency Action Level (EAL) relax any safety system settings. The safety thus, ensuring the integrity of the fuel rod
scheme described in NEI 99–01, analysis acceptance criteria are not affected cladding. It is noted that maintaining AFD
Revision 6, ‘‘Development of Emergency by these changes. The proposed changes will within the COLR limits at-all-times when
Action Levels for Non-Passive Reactors’’
not result in plant operation in a ≥50% RTP is the normal operating practice
configuration outside the design basis. The as specified in plant procedures.
(ADAMS Accession No. ML110240324), proposed changes do not adversely affect The proposed change will have no impact
which has been endorsed by the NRC systems that respond to safely shut down the on accident initiators or precursors; does not
(ADAMS Accession No. ML12346A463). plant and to maintain the plant in a safe alter accident analysis assumptions; does not
Basis for proposed no significant shutdown condition. involve any physical plant modifications that
hazards consideration determination: Therefore, the changes do not involve a would alter the design or configuration of the
As required by 10 CFR 50.91(a), the significant reduction in a margin of safety. facility, or the manner in which the plant is
licensee has provided its analysis of the The NRC staff has reviewed the maintained; and does not impact the
amozie on DSK3GDR082PROD with NOTICES1
issue of no significant hazards licensee’s analysis and, based on this probability of operator error.
The proposed amendment will not impact
consideration, which is presented review, it appears that the three
the ability of structures, systems, and
below: standards of 10 CFR 50.92(c) are components (SSCs) from performing their
1. Do the proposed changes involve a satisfied. Therefore, the NRC staff intended functions to mitigate the
significant increase in the probability or proposes to determine that the consequences of an accident. All accident
consequences of an accident previously amendment request involves no analysis acceptance criteria will continue to
evaluated? significant hazards consideration. be met as the proposed change will not affect
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![Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 26105
the source term, containment isolation Generation Company, LLC, 4300 that the specification only applies to Alloy
function, or radiological release assumptions Winfield Road, Warrenville, IL 60555. 800 tube sleeves installed in the steam
for any accident previously evaluated. NRC Branch Chief: David J. Wrona. generator tubesheet. The design of Alloy 800
Based on the above discussion, the sleeves installed in steam generator tube
proposed amendment does not involve a FirstEnergy Nuclear Operating locations other than the tubesheet does not
significant increase in the probability or Company, Docket No. 50–412, Beaver include a nickel band. For these sleeves,
consequences of an accident previously Valley Power Station, Unit No. 2, Beaver nondestructive examination methods have
evaluated. County, Pennsylvania been demonstrated to be effective and limits
2. Does the proposed change create the on sleeve operating life are not necessary.
possibility of a new or different kind of Date of amendment request: March This proposed amendment does not change
accident from any accident previously 28, 2018. A publicly-available version is the intent of the specification.
evaluated? in ADAMS under Accession No. The second change to TS 5.5.5.2.f.3,
Response: No. ML18087A293. increases the number of fuel cycles Alloy 800
The proposed change formalizes the Description of amendment request: tube sleeves installed in the tubesheet may
existing operating practice of maintaining the The amendment would revise Technical remain in service. The leak-limiting Alloy
AFD within the limits specified in the COLR 800 sleeves are designed using the applicable
Specification (TS) 5.5.5.2.d, ‘‘Provisions
at-all-times during MODE 1 when reactor American Society of Mechanical Engineers
for SG [Steam Generator] Tube
power is ≥ 50% RTP. This change ensures (ASME) Boiler and Pressure Vessel Code and,
Inspection,’’ and TS 5.5.5.2.f, therefore, meet the design objectives of the
that all FRD performance criteria remain
satisfied during ANS Condition II events. The ‘‘Provisions for SG Tube Repair original steam generator tubing. The applied
ANS Condition II events have all been Methods.’’ More specifically, TSs stresses and fatigue usage for the sleeves are
previously evaluated in the Updated Final 5.5.5.2.d.5 and 5.5.5.2.f.3 would be bounded by the limits established in the
Safety Analysis Report. simplified and clarified, respectively, ASME Code. Mechanical testing has shown
The proposed change does not involve a without changing the intent of the that the structural strength of sleeves under
design change or other changes that would specifications. Specification 5.5.5.2.f.3 normal, upset, emergency, and faulted
impact safety-related SSCs from performing conditions provides margin to the acceptance
would also be amended by changing the limits. These acceptance limits bound the
their specified safety functions. number of fuel cycles that Westinghouse
The proposed change does not result in the most limiting (three times normal operating
Electric Company, LLC leak-limiting pressure differential) burst margin of NRC
creation of any new accident precursors; does
not result in changes to any existing accident Alloy 800 sleeves may remain in Regulatory Guide 1.121, ‘‘Bases for Plugging
scenarios; and does not introduce any operation. Degraded PWR Steam Generator Tubes.’’
operational changes or mechanisms that Basis for proposed no significant The leak-limiting Alloy 800 sleeve depth-
would create the possibility of a new or hazards consideration determination: based structural limit is determined using
different kind of accident. As required by 10 CFR 50.91(a), the NRC guidance and the pressure stress
Therefore, the proposed amendment does licensee has provided its analysis of the equation of ASME Code, Section III with
not create the possibility of a new or different issue of no significant hazards margin added to account for the
kind of accident from any previously configuration of long axial cracks.
consideration, which is presented Calculations show that a depth-based limit of
evaluated. below:
3. Does the proposed change involve a 45 percent through-wall degradation is
significant reduction in a margin of safety? 1. Does the proposed amendment involve acceptable. However, Technical
Response: No. a significant increase in the probability or Specifications 5.5.5.2.c.2 and 5.5.5.2.c.3
The proposed change to maintain the AFD consequences of an accident previously provide additional margin by requiring an
within the limits specified in the COLR at- evaluated? Alloy 800 sleeved tube to be plugged on
all-times during MODE 1 when reactor power Response: No. detection of any flaw in the sleeve or in the
is ≥ 50% RTP ensures that all FRD Proposed amendment of Technical pressure boundary portion of the original
performance criteria remain satisfied during Specification 5.5.5.2.d.5 to simplify the tube wall in the sleeve to tube joint.
ANS Condition II events; and thus, will description of the required inspection region, Degradation of the original tube adjacent to
maintain the existing margin of safety related and Technical Specification 5.5.5.2.f.3 to the nickel band of an Alloy 800 sleeve
to FRD performance criteria and ensure the clarify that this specification is only installed in the tubesheet, regardless of
integrity of the fuel rod cladding. The AFD applicable to sleeves installed in the steam depth, would not prevent the sleeve from
limits specified in the COLR have been generator tubesheet and change the number satisfying design requirements. Thus, flaw
established in accordance with the analysis of fuel cycles that an Alloy 800 steam detection capabilities within the original tube
approach described in NRC-approved generator tubesheet sleeve may remain in adjacent to the sleeve nickel band are a
Westinghouse Topical Reports. service from five to eight fuel cycles of defense-in-depth measure, and are not
In addition, this change will have no operation, does not affect structures, systems necessary in order to justify continued
impact on the margin of safety associated or components of the plant, plant operations, operation of the sleeved tube.
with other reactor core safety parameters design functions or analyses that verify the Evaluation of repaired steam generator tube
such as fuel hot channel factors, core power capability of structures, systems or testing and analysis indicates that there are
tilt ratios, loss of coolant accident peak components to perform a design function. no detrimental effects on the leak-limiting
cladding temperature and peak local power The proposed amendment does not increase Alloy 800 sleeve or sleeved tube assembly
density. the likelihood of steam generator tube sleeve from reactor coolant system flow, primary or
Therefore, the proposed amendment does leakage. secondary coolant chemistries, thermal
not involve a significant reduction in a The proposed amendment of Technical conditions or transients, or pressure
margin of safety. Specification 5.5.5.2.d.5 to simplify the conditions that may be experienced at Beaver
description of the required inspection region, Valley Power Station, Unit No. 2.
The NRC staff has reviewed the makes it clear that the steam generator parent Westinghouse is not aware of, and has no
licensee’s analysis and, based on this tube is to be inspected in the areas where the knowledge of any reports of parent-tube
review, it appears that the three joints will be established prior to installation stress corrosion cracking (SCC) in the sleeve
amozie on DSK3GDR082PROD with NOTICES1
standards of 10 CFR 50.92(c) are of the sleeve, regardless of the sleeve roll joint region for any Westinghouse sleeve
satisfied. Therefore, the NRC staff location. This proposed amendment does not design.
change the intent of the specification. The proposed increase in the number of
proposes to determine that the
The proposed amendment of TS 5.5.5.2.f.3 fuel cycles Alloy 800 tube sleeves installed
requested amendments involve no includes two changes. The first change in the tubesheet may remain in service has
significant hazards consideration. would add the words ‘‘installed in the hot- no effect on sleeve operation or capability of
Attorney for licensee: Tamra Domeyer, leg or cold-leg tubesheet region’’ after the the sleeve to perform its design function. The
Associate General Counsel, Exelon words ‘‘An Alloy 800 sleeve’’ to make it clear mechanical and leakage tests have confirmed
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![26106 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices
that degradation of the parent tube adjacent Therefore, the proposed amendment does Attorney for licensee: David W.
to the nickel band will not prevent the sleeve not create the possibility of a new or different Jenkins, FirstEnergy Nuclear Operating
from satisfying its design function. kind of accident from any previously Company, FirstEnergy Corporation, 76
Consequences of a hypothetical failure of evaluated.
the leak-limiting Alloy 800 sleeve and tube 3. Does the proposed amendment involve
South Main Street, Akron, OH 44308.
assembly are bounded by the current main a significant reduction in a margin of safety?
NRC Branch Chief: James Danna.
steam line break and steam generator tube Response: No. PSEG Nuclear LLC, Docket No. 50–354,
rupture accident analyses described in the Proposed amendment of Technical Hope Creek Generating Station, Salem
Beaver Valley Power Station, Unit No. 2 Specification 5.5.5.2.d.5 to simplify the
Updated Final Safety Analysis Report. The
County, New Jersey
description of the required inspection region,
total number of plugged steam generator and Technical Specification 5.5.5.2.f.3 to Date of amendment request: March
tubes (including equivalency associated with clarify that this specification is only 28, 2018. A publicly-available version is
installed sleeves) is required to be consistent applicable to sleeves installed in the steam in ADAMS under Accession No.
with accident analysis assumptions. The generator tubesheet, do not change the intent ML18087A095.
sleeve and tube assembly leakage during of these requirements or reduce the margin Description of amendment request:
plant operation is required to be within the of safety. The proposed amendment to
allowable Technical Specification leakage The amendment would revise Technical
Technical Specification 5.5.5.2.f.3 to change
limits and accident analysis assumptions. the number of fuel cycles that an Alloy 800 Specification (TS) 3/4.8.1, ‘‘AC
Therefore, the proposed amendment does steam generator tubesheet sleeve may remain [Alternating Current] Sources—
not involve a significant increase in the in service from five to eight fuel cycles of Operating’’; specifically, ACTION b
probability or consequences of an accident operation, does not affect a design basis or concerning one inoperable emergency
previously evaluated. safety limit (that is, the controlling numerical diesel generator (EDG). The proposed
2. Does the proposed amendment create value for a parameter established in the change would remove the Salem
the possibility of a new or different kind of Updated Final Safety Analysis Report or the
accident from any accident previously Nuclear Generating Station, Unit No. 3
license) or reduce the margin of safety.
evaluated? (Salem Unit 3), gas turbine generator
The proposed amendment to Technical
Response: No. Specification 5.5.5.2.f.3 increases the number and replace it with portable diesel
Proposed amendment of Technical of fuel cycles Alloy 800 tube sleeves installed generators.
Specification 5.5.5.2.d.5 to simplify the in the tubesheet may remain in service to Basis for proposed no significant
description of the required inspection region, eight fuel cycles of operation. hazards consideration determination:
and Technical Specification 5.5.5.2.f.3 to Implementation of this proposed amendment As required by 10 CFR 50.91(a), the
clarify that this specification is only would not affect a design basis or safety limit licensee has provided its analysis of the
applicable to sleeves installed in the steam or reduce the margin of safety. The repair of
generator tubesheet do not change the intent issue of no significant hazards
degraded steam generator tubes with leak- consideration, which is presented
of these specifications, and do not affect the limiting Alloy 800 sleeves restores the
design function or operation of the tube below:
structural integrity of the degraded tube
sleeves. The proposed amendment of under normal operating and postulated 1. Does the proposed change involve a
Technical Specification 5.5.5.2.f.3 to change accident conditions. Minimum reactor significant increase in the probability or
the number of fuel cycles that an Alloy 800 coolant system flow rate from the cumulative consequences of an accident previously
steam generator tubesheet sleeve may remain effect of repaired (sleeved) and plugged tubes evaluated?
in service from five to eight fuel cycles of will be greater than the flow rate limit Response: No.
operation, does not affect the design function established in the Technical Specification The proposed change removes the
or operation of the tube sleeves. Since these limiting condition for operation 3.4.1. The requirement for the Salem Unit 3 gas turbine
changes do not create any credible new design safety factors utilized for the sleeves generator (GTG) and replaces it with the
failure mechanisms, malfunctions, or are consistent with the safety factors in the supplemental power source during the
accident initiators not considered in the American Society of Mechanical Engineers existing extended allowable outage time for
design or licensing bases, the changes do not Boiler and Pressure Vessel Code used in the the A or B EDG. The emergency diesel
create the possibility of a new or different original steam generator design. Tubes with generators are safety related components
kind of accident from any previously sleeves are subject to the same safety factors which provide backup electrical power
evaluated. as the original tubes, which are described in supply to the onsite Safeguards Distribution
The leak-limiting Alloy 800 sleeves are the performance criteria for steam generator System. The emergency diesel generators are
designed using the applicable ASME Code, not accident initiators; the EDGs are designed
tube integrity in the existing Technical
and therefore meet the objectives of the to mitigate the consequences of previously
Specifications. The sleeve and portions of the
original steam generator tubing. As a result, evaluated accidents including a loss of offsite
installed sleeve and tube assembly that
the functions of the steam generator will not power. (During normal operation, the
represent the reactor coolant pressure
be significantly affected by the installation of proposed portable diesel generators will not
boundary will be monitored, and a sleeved
the proposed sleeve. Therefore, the only be connected to the plant.)
tube will be plugged if a flaw is detected in
credible failure modes for the sleeve and tube The proposed change does not adversely
the sleeve or in the pressure boundary
are to leak or rupture, which has already affect accident initiators or precursors nor
portion of the original tube wall in the leak-
been evaluated. The continued integrity of alter the design assumptions, conditions, or
limiting sleeve and tube assembly. Use of the
the installed sleeve and tube assembly is configuration of the facility or the manner in
periodically verified as required by the previously-identified design criteria and
which the plant is operated and maintained.
Technical Specifications, and a sleeved tube design verification testing ensures that the The proposed change does not alter or
will be plugged on detection of a flaw in the margin of safety is not significantly different prevent the ability of structures, systems, and
sleeve or in the pressure boundary portion of from the original steam generator tubes. components (SSCs) from performing their
the original tube wall in the sleeve to tube Therefore, the proposed amendment does intended function to mitigate the
joint. not involve a significant reduction in a consequences of an initiating event within
The proposed amendment to Technical margin of safety. the assumed acceptance limits. The proposed
Specification 5.5.5.2.f.3 increases the number The NRC staff has reviewed the change does not affect the source term,
amozie on DSK3GDR082PROD with NOTICES1
of fuel cycles Alloy 800 tube sleeves installed licensee’s analysis and, based on this containment isolation, or radiological release
in the tubesheet may remain in service to review, it appears that the three assumptions used in evaluating the
eight fuel cycles of operation. radiological consequences of an accident
Implementation of this proposed amendment
standards of 10 CFR 50.92(c) are previously evaluated. The proposed change
has no significant effect on either the satisfied. Therefore, the NRC staff is consistent with safety analysis
configuration of the plant, the manner in proposes to determine that the assumptions and resultant consequences.
which it is operated, or ability of the sleeve amendment request involves no Therefore, the proposed change does not
to perform its design function. significant hazards consideration. involve a significant increase in the
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![Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 26107
probability or consequences of an accident Accession Nos. ML18071A349 and credited in the SQN accident analysis and do
previously evaluated. ML18102B430, respectively. not provide any additional safety margin.
2. Does the proposed change create the Description of amendment request: The results of accident analyses remain
possibility of a new or different kind of The amendments would make changes unchanged by this request. Therefore, the
accident from any accident previously proposed change does not involve a
to the SQN Essential Raw Cooling Water significant reduction in a margin of safety.
evaluated?
Response: No. (ERCW) Motor Control Centers (MCCs)
and revise the Updated Final Safety The NRC staff has reviewed the
The proposed change removes the
Analysis Report (UFSAR). licensee’s analysis and, based on this
requirement for the Salem Unit 3 gas turbine
generator (GTG) and replaces it with the Basis for proposed no significant review, it appears that the three
supplemental power source during the hazards consideration determination: standards of 10 CFR 50.92(c) are
existing extended allowable outage time for As required by 10 CFR 50.91(a), the satisfied. Therefore, the NRC staff
the A or B EDG. The proposed change does licensee has provided its analysis of the proposes to determine that the
not alter or involve any design basis accident issue of no significant hazards amendment request involves no
initiators. Equipment will be operated in the consideration, which is presented significant hazards consideration.
same configuration and manner that is
below: Attorney for licensee: General
currently allowed and designed for. Counsel, Tennessee Valley Authority,
Therefore, the proposed change does not 1. Does the proposed amendment involve 400 West Summit Hill Drive, 6A West
create the possibility of a new or different a significant increase in the probability or
Tower, Knoxville, TN 37902.
kind of accident from any [accident] consequence of an accident previously
evaluated?
NRC Acting Branch Chief: Brian W.
previously evaluated.
Response: No. Tindell.
3. Does the proposed change involve a
significant reduction in a margin of safety? The proposed change does not alter the Vistra Operations Company LLC, Docket
Response: No. safety function of any structure, system, or Nos. 50–445 and 50–446, Comanche
The proposed change does not alter the component, does not modify the manner in
which the plant is operated, and does not
Peak Nuclear Power Plant, Unit Nos. 1
permanent plant design, including
alter equipment out-of-service time. In and 2, Somervell County, Texas
instrument set points, nor does it change the
assumptions contained in the safety analyses. addition, this request does not degrade the Date of amendment request: March
The proposed change does not impact the ability of the ERCW to perform its intended 29, 2018. A publicly-available version is
redundancy or availability requirements of safety function. Therefore, the proposed in ADAMS under Accession No.
offsite power supplies or change the ability change does not involve a significant ML18102A516.
of the plant to cope with station blackout increase in the probability or consequence of
Description of amendment request:
[(SBO)] events. an accident previously evaluated.
2. Does the proposed amendment create The amendments would revise
The EDGs continue to meet their design Technical Specification 3.3.2,
requirements; there is no reduction in the possibility of a new or different kind of
capability or change in design configuration. accident from any accident previously ‘‘Engineered Safety Feature Actuation
The EDG response to LOOP [loss of offsite evaluated? System (ESFAS) Instrumentation,’’ to
power], LOCA [loss-of-coolant accident], Response: No. change the applicability of when the
SBO, or fire is not changed by this proposed The proposed change does not involve any automatic auxiliary feedwater actuation
amendment; there is no change to the EDG physical changes to plant safety related due to the trip of all main feedwater
operating parameters. The remaining structure, system or component or alter the pumps is required to be operable at
operable emergency diesel generators are modes of plant operation in a manner that is Comanche Peak Nuclear Power Plant,
adequate to supply electrical power to the outside the bounds of the system design
Unit Nos. 1 and 2.
onsite Safeguards Distribution System. The analyses. The proposed change to complete
Basis for proposed no significant
proposed change does not alter a design basis the design change for the removal of
mechanical interlock device from the feeder hazards consideration determination:
or safety limit; therefore it does not
breakers and tie breakers for the ERCW MCCs As required by 10 CFR 50.91(a), the
significantly reduce the margin of safety. The
EDGs will continue to operate per the and to revise the ERCW System Description licensee has provided its analysis of the
existing design and regulatory requirements. in Section 9.2.2.2 of the SQN UFSAR to issue of no significant hazards
Therefore, it is concluded that the describe the normal and alternate power consideration, which is presented
proposed change does not involve a sources for the ERCW system does not create below:
significant reduction in a margin of safety. the possibility for an accident or malfunction
1. Do the proposed changes involve a
of a different type than any evaluated
The NRC staff has reviewed the significant increase in the probability or
previously in SQN’s UFSAR. The proposal
licensee’s analysis and, based on this consequences of an accident previously
does not alter the way any safety related
evaluated?
review, it appears that the three structure, system or component functions
Response: No.
standards of 10 CFR 50.92(c) are and does not modify the manner in which The design basis events which impose
satisfied. Therefore, the NRC staff the plant is operated. Therefore, the proposed auxiliary feedwater safety function
proposes to determine that the change does not create the possibility of a requirements are loss of all AC [alternating
new or different kind of accident from any current] power to plant auxiliaries, loss of
amendment request involves no
accident previously evaluated. normal feedwater, steam generator fault in
significant hazards consideration. 3. Does the proposed amendment involve
Attorney for licensee: Jeffrie J. Keenan, either the feedwater or steam lines, and small
a significant reduction in a margin of safety? break loss of coolant accidents. These design
PSEG Nuclear LLC–N21, P.O. Box 236, Response: No. basis event evaluations assume actuation of
Hancocks Bridge, NJ 08038. The proposed change to remove the auxiliary feedwater due to station blackout,
NRC Branch Chief: James G. Danna. mechanical interlock device from the feeder low-low steam generator level or a safety
breakers and tie breakers for ERCW MCCs injection signal. The anticipatory auxiliary
Tennessee Valley Authority, Docket 1B–B and 2B–B and to revise the ERCW feedwater automatic start signals from the
amozie on DSK3GDR082PROD with NOTICES1
Nos. 50–327 and 50–328, Sequoyah System Description in Section 9.2.2.2 of the main feedwater pumps are not credited in
Nuclear Plant, Unit Nos. 1 and 2 (SQN), SQN UFSAR to describe the normal and any design basis accidents and are, therefore,
Hamilton County, Tennessee alternate power sources for the ERCW system not part of the primary success path for
does not reduce the margin of safety because postulated accident mitigation as defined by
Date of amendment request: March 9, ERCW will continue to perform its safety 10 CFR 50.36(c)(2)(ii), Criterion 3. Modifying
2018, as supplemented by letter dated function. The design features provided by the MODE 2 Applicability for this function will
April 11, 2018. Publicly-available mechanical interlock device are not not impact any previously evaluated design
versions are in ADAMS under described in the SQN UFSAR, are not basis accidents.
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![26108 Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices
Therefore, the proposed change does not 10 CFR chapter I, which are set forth in Amendment No.: 180. A publicly-
involve a significant increase in the the license amendment. available version is in ADAMS under
probability or consequences of an accident A notice of consideration of issuance Accession No. ML18110A133;
previously evaluated. of amendment to facility operating
2. Do the proposed changes create the
documents related to this amendment
possibility of a new or different kind of
license or combined license, as are listed in the Safety Evaluation
accident from any accident previously applicable, proposed no significant enclosed with the amendment.
evaluated? hazards consideration determination, Facility Operating License No. NPF–
Response: No. and opportunity for a hearing in 58: The amendment revised the Facility
This technical specification change allows connection with these actions, was Operating License and TSs.
for an operational allowance during MODE 2 published in the Federal Register as Date of initial notice in Federal
while placing main feedwater pumps in indicated.
service. This change involves an anticipatory
Register: August 1, 2017 (82 FR 35841).
Unless otherwise indicated, the The supplemental letter dated January
auxiliary feedwater automatic start function Commission has determined that these
that is not credited in the accident analysis. 30, 2018, provided additional
Since this change only affects the conditions amendments satisfy the criteria for information that clarified the
at which this automatic start function needs categorical exclusion in accordance application, did not expand the scope of
to be operable and does not affect the with 10 CFR 51.22. Therefore, pursuant the application as originally noticed,
function that actuates auxiliary feedwater to 10 CFR 51.22(b), no environmental and did not change the NRC staff’s
due to loss of offsite power, low-low steam impact statement or environmental original proposed no significant hazards
generator level or a safety injection signal, it assessment need be prepared for these
will not be an initiator to a new or different consideration determination as
amendments. If the Commission has published in the Federal Register.
kind of accident from any accident prepared an environmental assessment
previously evaluated. The Commission’s related evaluation
Therefore, the proposed change does not under the special circumstances of the amendment is contained in a
create the possibility of a new or different provision in 10 CFR 51.22(b) and has Safety Evaluation dated May 16, 2018.
kind of accident from any previously made a determination based on that No significant hazards consideration
evaluated. assessment, it is so indicated. comments received: No.
3. Do the proposed changes involve a For further details with respect to the
significant reduction in a margin of safety? action see (1) the applications for NextEra Energy Duane Arnold, LLC,
Response: No. amendment; (2) the amendment; and (3) Docket No. 50–331, Duane Arnold
This technical [s]pecification change the Commission’s related letter, Safety Energy Center (DAEC), Linn County,
involves the automatic start of the auxiliary Iowa
Evaluation and/or Environmental
feedwater pumps due to trip of both main
feedwater pumps, which is not an assumed Assessment, as indicated. All of these Date of amendment request: March
start signal for design basis events. This items can be accessed as described in 24, 2017.
change does not modify any values or limits the ‘‘Obtaining Information and Brief description of amendment: The
involved in a safety related function or Submitting Comments’’ section of this amendment revised the DAEC Technical
accident analysis. document.
Therefore, the proposed change does not Specification (TS) Table 3.3.2.1–1,
involve a significant reduction in a margin of FirstEnergy Nuclear Operating ‘‘Control Rod Block Instrumentation,’’
safety. Company, Docket No. 50–440, Perry by relocating certain cycle-specific
Nuclear Power Plant, Unit No. 1, Lake Minimum Critical Power Ratio values to
The NRC staff has reviewed the
County, Ohio the DAEC Core Operating Limits Report.
licensee’s analysis and, based on this
Date of amendment request: June 8, The amendment also added a
review, it appears that the three
2017. requirement to DAEC TS 5.6.5, ‘‘Core
standards of 10 CFR 50.92(c) are
Brief description of amendment: The Operating Limits Report.’’
satisfied. Therefore, the NRC staff
amendment revised technical Date of issuance: March 7, 2018.
proposes to determine that the
amendment request involves no specifications (TSs) to reflect previously Effective date: As of the date of its
significant hazards consideration. approved changes made as part of the issuance and shall be implemented by
Attorney for licensee: Timothy P. alternative source term initiative. The September 27, 2018. (Note: This Notice
Matthews, Esq., Morgan, Lewis, and amendment revised the surveillance of Issuance corrects the ‘‘Effective date’’
Bockius, 1111 Pennsylvania Avenue requirements for the control room of Amendment No. 303 originally
NW, Washington, DC 20004. emergency recirculation and annulus noticed in the Federal Register on
NRC Branch Chief: Robert J. exhaust gas treatment systems, which March 27, 2018 (83 FR 13153).
Pascarelli. are consistent with Technical Amendment No.: 303. A publicly-
Specification Task Force (TSTF) available version is in ADAMS under
III. Notice of Issuance of Amendments Accession No. ML18011A059;
Traveler TSTF–522, ‘‘Revise Ventilation
to Facility Operating Licenses and documents related to this amendment
System Surveillance Requirement to
Combined Licenses are listed in the Safety Evaluation
Operate for 10 Hours per Month.’’ The
During the period since publication of amendment also deleted two TS enclosed with the amendment.
the last biweekly notice, the sections related to the fuel handling Amendment No. 303 was corrected by
Commission has issued the following building and fuel handling building letter dated May 7, 2018 (ADAMS
amendments. The Commission has ventilation exhaust system and Accession No. ML18081A074).
determined for each of these increased the allowable secondary Renewed Facility Operating License
amendments that the application containment leakage. Lastly, the No. DPR–49: The amendment revised
amozie on DSK3GDR082PROD with NOTICES1
complies with the standards and amendment revised the TS Table of the Renewed Facility Operating License
requirements of the Atomic Energy Act Contents to reflect administrative and TSs.
of 1954, as amended (the Act), and the changes to the titles of TS sections. Date of initial notice in Federal
Commission’s rules and regulations. Date of issuance: May 16, 2018. Register: May 23, 2017 (82 FR 23627).
The Commission has made appropriate Effective date: As of the date of The Commission’s related evaluation
findings as required by the Act and the issuance and shall be implemented of the amendment is contained in a
Commission’s rules and regulations in within 180 days of issuance. Safety Evaluation dated March 7, 2018.
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![Federal Register / Vol. 83, No. 108 / Tuesday, June 5, 2018 / Notices 26109
No significant hazards consideration presentation at least thirty minutes design certification application. The
comments received: No. before the meeting. Electronic Subcommittee will hear presentations
Dated at Rockville, Maryland, this 29th day recordings will be permitted only by and hold discussions with the NRC
of May, 2018. during those portions of the meeting staff and other interested persons
For the Nuclear Regulatory Commission. that are open to the public. Detailed regarding this matter. The
Gregory F. Suber,
procedures for the conduct of and Subcommittee will gather information,
participation in ACRS meetings were analyze relevant issues and facts, and
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
published in the Federal Register on formulate proposed positions and
Regulation. October 4, 2017 (82 FR 46312). The actions, as appropriate, for deliberation
[FR Doc. 2018–11843 Filed 6–4–18; 8:45 am]
bridgeline number for this meeting is by the Full Committee.
866–822–3032, passcode 8272423#.
BILLING CODE 7590–01–P Members of the public desiring to
Detailed meeting agendas and meeting provide oral statements and/or written
transcripts are available on the NRC comments should notify the Designated
NUCLEAR REGULATORY website at http://www.nrc.gov/reading- Federal Official (DFO), Michael
COMMISSION rm/doc-collections/acrs. Information
Snodderly (Telephone 301–415–2241 or
regarding topics to be discussed,
Email: Michael.Snodderly@nrc.gov) five
Meeting of the Advisory Committee on changes to the agenda, whether the
days prior to the meeting, if possible, so
Reactor Safeguards (ACRS) meeting has been canceled or
that appropriate arrangements can be
Subcommittee on APR1400 rescheduled, and the time allotted to
made. Thirty-five hard copies of each
present oral statements can be obtained
The ACRS Subcommittee on APR1400 presentation or handout should be
from the website cited above or by
will hold a meeting on June 5, 2018, at provided to the DFO thirty minutes
contacting the identified DFO.
11545 Rockville Pike, Room T–2B1, before the meeting. In addition, one
Moreover, in view of the possibility that
Rockville, Maryland 20852. electronic copy of each presentation
the schedule for ACRS meetings may be
The meeting will be open to public should be emailed to the DFO one day
adjusted by the Chairman as necessary
attendance with the exception of before the meeting. If an electronic copy
to facilitate the conduct of the meeting,
portions that may be closed to protect persons planning to attend should check cannot be provided within this
information that is proprietary pursuant with these references if such timeframe, presenters should provide
to 5 U.S.C. 552b(c)(4). The agenda for rescheduling would result in a major the DFO with a CD containing each
the subject meeting shall be as follows: inconvenience. presentation at least thirty minutes
before the meeting. Electronic
Tuesday, June 5, 2018, 8:30 a.m. Until If attending this meeting, please enter
recordings will be permitted only
5:00 p.m. through the One White Flint North
during those portions of the meeting
The Subcommittee will review the Building, 11555 Rockville Pike,
Rockville, Maryland 20852. After that are open to the public. Detailed
APR1400 Design Control Document and procedures for the conduct of and
Safety Evaluation Report with No Open registering with Security, please contact
Ms. Kendra Freeland (Telephone 301– participation in ACRS meetings were
Items, Chapter 17 (Quality Assurance & published in the Federal Register on
Reliability Assurance), Chapter 19.1 415–6207) to be escorted to the meeting
room. October 4, 2017 (82 FR 46312). The
(Probabilistic Risk Assessment), and bridgeline number for this meeting is
Chapter 19.2 (Severe Accident Dated: May 23, 2018. 866–822–3032, passcode 8272423#.
Evaluation). Mark L. Banks,
The Subcommittee will hear Detailed meeting agendas and meeting
Chief, Technical Support Branch, Advisory
presentations by and hold discussions Committee on Reactor Safeguards.
transcripts are available on the NRC
with the NRC staff and Korea Hydro & website at http://www.nrc.gov/reading-
[FR Doc. 2018–12022 Filed 6–4–18; 8:45 am]
Nuclear Power Company regarding this rm/doc-collections/acrs. Information
BILLING CODE 7590–01–P
matter. The Subcommittee will gather regarding topics to be discussed,
information, analyze relevant issues and changes to the agenda, whether the
facts, and formulate proposed positions meeting has been canceled or
NUCLEAR REGULATORY rescheduled, and the time allotted to
and actions, as appropriate, for
COMMISSION present oral statements can be obtained
deliberation by the Full Committee.
Members of the public desiring to from the website cited above or by
Advisory Committee on Reactor contacting the identified DFO.
provide oral statements and/or written Safeguards (ACRS) Meeting of the
comments should notify the Designated Moreover, in view of the possibility that
ACRS Subcommittee on Nuscale; the schedule for ACRS meetings may be
Federal Official (DFO), Christopher Notice of Meeting
Brown (Telephone 301–415–7111 or adjusted by the Chairman as necessary
Email: Christopher.Brown@nrc.gov) five The ACRS Subcommittee on NuScale to facilitate the conduct of the meeting,
days prior to the meeting, if possible, so will hold a meeting on June 6, 2018, at persons planning to attend should check
that appropriate arrangements can be 11545 Rockville Pike, Room T–2B1, with these references if such
made. Thirty-five hard copies of each Rockville, Maryland 20852. rescheduling would result in a major
presentation or handout should be The meeting will be open to public inconvenience.
provided to the DFO thirty minutes attendance. The agenda for the subject If attending this meeting, please enter
amozie on DSK3GDR082PROD with NOTICES1
before the meeting. In addition, one meeting shall be as follows: through the One White Flint North
electronic copy of each presentation building, 11555 Rockville Pike,
Wednesday, June 6, 2018, 8:30 a.m.
should be emailed to the DFO one day Rockville, Maryland. After registering
Until 12:00 p.m.
before the meeting. If an electronic copy with Security, please contact Mr.
cannot be provided within this The Subcommittee will review the Theron Brown (Telephone 301–415–
timeframe, presenters should provide staff’s SER with open items for Chapter 6702 or 301–415–8066) to be escorted to
the DFO with a CD containing each 8, ‘‘Electrical Systems,’’ of the NuScale the meeting room.
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Document Created: 2018-11-02 11:46:04
Document Modified: 2018-11-02 11:46:04
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by July 5, 2018. A request for a hearing must be filed by August 6, 2018. | |
| Contact | Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1384, email: [email protected] | |
| FR Citation | 83 FR 26098 |
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