Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 14142-14156 [2019-06449]

Download as PDF 14142 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices The NRC provides reasonable accommodation to individuals with disabilities where appropriate. If you need a reasonable accommodation to participate in these public meetings, or need this meeting notice or the transcript or other information from the public meetings in another format (e.g., Braille, large print), please notify Kimberly Meyer-Chambers, NRC Disability Program Manager, at 301– 287–0739, by videophone at 240–428– 3217, or by email at Kimberly.MeyerChambers@nrc.gov. Determinations on requests for reasonable accommodation will be made on a case-by-case basis. Members of the public may request to receive this information electronically. If you would like to be added to the distribution, please contact the Nuclear Regulatory Commission, Office of the Secretary, Washington, DC 20555 (301– 415–1969), or by email at Wendy.Moore@nrc.gov. Dated at Rockville, Maryland, this 5th day of April 2019. For the Nuclear Regulatory Commission. Denise L. McGovern, Policy Coordinator, Office of the Secretary. [FR Doc. 2019–07068 Filed 4–5–19; 11:15 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [NRC–2019–0087] I. Obtaining Information and Submitting Comments Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations Nuclear Regulatory Commission. ACTION: Biweekly notice. AGENCY: Pursuant to the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued, from March 12, 2019 to March 25, 2019. The last amozie on DSK9F9SC42PROD with NOTICES SUMMARY: VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 biweekly notice was published on March 26, 2019. DATES: Comments must be filed by May 9, 2019. A request for a hearing must be filed by June 10, 2019. ADDRESSES: You may submit comments by any of the following methods: • Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC–2019–0087. Address questions about NRC dockets IDs in Regulations.gov to Jennifer Borges; telephone: 301–287–9127; email: Jennifer.Borges@nrc.gov. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document. • Mail comments to: Office of Administration, Mail Stop: TWFN–7– A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, ATTN: Program Management, Announcements and Editing Staff. • 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: Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555–0001; telephone: 301–415–5411, email: Shirley.Rohrer@nrc.gov. SUPPLEMENTARY INFORMATION: A. Obtaining Information Please refer to Docket ID NRC–2019– 0087, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publiclyavailable information related to this action by any of the following methods: • Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC–2019–0087. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may obtain publiclyavailable documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/ adams.html. To begin the search, select ‘‘Begin Web-based ADAMS Search.’’ For problems with ADAMS, please contact the NRC’s Public Document Room (PDR) reference staff at 1–800–397–4209, 301– 415–4737, or by email to pdr.resource@ nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document. PO 00000 Frm 00060 Fmt 4703 Sfmt 4703 • 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–2019– 0087, 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. Background Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the 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. III. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission’s ‘‘Agency Rules of Practice and Procedure’’ in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC’s regulations are accessible electronically from the NRC Library on the NRC’s website at http://www.nrc.gov/reading-rm/doccollections/cfr/. Alternatively, a copy of the regulations is available at the NRC’s Public Document Room, located at One White Flint North, Room O1–F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 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 PO 00000 Frm 00061 Fmt 4703 Sfmt 4703 14143 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the ‘‘Electronic Submissions (E-Filing)’’ section of this document. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner’s interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ‘‘Electronic Submissions (E-Filing)’’ section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or Federallyrecognized 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, Federallyrecognized 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 E:\FR\FM\09APN1.SGM 09APN1 14144 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC’s E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The EFiling process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/ e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301–415–1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on the NRC’s public website at http:// www.nrc.gov/site-help/e-submittals/ getting-started.html. Once a participant VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC’s public website at http://www.nrc.gov/ site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC’s E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the EFiling system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC’s Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system. A person filing electronically using the NRC’s adjudicatory E-Filing system may seek assistance by contacting the NRC’s Electronic Filing Help Desk through the ‘‘Contact Us’’ link located on the NRC’s public website at http:// www.nrc.gov/site-help/esubmittals.html, by email to MSHD.Resource@nrc.gov, or by a tollfree call at 1–866–672–7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays. Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. PO 00000 Frm 00062 Fmt 4703 Sfmt 4703 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 application(s), 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. DTE Electric Company, Docket No. 50– 341, Fermi 2, Monroe County, Michigan Date of amendment request: February 8, 2019. A publicly-available version is in ADAMS under Accession No. ML19039A126. Description of amendment request: The amendment would adopt TSTF– 564, ‘‘Safety Limit MCPR [minimum critical power ratio],’’ Revision 2, which E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES revises the Fermi 2 technical specification safety limit on minimum critical power ratio (SLMCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for a safety limit. In addition, technical specification 5.6.5, ‘‘Core Operating Limits Report (COLR),’’ is revised to require the current SLMCPR value to be included in the COLR. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated. 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 previously evaluated? Response: No. The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered 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 amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95% probability with 95% confidence that no fuel rods experience transition boiling during an VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 anticipated transient instead of the current limit based on ensuring that 99.9% of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Llimiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jon P. Christinidis, DTE Energy, Expert Attorney—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226–1279. NRC Branch Chief: David J. Wrona. Entergy Operations, Inc., Docket Nos. 50–313 and 50–368, Arkansas Nuclear One, Units 1 and 2, Pope County, Arkansas Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative Energy, A Mississippi Electric Cooperative; and Entergy Mississippi, LLC, Docket No. 50–416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Entergy Nuclear Operations, Inc., Docket Nos. 50–247 and 50–286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York Entergy Nuclear Operations, Inc., Docket No. 50–255, Palisades Nuclear Plant, Van Buren County, Michigan Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50–458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana Entergy Operations, Inc., Docket No. 50– 382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: January 31, 2019. A publicly-available version is in ADAMS under Accession No. ML19032A256. Description of amendment request: The amendments would revise the technical specifications (TSs) for each of these facilities based on Technical Specifications Task Force (TSTF) Traveler TSTF–529, ‘‘Clarify Use and Application Rules,’’ Revision 4 (ADAMS Accession No. ML16062A271). Specifically, the changes would revise and clarify the TS usage rules for completion times, limiting conditions for operation (LCOs), and surveillance requirements (SRs). PO 00000 Frm 00063 Fmt 4703 Sfmt 4703 14145 Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to Section 1.3 and LCO 3.0.4 have no effect on the requirement for systems to be Operable and have no effect on the application of TS actions. The proposed change to SR 3.0.3 (or equivalent) states that the allowance may only be used when there is a reasonable expectation the surveillance will be met when performed. Since the proposed changes do not significantly affect system Operability, the proposed changes will have no significant effect on the initiating events for accidents previously evaluated and will have no significant effect on the ability of the systems to mitigate accidents previously evaluated. Therefore, it is concluded that these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the TS usage rules do not affect the design or function of any plant systems. The proposed changes do not change the Operability requirements for plant systems or the actions taken when plant systems are not operable. Therefore, it is concluded that the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes clarify the application of Section 1.3 and LCO 3.0.4 and do not result in changes in plant operation. SR 3.0.3 (or equivalent) is revised to allow application of SR 3.0.3 when an SR has not been previously performed if there is reasonable expectation that the SR will be met when performed. This expands the use of SR 3.0.3 while ensuring the affected system is capable of performing its safety function. As a result, plant safety is either improved or unaffected. Therefore, it is concluded that 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, Entergy Services, E:\FR\FM\09APN1.SGM 09APN1 14146 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Exelon Generation Company, LLC, Docket No. 50–461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 50– 333, James A. FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York Exelon Generation Company, LLC, Docket Nos. 50–373 and 50–374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–352 and 50–353, Limerick Generating Station (Limerick), Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, Docket No. 50–410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York amozie on DSK9F9SC42PROD with NOTICES Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50–277 and 50–278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: February 1, 2019, as supplemented by letter dated March 7, 2019. Publicly-available versions are in ADAMS under Accession Nos. ML19032A624 and ML19066A162, respectively. Description of amendment request: The proposed amendments would revise the technical specification (TS) requirements for these facilities related to the safety limit minimum critical power ratio (MCPR) and the core operating limits report (COLR). The proposed amendments are based on Technical Specification Task Force (TSTF) traveler TSTF–564, Revision 2, ‘‘Safety Limit MCPR’’ (ADAMS Accession No. ML18297A361). The proposed amendments for Limerick and FitzPatrick would also make changes to the MCPR and COLR requirements that are outside the scope of TSTF–564, Revision 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. The NRC staff has reviewed the licensee’s analysis against the standards of 10 CFR 50.92(c). The NRC staff’s analysis is presented below: 1. Do the proposed amendments involve a significant increase in the probability or VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 consequences of an accident previously evaluated? Response: No. The proposed amendments revise the TS requirements for the safety limit MCPR and the list of core operating limits to be included in the COLR. The safety limit MCPR is not an initiator of any accident previously evaluated. The revised safety limit values will continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed amendments for Limerick, Units 1 and 2, also include a revision to point to MCPR limits specified in the COLR and clarify references to other specifications. The proposed amendment for FitzPatrick also revises the COLR methodology references by deleting references that are no longer needed and clarifying the remaining reference. The proposed changes do not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendments revise the TS requirements for the safety limit MCPR and the list of core operating limits to be included in the COLR. The proposed amendments for Limerick, Units 1 and 2, also include a revision to point to MCPR limits specified in the COLR and clarify references to other specifications. The proposed amendment for FitzPatrick also revises the COLR methodology references by deleting references that are no longer needed and clarifying the remaining reference. The proposed change will not affect the design function or operation of any structures, systems or components. No new equipment will be installed. As a result, the proposed changes will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed amendments involve a significant reduction in a margin of safety? Response: No. The proposed amendments revise the TS safety limit MCPR and the list of core operating limits to be included in the COLR. The proposed amendments for Limerick, Units 1 and 2, also include a revision to point to MCPR limits specified in the COLR and clarify references to other specifications. The proposed amendment for FitzPatrick also revises the COLR methodology references by deleting references that are no longer needed and clarifying the remaining reference. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety PO 00000 Frm 00064 Fmt 4703 Sfmt 4703 limit. As discussed in the application, changing the safety limit MCPR methodology to one based on a 95 percent probability with 95 percent confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9 percent of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, Docket Nos. STN 50–456 and STN 50– 457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50–454 and STN 50– 455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50–461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–010, 50–237, and 50– 249, Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–373 and 50–374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–352 and 50–353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50–171, 50–277, and 50–278, Peach Bottom Atomic Power Station, Units 1, 2, and 3, York and Lancaster Counties, Pennsylvania Exelon Generation Company, LLC, Docket Nos. 50–254 and 50–265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: March 1, 2019. Publicly-available version is in ADAMS under Accession No. ML19063A685. Description of amendment request: The amendments would revise the E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES emergency action levels (EALs) in the emergency plan for each site. The proposed changes are based primarily on the resolution of emergency preparedness frequently asked questions (EPFAQs) and industry best-practices. Editorial changes are also proposed. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes involving revisions to existing NRC-approved [Nuclear Energy Institute guidance document] NEI 99–01, Revision 6, EALs, as clarified by the NRC through the EPFAQ process, for the affected facilities do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of Exelon’s ERO [emergency response organization] to respond in mitigating the consequences of any design basis accident. The probability of a reactor accident requiring implementation of Emergency Plan EALs has no relevance in determining whether the proposed changes to the EALs reduce the effectiveness of the Emergency Plans. As discussed in Section D, ‘‘Planning Basis,’’ of NUREG–0654, Revision 1, ‘‘Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants;’’ ‘‘. . . The overall objective of emergency response plans is to provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that could produce offsite doses in excess of Protective Action Guides (PAGs). No single specific accident sequence should be isolated as the one for which to plan because each accident could have different consequences, both in nature and degree. Further, the range of possible selection for a planning basis is very large, starting with a zero point of requiring no planning at all because significant offsite radiological accident consequences are unlikely to occur, to planning for the worst possible accident, regardless of its extremely low likelihood. . . .’’ Therefore, Exelon did not consider the risk insights regarding any specific accident initiation or progression in evaluating the proposed changes. The proposed changes do not involve any physical changes to plant equipment or systems, nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors nor do they alter the design assumptions, conditions, and configuration or the manner in which the plants are operated and maintained. The proposed VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 changes do not adversely affect the ability of Structures, Systems, or Components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes involving revisions to existing NRC-approved NEI 99–01, Revision 6, EALs, as clarified by the NRC through the EPFAQ process, for the affected facilities do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. Exelon ERO functions will continue to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those that have been previously evaluated. 3. Do the proposed amendments involve a significant reduction in a margin of safety? Response: No. The proposed changes involving revisions to existing NRC-approved NEI 99–01, Revision 6, EALs, as clarified by the NRC through the EPFAQ process, for the affected facilities do not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated. Margins of safety are unaffected by the proposed changes to the EALs based on further NRC clarification through the EPFAQ. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met. Therefore, the proposed changes do not involve any reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the 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 PO 00000 Frm 00065 Fmt 4703 Sfmt 4703 14147 Exelon Generation Company, LLC, Docket Nos. 50–237 and 50–249, Dresden Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois, and Docket Nos. 50–254 and 50–265, Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, Rock Island County, Illinois Date of amendment request: December 5, 2018. A publicly-available version is in ADAMS under Accession No. ML18339A009. Description of amendment request: The amendments would revise the technical specifications for both the single recirculation loop and two recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR) limits for the DNPS and QCNPS units. The proposed decrease in these limits improves operational flexibility through the recapture of margins that are available as a result of the transition to Framatome, Inc. using NRC-approved SLMCPR calculation methodology. 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 SLMCPR values have been determined using NRC-approved methods discussed in AREVA Topical Report ANP– 10307PA, Revision 0, ‘‘AREVA MCPR Safety Limit Methodology for Boiling Water Reactors,’’ dated June 2011. The proposed SLMCPRs for two recirculation loop and single recirculation loop operation ensure that the acceptance criterion continues to be met (i.e., at least 99.9 percent of all fuel rods in the core do not experience boiling transition). The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed license amendments do not involve any plant modifications or operational changes that could affect system reliability or performance, or that could affect the probability of operator error. As such, the proposed changes do not affect any postulated accident precursors. Since no individual precursors of an accident are affected, the proposed license amendments do not involve a significant increase in the probability of a previously analyzed event. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. The basis for the SLMCPR calculations is to ensure that during normal operation and during anticipated operational occurrences, at least 99.9 percent of all fuel E:\FR\FM\09APN1.SGM 09APN1 14148 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES rods in the core do not experience boiling transition if the safety limit is not exceeded. Based on these considerations, the proposed changes do not involve a significant increase in the consequences of a previously analyzed accident. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The SLMCPR is a TS numerical value calculated for two recirculation loop operation and single recirculation loop operation to ensure at least 99.9 percent of all fuel rods in the core do not experience boiling transition if the safety limit is not exceeded. SLMCPR values are calculated using NRC-approved methodology identified in the TSs. The proposed SLMCPR values do not involve any new modes of plant operation or any plant modifications and do not directly or indirectly affect the failure modes of any plant systems or components. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The SLMCPR provides a margin of safety by ensuring that at least 99.9 percent of the fuel rods do not experience boiling transition during normal operation and anticipated operational occurrences if the MCPR Safety Limit is not exceeded. Revision of the SLMCPR values in TS 2.1.1.2, using an NRCapproved methodology, will ensure that the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR Safety Limit is not exceeded). The SLMCPRs are verified to be bounding by cycle specific analyses prior to power operations for each operating cycle. Therefore, the proposed amendments do not result in a significant reduction in the margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorneys for licensee: Tamra (Tami) Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555 NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, Docket No. 50–244, R. E. Ginna Nuclear Power Plant, Wayne County, New York Date of amendment request: February 15, 2018. A publicly-available version is VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 in ADAMS under Accession No. ML19045A282. Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.15, ‘‘Containment Leakage Rate Testing Program,’’ to reflect an increase to the existing Type A integrated leak rate test program test interval from 10 years to 15 years, in accordance with Nuclear Energy Institute (NEI) Report NEI 94–01, Revision 2–A, ‘‘Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J.’’ The proposed change would also reflect adoption of both the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8–2002, ‘‘Containment System Leakage Testing Requirements,’’ and a more conservative allowable test interval extension of 9 months for Type A leakage tests in accordance with NEI 94–01, Revision 2– A. The amendment would also make an administrative change to remove the exception under TS 5.5.15 for the onetime 15-year Type A test internal being performed after May 31, 1996, and performed prior to May 31, 2011, as this has already occurred. 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 activity involves the revision of the R. E. Ginna Nuclear Power Plant (GNPP) Technical Specification (TS) 5.5.15, ‘‘Primary Containment Leakage Rate Testing Program,’’ to allow the extension of the Type A Integrated Leakage Rate Test (ILRT) containment test interval to 15 years. Per the guidance provided in Nuclear Energy Institute (NEI) 94–01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 2–A, the current Type A test interval of 10 years would be extended on a permanent basis to no longer than 15 years from the last Type A test. The proposed interval extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant’s ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. PO 00000 Frm 00066 Fmt 4703 Sfmt 4703 The change in Type A test frequency to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the probabilistic risk assessment (PRA) is 0.29 person-Roentgen equivalent man (rem)/year. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2A states that a very small population dose is defined as an increase of less than 1.0 person-rem per year or less than 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation which endorsed NEI 94–01 and EPRI Report No. 1009325, Revision 2A. Moreover, the risk impact when compared to other severe accident risks is negligible. Therefore, the proposed extension does not involve a significant increase in the probability of an accident previously evaluated. In addition, as documented in NUREG– 1493, ‘‘Performance-Based Containment Leak-Test Program,’’ dated September 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The GNPP Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, ‘‘Rules for Inservice Inspection of Nuclear Power Plant Components,’’ Containment Maintenance Rule Inspections, Containment Coatings Program and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test (ILRT). Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated. This proposed amendment also deletes the exception previously granted to allow onetime extension of the ILRT test frequency for GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires the first Type A test performed after May 31, 1996, to be performed by May 31, 2011. This exception was included in the TS for one-time testing activities that would have already taken place by the time this amendment is approved; therefore, deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. E:\FR\FM\09APN1.SGM 09APN1 amozie on DSK9F9SC42PROD with NOTICES Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices Therefore, the proposed changes do not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the GNPP TS 5.5.15 involves the extension of the GNPP Type A containment test interval from 10 years to 15 years. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant’s ability to mitigate the consequences of an accident; thereby, do not involve any accident precursors or initiators. The proposed change does not involve a physical modification to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment. This proposed amendment also deletes the exception previously granted to allow onetime extension of the ILRT test frequency for GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires the first Type A test performed after May 31, 1996, to be performed by May 31, 2011. This exception was included in the TS for one-time testing activities that would have already taken place by the time this amendment is approved; therefore, deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.15 involves the extension of the GNPP Type A containment test interval to 15 years. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves the extension of the interval between Type A containment leak rate tests for GNPP. The proposed surveillance interval extension is bounded by the 15-year ILRT interval currently authorized within NEI 94–01, Revision 2–A. Industry experience supports the conclusion that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with Option B to 10 CFR 50, VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 Appendix J and the overlapping inspection activities performed as part of ASME Section Xl, and the TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A test intervals. In addition, this proposed amendment also deletes the exception previously granted to allow one-time extension of the ILRT test frequency for GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires the first Type A test performed after May 31, 1996, to be performed by May 31, 2011. This exception was included in the TS for one-time testing activities that would have already taken place by the time this amendment is approved; therefore, deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not 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. Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: James G. Danna. FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50–346, Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), Ottawa County, Ohio Date of amendment request: February 5, 2019. A publicly-available version is in ADAMS under Accession No. ML19036A523. Description of amendment request: By letter dated April 25, 2018 (ADAMS Accession No. ML18115A007), FENOC notified the NRC that DBNPS will permanently cease power operations by May 31, 2020. The proposed amendment would revise the DBNPS renewed facility operating license (RFOL) and technical specifications (TSs) following the permanent cessation PO 00000 Frm 00067 Fmt 4703 Sfmt 4703 14149 of power operations to reflect the postshutdown and permanently defueled condition. The proposed amendment would eliminate TS requirements and license conditions which would not be applicable once DBNPS ceases power operations and can no longer place fuel in the reactor vessel. The proposed amendment would also eliminate obsolete license conditions. In addition, the proposed amendment would revise several license conditions and TS requirements, including limiting conditions for operation (LCOs), usage rules, definitions, surveillance requirements (SRs), and administrative controls. FENOC also proposed to revise the licensing bases for DBNPS, including the design bases accident (DBA) analysis. 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 would not take effect until DBNPS has certified to the NRC that it has permanently ceased operation and entered a permanently defueled condition. Because the 10 CFR part 50 license for DBNPS will no longer authorize operation of the reactor, or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The remaining [Updated Final Safety Analysis Report] UFSAR Chapter 15 postulated design basis accident (DBA) events that could potentially occur at a permanently defueled facility would be a fuel handling accident (FHA) in the spent fuel pool (SFP), the waste gas decay tank rupture (WGDTR), and external causes. The FHA analyses for DBNPS shows that, following 95 days of decay time after reactor shutdown and provided the SFP water level requirements of TS LCO 3.7.14 are met, the dose consequences are acceptable without relying on structures, systems, and components (SSCs) to remain functional for accident mitigation during and following the event other than the passive SFP structure. The remaining DBAs that support the permanently shutdown and defueled condition do not rely on any active safety systems for mitigation. The probability of occurrence of previously evaluated accidents is not increased, since safe storage and handling of fuel will be the only operations performed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation E:\FR\FM\09APN1.SGM 09APN1 amozie on DSK9F9SC42PROD with NOTICES 14150 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices will no longer be credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents. The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled status of DBNPS has no impact on the remaining applicable DBAs. The removal of LCOs or SRs that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the permanently defueled condition. 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 delete or modify certain DBNPS RFOL, TS, and current licensing bases (CLB) have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor, or only to the prevention, diagnosis, or mitigation of reactor related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled. The proposed modification or deletion of requirements of the DBNPS RFOL, TS, and CLB do not affect systems credited in the accident analysis for the remaining credible DBAs at DBNPS. The proposed RFOL and PDTS [permanently defueled TSs] will continue to require proper control and monitoring of safety significant parameters and activities. The TS regarding SFP water level and spent fuel storage is retained to preserve the current requirements for safe storage of irradiated fuel. The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding, spent fuel racks, SFP integrity, and SFP water level). Since extended operation in a defueled condition and safe fuel handling will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes are to delete or modify certain RFOL, TS, and CLB once the VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 DBNPS facility has been permanently shutdown and defueled. Because the 10 CFR part 50 license for DBNPS will no longer authorize operation of the reactor, or emplacement or retention of fuel into the reactor vessel, the occurrence of postulated accidents associated with reactor operation is no longer credible. The remaining postulated DBA events that could potentially occur at a permanently defueled facility would be a[n] FHA, WGDTR, and external causes. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses. The proposed changes are limited to those portions of the RFOL, TS, and CLB that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the RFOL, TS, and CLB are not credited in the updated applicable accident analysis for the remaining applicable postulated accidents, and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated design basis accidents involving the reactor will no longer be possible because the reactor will be permanently shutdown and defueled, and DBNPS will no longer be authorized to operate the reactor. 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: Rick Giannantonio, General Counsel, FirstEnergy Corporation, Mail Stop A– GO–15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: David J. Wrona. Indiana Michigan Power Company, Docket Nos. 50–315 and 50–316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: February 26, 2019. A publicly-available version is in ADAMS under Accession No. ML19060A060. Description of amendment request: The proposed amendment would expand the criteria within technical specification (TS) 3.2.1 surveillance requirements to apply a revised penalty factor to measured transient FQ(Z) in response to Westinghouse Nuclear Safety Advisory Letter, NSAL–15–1, ‘‘Heat Flux Hot Channel Factor Technical Specification Surveillance.’’ 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 PO 00000 Frm 00068 Fmt 4703 Sfmt 4703 consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated? Response: No. The proposed amendment to add an additional surveillance requirement, to apply the penalty factor of 1.02 or a factor specified in the COLR [core operating limit report], whichever is greater, to the transient FQ(Z) calculation, ensures that the assumptions and inputs to the safety analyses remain valid and does not result in actions that would increase the probability or consequences of any accident previously evaluated. The design of the protection systems will be unaffected. The reactor protection system and engineered safety feature actuation system will continue to function in a manner consistent with the plant design basis. All design, material and construction standards that were applicable prior to the request are maintained. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accidentpreviously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Operation in accordance with the revised TS and its limits precludes new challenges to systems or structures that might introduce a new type of accident. All design and performance criteria will continue to be met and no new single failure mechanisms will be created. The proposed change for resolution of Westinghouse NSAL–15–1 does not involve the alteration of plant equipment or introduce unique operational modes or accident precursors. Therefore it does not create the potential for a different kind of accident. Therefore, the proposed changes do not create the possibility of a new or, different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Operation in accordance with the revised TS and its limits preserves the margins assumed in the safety analyses. This ensures that all design and performance criteria associated with the safety analysis will continue to be met and that the margin of safety is not affected. 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 requests involve no significant hazards consideration. Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices NRC Branch Chief: David J. Wrona. amozie on DSK9F9SC42PROD with NOTICES Indiana Michigan Power Company, Docket Nos. 50–315 and 50–316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: February 26, 2019. A publicly-available version is in ADAMS under Accession No. ML19063A498. Description of amendment request: The proposed amendment would adopt Technical Specification Task Force (TSTF) Traveler TSTF–563, ‘‘Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program.’’ TSTF–563 revises the TS definitions of Channel Calibration, Channel Operational Test, and Trip Actuating Device Operational Test. 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 of occurrence or consequences of an accident previously evaluated? Response: No. The proposed change revises the TS [technical specification] definitions of Channel Calibration, COT [channel operational test], and TADOT [trip actuating device operational test] to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. All components in the channel continue to be tested. The frequency at which a channel test is performed is not an initiator of any accident previously evaluated, so the probability of an accident is not affected by the proposed change. The channels surveilled in accordance with the affected definitions continue to be required to be operable and the acceptance criteria of the surveillances are unchanged. As a result, any mitigating functions assumed in the accident analysis will continue to be performed. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accidentpreviously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises the TS definitions of Channel Calibration, COT, and TADOT to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. The design function or operation of the components involved are not affected and there is no physical alteration of the plant (i.e., no new or different type of equipment will be installed). No credible new failure VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions. Therefore, the proposed changes do not create the possibility of a new or, different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises the TS definitions of Channel Calibration, COT, and TADOT to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. The Surveillance Frequency Control Program assures sufficient safety margins are maintained, and that design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants’ licensing basis. The proposed change does not adversely affect existing plant safety margins, or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the method of determining surveillance test intervals under an NRCapproved licensee-controlled program. 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 requests involve no significant hazards consideration. Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: David J. Wrona. NextEra Energy Seabrook, LLC, Docket No. 50–443, Seabrook Station, Unit No. 1, Rockingham wCounty, New Hampshire Date of amendment request: October 4, 2018. A publicly-available version is in ADAMS under Accession No. ML18277A377. Description of amendment request: The amendment would revise the Seabrook Station, Unit No. 1 (Seabrook), Technical Specifications (TSs) and Surveillance Requirements (SRs) associated with the control rods. The amendment would adopt changes provided in Technical Specifications Task Force (TSTF) Traveler TSTF–234, ‘‘Add Action for More than One [D]RPI PO 00000 Frm 00069 Fmt 4703 Sfmt 4703 14151 [Digital Rod Position Indicator] Inoperable,’’ and TSTF–547, ‘‘Clarification of Rod Position Requirements,’’ and make various other changes to align the Seabrook TSs more closely with NUREG–1431, ‘‘Standard Technical Specifications— Westinghouse Plants.’’ In all, the amendment would revise SR 4.1.1.1.1, SR 4.1.1.2, TS 3.1.3.1, SR 4.1.3.1.1, TS 3.1.3.2, SR 4.1.3.2, TS 3.1.3.3, SR 4.1.3.3, TS 3.1.3.5, SR 4.1.3.5, TS 3.1.3.6, SR 4.1.3.6, TS 3.10.5, SR 4.10.5, and TS 6.8.1.6.b. 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. Control and shutdown rods are assumed to insert into the core to shut down the reactor in evaluated accidents. Rod insertion limits ensure that adequate negative reactivity is available to provide the assumed shutdown margin (SDM). Rod alignment limits maintain an appropriate power distribution and reactivity insertion profile. Control and shutdown rods are initiators to several accidents previously evaluated, such as rod ejection. The proposed change does not change the limiting conditions for operation for the rods or make any technical changes to the surveillance requirements governing the rods. Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated. Adding new TS Actions to provide a limited time to repair rod control system failures has no effect on the SDM assumed in the accident analysis as the proposed Actions require verification that SDM is maintained. The effects on power distribution will not cause a significant increase in the consequences of any accident previously evaluated as all TS requirements on power distribution continue to be applicable. The proposed change to resolve the conflicts in the TS ensures that the intended Actions are followed when equipment is inoperable. Actions taken with inoperable equipment are not assumptions in the accidents previously evaluated and have no significant effect on the consequences. The capability of any operable TS-required equipment to perform its specified safety function is not impacted by the proposed change. As a result, the outcomes of accidents previously evaluated are unaffected. Therefore, the proposed changes do not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? E:\FR\FM\09APN1.SGM 09APN1 14152 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices Response: No. The proposed change does not challenge the integrity or performance of any safetyrelated 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 system or component. No physical changes are made to the plant, so no new causal mechanisms are introduced. Therefore, the proposed changes to the TS 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 ability of the control rods to perform their designated safety function is unaffected by the proposed changes. The proposed changes do not alter any safety analyses assumptions, safety limits, limiting safety system settings, or method of operating the plant. The proposed change to provide time to repair rods that are operable but immovable does not result in a significant reduction in the margin of safety because all rods must be verified to be operable, and all other banks must be within the insertion limits. The changes do not adversely affect plant operating margins or the reliability of equipment credited in the safety analyses. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. amozie on DSK9F9SC42PROD with NOTICES The NRC staff has reviewed the licensee’s analysis, and based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Debbie Hendell, Managing Attorney, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408–0420. NRC Branch Chief: James G. Danna. PSEG Nuclear LLC, Docket No. 50–354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: February 27, 2019. A publicly-available version is in ADAMS under Accession No. ML19058A221. Description of amendment request: The proposed change is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF–546, Revision 0, ‘‘Revise APRM [Average Power Range Monitor] Channel Adjustment Surveillance Requirement’’ (ADAMS Accession No. ML17205A444). The amendment would alter Surveillance Requirement (SR) 4.3.1.1 of Technical Specification 3.3.1, ‘‘Reactor Protection System Instrumentation.’’ The change would revise the SR to verify that calculated (i.e., calorimetric heat balance) power is no more than 2 percent greater than the APRM channel output. The SR requires the APRM VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 channel to be adjusted such that calculated power is no more than 2 percent greater than the APRM indicated power when operating at ≥24 percent of rated thermal power. This change would revise the SR to distinguish between APRM indications that are consistent with the accident analyses and those that provide additional margin. 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 APRM system and the RPS are not initiators of any accidents previously evaluated. As a result, the proposed change does not affect the probability of any accident previously evaluated. The APRM system and the Reactor Protection System (RPS) functions act to mitigate the consequences of accidents previously evaluated. The reliability of APRM system and the RPS is not significantly affected by removing the gain adjustment requirement on the APRM channels when the APRMs are calibrated conservatively with respect to the calculated heat balance. This is because the actual core thermal power at which the reactor will automatically trip is lower, thereby increasing the margin to the core thermal limits and the limiting safety system settings assumed in the safety analyses. The consequences of an accident during the adjustment of the APRM instrumentation are no different from those during the existing surveillance testing period or the existing time allowed to restore the instruments to operable status. As a result, the ability of the APRM system and the RPS to mitigate any accident previously evaluated is not significantly affected. 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. The proposed change does not alter the protection system design, create new failure modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. PO 00000 Frm 00070 Fmt 4703 Sfmt 4703 The margin of safety provided by the APRM system and the RPS is to ensure that the reactor is shut down automatically when plant parameters exceed the setpoints for the system. Any reduction in the margin of safety resulting from the adjustment of the APRM channels while continuing operation is considered to be offset by delaying a plant shutdown (i.e., a transient) for a short time with the APRM system, the primary indication of core power and an input to the RPS, not calibrated. Additionally, the short time period required for adjustment is consistent with the time allowed by Technical Specifications to restore the core power distribution parameters to within limits and is acceptable based on the low probability of a transient or design basis accident occurring simultaneously with inaccurate APRM channels. The proposed change does not alter setpoints or limits established or assumed by the accident analyses. The Technical Specifications continue to require operability of the RPS functions, which provide core protection for postulated reactivity insertion events occurring during power operating conditions consistent with the plant safety analyses. 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: Steven Fleischer, PSEG Services Corporation, 80 Park Plaza, T–5, Newark, NJ 07102. NRC Branch Chief: James G. Danna. Tennessee Valley Authority (TVA), Docket Nos. 50–390 and 50–391, Watts Bar Nuclear Plant, Units 1 and 2, Rhea County, Tennessee Date of amendment request: October 12, 2018. A publicly-available version is in ADAMS under Accession No. ML18288A352. Description of amendment request: The amendments would revise the Technical Specifications (TS) by the adoption, with administrative and technical variations, of Technical Specification Task Force (TSTF) Traveler TSTF–425, Revision 3, ‘‘Relocate Surveillance Frequencies to Licensee Control—Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b.’’ TSTF–425, Revision 3, provides for the relocation of specific surveillance frequencies to a licensee-controlled program. Additionally, the change would add a new program, the Surveillance Frequency Control Program (SFCP), to TS Section 5.0, ‘‘Administrative Controls.’’ E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices amozie on DSK9F9SC42PROD with NOTICES 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 relocates the specified frequencies for periodic surveillance requirements to licensee control under a new 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 for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new or different accidents result from utilizing the proposed change. The change does 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 change does not impose any new or different requirements. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The design, operation, testing methods, and acceptance criteria for structures, systems, [and] components, 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), because these are not affected by changes to the surveillance frequencies. Similarly, there is no effect to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, TVA will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 [Nuclear Energy Institute] 04–10, Revision 1, in accordance with the TS SFCP. This 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 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 Branch Chief: Undine Shoop. Tennessee Valley Authority, Docket Nos. 50–327 and 50–328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date of amendment request: February 1, 2019. A publicly-available version is in ADAMS under Accession No. ML19032A632. Description of amendment request: The amendments would adopt Technical Specification Task Force Traveler TSTF–563, ‘‘Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program.’’ 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 revises the TS [Technical Specification] definitions of Channel Calibration, COT [Channel Operational Test], and TADOT [Trip Actuation Device Operational Test] to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. All components in the channel continue to be tested. The frequency at which a channel test is performed is not an initiator of any accident previously evaluated, so the probability of an accident is not affected by the proposed change. The channels surveilled in accordance with the affected definitions continue to be required to be operable and the acceptance criteria of the surveillances are unchanged. As a result, any mitigating functions assumed in the accident analysis will continue to be performed. PO 00000 Frm 00071 Fmt 4703 Sfmt 4703 14153 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises the TS definitions of Channel Calibration, COT, and TADOT to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. The design function or operation of the components involved are not affected and there is no physical alteration of the plant (i.e., no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises the TS definitions of Channel Calibration, COT, and TADOT to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program. The Surveillance Frequency Control Program assures sufficient safety margins are maintained, and that design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the Nuclear Regulatory Commission (NRC)) will continue to be met as described in the plants’ licensing basis. The proposed change does not adversely affect existing plant safety margins, or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by method of determining surveillance test intervals under an NRCapproved licensee-controlled program. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. E:\FR\FM\09APN1.SGM 09APN1 14154 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices NRC Branch Chief: Undine Shoop. Wolf Creek Nuclear Operating Corporation, Docket No. 50–482, Wolf Creek Generating Station, Unit 1, Coffey County, Kansas amozie on DSK9F9SC42PROD with NOTICES Date of amendment request: January 23, 2019, as supplemented by letter dated March 11, 2019. Publiclyavailable versions are in ADAMS under Accession Nos. ML19036A772 and ML19078A131, respectively.) Description of amendment request: The amendment would revise technical specification (TS) requirements in Section 1.3, ‘‘Completion Times,’’ and Section 3.0, ‘‘Limiting Condition for Operation (LCO) Applicability,’’ regarding LCO and surveillance requirement (SR) usage. The proposed changes are consistent with the NRCapproved Technical Specifications Task Force (TSTF) Traveler TSTF–529, Revision 4, ‘‘Clarify Use and Application Rules,’’ using the consolidated line item improvement process (ADAMS Accession No. ML16062A271). The model safety evaluation was approved by the NRC in a letter dated April 21, 2016 (ADAMS Package Accession No. ML16060A441). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to Section 1.3 and LCO 3.0.4 have no effect on the requirement for systems to be Operable and have no effect on the application of TS actions. The proposed change to SR 3.0.3 states that the allowance may only be used when there is a reasonable expectation the surveillance will be met when performed. Since the proposed change does not significantly affect system Operability, the proposed change will have no significant effect on the initiating events for accidents previously evaluated and will have no significant effect on the ability of the systems to mitigate accidents previously evaluated. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the TS usage rules does not affect the design or function of any plant systems. The proposed change does not change the Operability requirements for plant VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 systems or the actions taken when plant systems are not operable. Therefore, it is concluded that this 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 clarifies the application of Section 1.3 and LCO 3.0.4 and does not result in changes in plant operation. SR 3.0.3 is revised to allow application of SR 3.0.3 when an SR has not been previously performed if there is reasonable expectation that the SR will be met when performed. This expands the use of SR 3.0.3 while ensuring the affected system is capable of performing its safety function. As a result, plant safety is either improved or unaffected. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 1200 17th Street NW, Washington, DC 20036. NRC Branch Chief: Robert J. Pascarelli. IV. 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 PO 00000 Frm 00072 Fmt 4703 Sfmt 4703 impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ‘‘Obtaining Information and Submitting Comments’’ section of this document. Dominion Energy Nuclear Connecticut, Inc., Docket No. 50–423, Millstone Power Station, Unit No. 3, New London County, Connecticut Date of amendment request: April 4, 2018, as supplemented by letter dated October 22, 2018. Brief description of amendment: The amendment revised ACTION 18 in Technical Specifications Table 3.3–3, Functional Unit 7.e, ‘‘Control Building Inlet Ventilation Radiation,’’ for Millstone Power Station, Unit No. 3, to allow continued fuel handling and reactor operation with inoperable inlet radiation monitoring instrumentation provided that one train of the control room emergency ventilation system is operating in the emergency mode. The technical specification change specifies that one train of the control room emergency ventilation system be placed in the emergency mode of operation within 7 days if one radiation monitor channel is inoperable, or immediately, if both radiation monitor channels are inoperable. Date of issuance: March 21, 2019. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 272. A publiclyavailable version is in ADAMS under Accession No. ML19042A277; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF–49: The amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 17, 2018 (83 FR 33266). The supplemental letter dated October 22, 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 E:\FR\FM\09APN1.SGM 09APN1 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 21, 2019. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket Nos. 50–003 and 50–247, Indian Point Nuclear Generating Unit Nos. 1 and 2 (Indian Point 1 and Indian Point 2), Westchester County, New York Date of amendment request: June 20, 2018. A publicly-available version is in ADAMS under Accession No. ML18179A173. Brief description of amendments: The amendments deleted certain license conditions from the Indian Point 1 and Indian Point 2 Operating Licenses that impose specific requirements on the decommissioning trust agreement. With approval of these amendments, the provisions of 10 CFR 50.75(h), which specify the regulatory requirements for decommissioning trust funds, apply to the licensee, Entergy Nuclear Operations, Inc., for Indian Point 1 and Indian Point 2. Date of issuance: March 21, 2019. Effective date: As of the date of issuance, and shall be implemented within 60 days of issuance. Amendment Nos.: 61 (Unit No. 1) and 289 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML19065A101; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Provisional Operating License No. DPR–5 and Renewed Facility Operating License No. DPR–26: The amendments revised the Operating Licenses. Date of initial notice in Federal Register: September 11, 2018 (83 FR 45984). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 21, 2019. No significant hazards consideration comments received: No. amozie on DSK9F9SC42PROD with NOTICES Entergy Operations, Inc., System Energy Resources, Inc., Cooperative Energy, A Mississippi Electric Cooperative, and Entergy Mississippi, LLC, Docket No. 50–416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: March 26, 2018. Brief description of amendment: The amendment revised the Updated Final Safety Analysis Report descriptions for the replacement of the Turbine First Stage Pressure output signals with Power Range Neutron Monitoring System output signals. VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 Date of issuance: March 12, 2019. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No: 217. A publiclyavailable version is in ADAMS under Accession No. ML18215A196; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF–29: The amendment revised the Updated Final Safety Analysis Report. Date of initial notice in Federal Register: June 5, 2018 (83 FR 26115). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 2019. No significant hazards consideration comments received: No. Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative Energy, A Mississippi Electric Cooperative; and Entergy Mississippi, LLC, Docket No. 50–416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: April 27, 2018, as supplemented by letter dated October 10, 2018. Brief description of amendment: The amendment revised the GGNS Emergency Plan to adopt an Emergency Action Level scheme based on Nuclear Energy Institute (NEI) guidance in NEI 99–01, Revision 6, ‘‘Development of Emergency Action Levels for NonPassive Reactors,’’ dated November 2012, which was endorsed by the NRC by letter dated March 28, 2013. Date of issuance: March 12, 2019. Effective date: As of the date of issuance and shall be implemented within 365 days of issuance. Amendment No: 216. A publiclyavailable version is in ADAMS under Accession No. ML19025A023; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF–29: The amendment revised the GGNS Emergency Plan. Date of initial notice in Federal Register: June 5, 2018 (83 FR 26104). The supplemental letter dated October 10, 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. PO 00000 Frm 00073 Fmt 4703 Sfmt 4703 14155 The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 2019. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. STN 50–456 and STN 50– 457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50–454 and STN 50– 455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50–461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–010, 50–237, and 50– 249, Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–373 and 50–374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50–254 and 50–265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: January 31, 2018, as supplemented by letters dated July 27 and November 29, 2018. Brief description of amendments: The amendments revise the emergency response organization positions identified in the emergency plan for each site. Date of issuance: March 21, 2019. Effective date: As of the date of issuance and shall be implemented on or before December 31, 2019. Amendment Nos.: Braidwood 201/ 201, Byron 206/206, Clinton 223, Dresden 46/261/254, LaSalle 236/222, and Quad Cities 274/269. A publiclyavailable version is in ADAMS under Accession No. ML19036A586. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF– 72, NPF–77, NPF–37, NPF–66, NPF–62, DPR–2, DPR–19, DPR–25, NPF–11, NPF– 18, DPR–29, and DPR–30: Amendments revised the emergency plans. Date of initial notice in Federal Register: April 10, 2018 (83 FR 15417). The Commission’s related evaluation of the amendments is contained in a safety evaluation dated March 21, 2019. No significant hazards consideration comments received: No. E:\FR\FM\09APN1.SGM 09APN1 14156 Federal Register / Vol. 84, No. 68 / Tuesday, April 9, 2019 / Notices Florida Power & Light Company, Docket Nos. 50–250 and 50–251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: May 14, 2018, as supplemented by letter dated November 20, 2018. Brief description of amendments: The amendments revised the Technical Specifications (TSs) to increase the minimum load required for the Emergency Diesel Generator (EDG) partial-load rejection Surveillance Requirement (SR). Additionally, the amendments modified the EDG voltage and frequency limits for the SR and established a recovery period for the EDG(s) to return to steady-state conditions. Date of issuance: March 18, 2019. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: Unit 1, 285 and Unit 2, 279. A publicly-available version is in ADAMS under Accession No. ML18354A673; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR–31 and DPR–41: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: July 3, 2018 (83 FR 31185). The supplemental letter dated November 20, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 18, 2019. No significant hazards consideration comments received: No. amozie on DSK9F9SC42PROD with NOTICES Southern Nuclear Operating Company, Docket No. 52–025, Vogtle Electric Generating Plant (VEGP), Unit 3, Burke County, Georgia Date of amendment request: October 19, 2018. Description of amendment: The amendment authorizes the Southern Nuclear Operating Company to depart from certified AP1000 Design Control Document (DCD) Tier 2* material that has been incorporated into the Updated Final Safety Analysis Report (UFSAR). Specifically, the proposed departure consists of changes to Tier 2* information in the UFSAR (which includes the plant-specific DCD VerDate Sep<11>2014 18:15 Apr 08, 2019 Jkt 247001 information) to change the vertical reinforcement information provided in the VEGP Unit 3 column line 1 wall from elevation 135′-3″ to 137′-0″ . Date of issuance: March 13, 2019. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 156 for Unit 3. Publicly-available versions are in an ADAMS package under Accession No. ML19044A500 which includes the Safety Evaluation that references documents, located in that ADAMS package, related to this amendment. Facility Combined Licenses No. NPF– 91: Amendment revised the Facility Combined License. Date of initial notice in Federal Register: November 20, 2018 (83 FR 58607). The Commission’s related evaluation of the amendment is contained in the Safety Evaluation dated March 13, 2019. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50–391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee Date of amendment request: March 5, 2018, as supplemented by letters dated April 27 and October 11, 2018. Brief description of amendment: The amendment revised License Condition 2.C.(4), concerning the use of the PAD4TCD computer program. While the current License Condition permits the use of PAD4TCD for Unit 2, Cycles 1 and 2 only, the revision allows the use of PAD4TCD until the Unit 2 steam generators (SGs) are replaced with SGs equivalent to the existing SGs at Unit 1. Date of issuance: March 20, 2019. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 26. A publiclyavailable version is in ADAMS under Accession No. ML19046A286; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF– 96: Amendment revised the Facility Operating License. Date of initial notice in Federal Register: December 4, 2018 (83 FR 62623). The supplemental letters dated April 27 and October 11, 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 March 20, 2019. PO 00000 Frm 00074 Fmt 4703 Sfmt 9990 No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket Nos. 50–390 and 50–391, Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2, Rhea County, Tennessee Date of amendment request: August 1, 2018, as supplemented by letter dated March 4, 2019. Brief description of amendments: The amendments revised the Technical Specifications (TS) to adopt, with minor variation, Technical Specification Task Force (TSTF) Traveler TSTF–266–A, Revision 3, ‘‘Eliminate the Remote Shutdown System Table of Instrumentation and Controls.’’ Specifically, the comparable TS Table 3.3.4–1, ‘‘Remote Shutdown System Instrumentation and Controls,’’ was deleted from Watts Bar, Units 1 and 2, TS 3.3.4, ‘‘Remote Shutdown System.’’ Date of issuance: March 18, 2019. Effective date: As of the date of issuance and shall be implemented by March 24, 2019. Amendment Nos.: 124 and 25. A publicly-available version is in ADAMS under Accession No. ML19066A009; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF– 90 and NPF–96: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: February 12, 2019 (84 FR 3510). The supplemental letter dated March 4, 2019, requested expedited completion of the NRC review of 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 (NSHC) determination as published in the Federal Register. The Commission’s related evaluation of the amendments and final NSHC determination are contained in a Safety Evaluation dated March 18, 2019. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 29th day of March 2019. For the Nuclear Regulatory Commission. Craig G. Erlanger, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2019–06449 Filed 4–8–19; 8:45 am] BILLING CODE 7590–01–P E:\FR\FM\09APN1.SGM 09APN1

Agencies

[Federal Register Volume 84, Number 68 (Tuesday, April 9, 2019)]
[Notices]
[Pages 14142-14156]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-06449]


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NUCLEAR REGULATORY COMMISSION

[NRC-2019-0087]


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 the Atomic Energy Act of 1954, as amended (the 
Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this 
regular biweekly notice. The Act requires the Commission to publish 
notice of any amendments issued, or proposed to be issued, and grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license or combined license, as 
applicable, upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from March 12, 2019 to March 25, 2019. The last 
biweekly notice was published on March 26, 2019.

DATES: Comments must be filed by May 9, 2019. A request for a hearing 
must be filed by June 10, 2019.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0087. Address 
questions about NRC dockets IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual(s) listed in the FOR FURTHER 
INFORMATION CONTACT section of this document.
     Mail comments to: Office of Administration, Mail Stop: 
TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, ATTN: Program Management, Announcements and Editing Staff.
     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: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-5411, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2019-0087, 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-2019-0087.
     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 ``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-2019-0087, 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. Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the 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.

III. 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

[[Page 14143]]

not (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated; or (2) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (3) involve a significant reduction in a 
margin of safety. The basis for this proposed determination for each 
amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment 
period such that failure to act in a timely way would result, for 
example in derating or shutdown of the facility. If the Commission 
takes action prior to the expiration of either the comment period or 
the notice period, it will publish in the Federal Register a notice of 
issuance. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. The petition 
must be filed in accordance with the filing instructions in the 
``Electronic Submissions (E-Filing)'' section of this document, and 
should meet the requirements for petitions set forth in this section, 
except that under 10 CFR 2.309(h)(2) a State, local governmental body, 
or Federally-recognized Indian Tribe, or agency thereof does not need 
to address the standing requirements in 10 CFR 2.309(d) if the facility 
is located within its boundaries. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not

[[Page 14144]]

otherwise participate in the proceeding. A limited appearance may be 
made at any session of the hearing or at any prehearing conference, 
subject to the limits and conditions as may be imposed by the presiding 
officer. Details regarding the opportunity to make a limited appearance 
will be provided by the presiding officer if such sessions are 
scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562; August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public website at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
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 
application(s), 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.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: February 8, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19039A126.
    Description of amendment request: The amendment would adopt TSTF-
564, ``Safety Limit MCPR [minimum critical power ratio],'' Revision 2, 
which

[[Page 14145]]

revises the Fermi 2 technical specification safety limit on minimum 
critical power ratio (SLMCPR) to reduce the need for cycle-specific 
changes to the value while still meeting the regulatory requirement for 
a safety limit. In addition, technical specification 5.6.5, ``Core 
Operating Limits Report (COLR),'' is revised to require the current 
SLMCPR value to be included in the COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment revises the TS SLMCPR and the list of 
core operating limits to be included in the Core Operating Limits 
Report (COLR). The SLMCPR is not an initiator of any accident 
previously evaluated. The revised safety limit values continue to 
ensure for all accidents previously evaluated that the fuel cladding 
will be protected from failure due to transition boiling. The 
proposed change does not affect plant operation or any procedural or 
administrative controls on plant operation that affect the functions 
of preventing or mitigating any accidents previously evaluated.
    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 previously evaluated?
    Response: No.
    The proposed amendment revises the TS SLMCPR and the list of 
core operating limits to be included in the COLR. The proposed 
change will not affect the design function or operation of any 
structures, systems or components (SSCs). No new equipment will be 
installed. As a result, the proposed change will not create any 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered 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 amendment revises the TS SLMCPR and the list of 
core operating limits to be included in the COLR. This will result 
in a change to a safety limit, but will not result in a significant 
reduction in the margin of safety provided by the safety limit. As 
discussed in the application, changing the SLMCPR methodology to one 
based on a 95% probability with 95% confidence that no fuel rods 
experience transition boiling during an anticipated transient 
instead of the current limit based on ensuring that 99.9% of the 
fuel rods are not susceptible to boiling transition does not have a 
significant effect on plant response to any analyzed accident. The 
SLMCPR and the TS Llimiting Condition for Operation (LCO) on MCPR 
continue to provide the same level of assurance as the current 
limits and do not reduce a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jon P. Christinidis, DTE Energy, Expert 
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
    NRC Branch Chief: David J. Wrona.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

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

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: January 31, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19032A256.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) for each of these facilities based on 
Technical Specifications Task Force (TSTF) Traveler TSTF-529, ``Clarify 
Use and Application Rules,'' Revision 4 (ADAMS Accession No. 
ML16062A271). Specifically, the changes would revise and clarify the TS 
usage rules for completion times, limiting conditions for operation 
(LCOs), and surveillance requirements (SRs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Section 1.3 and LCO 3.0.4 have no effect 
on the requirement for systems to be Operable and have no effect on 
the application of TS actions. The proposed change to SR 3.0.3 (or 
equivalent) states that the allowance may only be used when there is 
a reasonable expectation the surveillance will be met when 
performed.
    Since the proposed changes do not significantly affect system 
Operability, the proposed changes will have no significant effect on 
the initiating events for accidents previously evaluated and will 
have no significant effect on the ability of the systems to mitigate 
accidents previously evaluated.
    Therefore, it is concluded that these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the TS usage rules do not affect the 
design or function of any plant systems. The proposed changes do not 
change the Operability requirements for plant systems or the actions 
taken when plant systems are not operable.
    Therefore, it is concluded that the changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes clarify the application of Section 1.3 and 
LCO 3.0.4 and do not result in changes in plant operation. SR 3.0.3 
(or equivalent) is revised to allow application of SR 3.0.3 when an 
SR has not been previously performed if there is reasonable 
expectation that the SR will be met when performed. This expands the 
use of SR 3.0.3 while ensuring the affected system is capable of 
performing its safety function. As a result, plant safety is either 
improved or unaffected.
    Therefore, it is concluded that 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, Entergy 
Services,

[[Page 14146]]

Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 
50-333, James A. FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego 
County, New York

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station (Limerick), Units 1 and 2, Montgomery County, 
Pennsylvania

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of amendment request: February 1, 2019, as supplemented by 
letter dated March 7, 2019. Publicly-available versions are in ADAMS 
under Accession Nos. ML19032A624 and ML19066A162, respectively.
    Description of amendment request: The proposed amendments would 
revise the technical specification (TS) requirements for these 
facilities related to the safety limit minimum critical power ratio 
(MCPR) and the core operating limits report (COLR). The proposed 
amendments are based on Technical Specification Task Force (TSTF) 
traveler TSTF-564, Revision 2, ``Safety Limit MCPR'' (ADAMS Accession 
No. ML18297A361). The proposed amendments for Limerick and FitzPatrick 
would also make changes to the MCPR and COLR requirements that are 
outside the scope of TSTF-564, Revision 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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:

    1. Do the proposed amendments involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendments revise the TS requirements for the 
safety limit MCPR and the list of core operating limits to be 
included in the COLR. The safety limit MCPR is not an initiator of 
any accident previously evaluated. The revised safety limit values 
will continue to ensure for all accidents previously evaluated that 
the fuel cladding will be protected from failure due to transition 
boiling. The proposed amendments for Limerick, Units 1 and 2, also 
include a revision to point to MCPR limits specified in the COLR and 
clarify references to other specifications. The proposed amendment 
for FitzPatrick also revises the COLR methodology references by 
deleting references that are no longer needed and clarifying the 
remaining reference. The proposed changes do not affect plant 
operation or any procedural or administrative controls on plant 
operation that affect the functions of preventing or mitigating any 
accidents previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed amendments create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendments revise the TS requirements for the 
safety limit MCPR and the list of core operating limits to be 
included in the COLR. The proposed amendments for Limerick, Units 1 
and 2, also include a revision to point to MCPR limits specified in 
the COLR and clarify references to other specifications. The 
proposed amendment for FitzPatrick also revises the COLR methodology 
references by deleting references that are no longer needed and 
clarifying the remaining reference. The proposed change will not 
affect the design function or operation of any structures, systems 
or components. No new equipment will be installed. As a result, the 
proposed changes will not create any credible new failure 
mechanisms, malfunctions, or accident initiators not considered in 
the design and licensing bases.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed amendments involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed amendments revise the TS safety limit MCPR and the 
list of core operating limits to be included in the COLR. The 
proposed amendments for Limerick, Units 1 and 2, also include a 
revision to point to MCPR limits specified in the COLR and clarify 
references to other specifications. The proposed amendment for 
FitzPatrick also revises the COLR methodology references by deleting 
references that are no longer needed and clarifying the remaining 
reference. This will result in a change to a safety limit, but will 
not result in a significant reduction in the margin of safety 
provided by the safety limit. As discussed in the application, 
changing the safety limit MCPR methodology to one based on a 95 
percent probability with 95 percent confidence that no fuel rods 
experience transition boiling during an anticipated transient 
instead of the current limit based on ensuring that 99.9 percent of 
the fuel rods are not susceptible to boiling transition does not 
have a significant effect on plant response to any analyzed 
accident.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

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

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249, 
Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, 
Illinois

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
171, 50-277, and 50-278, Peach Bottom Atomic Power Station, Units 1, 2, 
and 3, York and Lancaster Counties, Pennsylvania

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: March 1, 2019. Publicly-available 
version is in ADAMS under Accession No. ML19063A685.
    Description of amendment request: The amendments would revise the

[[Page 14147]]

emergency action levels (EALs) in the emergency plan for each site. The 
proposed changes are based primarily on the resolution of emergency 
preparedness frequently asked questions (EPFAQs) and industry best-
practices. Editorial changes are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed amendments involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes involving revisions to existing NRC-
approved [Nuclear Energy Institute guidance document] NEI 99-01, 
Revision 6, EALs, as clarified by the NRC through the EPFAQ process, 
for the affected facilities do not reduce the capability to meet the 
emergency planning requirements established in 10 CFR 50.47 and 10 
CFR 50, Appendix E. The proposed changes do not reduce the 
functionality, performance, or capability of Exelon's ERO [emergency 
response organization] to respond in mitigating the consequences of 
any design basis accident.
    The probability of a reactor accident requiring implementation 
of Emergency Plan EALs has no relevance in determining whether the 
proposed changes to the EALs reduce the effectiveness of the 
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of 
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of 
Radiological Emergency Response Plans and Preparedness in Support of 
Nuclear Power Plants;''
    ``. . . The overall objective of emergency response plans is to 
provide dose savings (and in some cases immediate life saving) for a 
spectrum of accidents that could produce offsite doses in excess of 
Protective Action Guides (PAGs). No single specific accident 
sequence should be isolated as the one for which to plan because 
each accident could have different consequences, both in nature and 
degree. Further, the range of possible selection for a planning 
basis is very large, starting with a zero point of requiring no 
planning at all because significant offsite radiological accident 
consequences are unlikely to occur, to planning for the worst 
possible accident, regardless of its extremely low likelihood. . . 
.''

    Therefore, Exelon did not consider the risk insights regarding 
any specific accident initiation or progression in evaluating the 
proposed changes.
    The proposed changes do not involve any physical changes to 
plant equipment or systems, nor do they alter the assumptions of any 
accident analyses. The proposed changes do not adversely affect 
accident initiators or precursors nor do they alter the design 
assumptions, conditions, and configuration or the manner in which 
the plants are operated and maintained. The proposed changes do not 
adversely affect the ability of Structures, Systems, or Components 
(SSCs) to perform their intended safety functions in mitigating the 
consequences of an initiating event within the assumed acceptance 
limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed amendments create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involving revisions to existing NRC-
approved NEI 99-01, Revision 6, EALs, as clarified by the NRC 
through the EPFAQ process, for the affected facilities do not 
involve any physical changes to plant systems or equipment. The 
proposed changes do not involve the addition of any new plant 
equipment. The proposed changes will not alter the design 
configuration, or method of operation of plant equipment beyond its 
normal functional capabilities. Exelon ERO functions will continue 
to be performed as required. The proposed changes do not create any 
new credible failure mechanisms, malfunctions, or accident 
initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from those that have been 
previously evaluated.
    3. Do the proposed amendments involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed changes involving revisions to existing NRC-
approved NEI 99-01, Revision 6, EALs, as clarified by the NRC 
through the EPFAQ process, for the affected facilities do not alter 
or exceed a design basis or safety limit. There is no change being 
made to safety analysis assumptions, safety limits, or limiting 
safety system settings that would adversely affect plant safety as a 
result of the proposed changes. There are no changes to setpoints or 
environmental conditions of any SSC or the manner in which any SSC 
is operated. Margins of safety are unaffected by the proposed 
changes to the EALs based on further NRC clarification through the 
EPFAQ. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, 
Appendix E will continue to be met.
    Therefore, the proposed changes do not involve any reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois, 
and Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station 
(QCNPS), Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: December 5, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18339A009.
    Description of amendment request: The amendments would revise the 
technical specifications for both the single recirculation loop and two 
recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR) 
limits for the DNPS and QCNPS units. The proposed decrease in these 
limits improves operational flexibility through the recapture of 
margins that are available as a result of the transition to Framatome, 
Inc. using NRC-approved SLMCPR calculation methodology.
    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 SLMCPR values have been determined using NRC-
approved methods discussed in AREVA Topical Report ANP-10307PA, 
Revision 0, ``AREVA MCPR Safety Limit Methodology for Boiling Water 
Reactors,'' dated June 2011. The proposed SLMCPRs for two 
recirculation loop and single recirculation loop operation ensure 
that the acceptance criterion continues to be met (i.e., at least 
99.9 percent of all fuel rods in the core do not experience boiling 
transition).
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
proposed license amendments do not involve any plant modifications 
or operational changes that could affect system reliability or 
performance, or that could affect the probability of operator error. 
As such, the proposed changes do not affect any postulated accident 
precursors. Since no individual precursors of an accident are 
affected, the proposed license amendments do not involve a 
significant increase in the probability of a previously analyzed 
event.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The basis for the SLMCPR calculations is to ensure 
that during normal operation and during anticipated operational 
occurrences, at least 99.9 percent of all fuel

[[Page 14148]]

rods in the core do not experience boiling transition if the safety 
limit is not exceeded.
    Based on these considerations, the proposed changes do not 
involve a significant increase in the consequences of a previously 
analyzed accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The SLMCPR is a TS numerical value calculated for two recirculation 
loop operation and single recirculation loop operation to ensure at 
least 99.9 percent of all fuel rods in the core do not experience 
boiling transition if the safety limit is not exceeded. SLMCPR 
values are calculated using NRC-approved methodology identified in 
the TSs. The proposed SLMCPR values do not involve any new modes of 
plant operation or any plant modifications and do not directly or 
indirectly affect the failure modes of any plant systems or 
components. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SLMCPR provides a margin of safety by ensuring that at least 
99.9 percent of the fuel rods do not experience boiling transition 
during normal operation and anticipated operational occurrences if 
the MCPR Safety Limit is not exceeded. Revision of the SLMCPR values 
in TS 2.1.1.2, using an NRC-approved methodology, will ensure that 
the current level of fuel protection is maintained by continuing to 
ensure that the fuel design safety criterion is met (i.e., that no 
more than 0.1 percent of the rods are expected to be in boiling 
transition if the MCPR Safety Limit is not exceeded). The SLMCPRs 
are verified to be bounding by cycle specific analyses prior to 
power operations for each operating cycle. Therefore, the proposed 
amendments do not result in a significant reduction in the margin of 
safety.

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

Exelon Generation Company, LLC, Docket No. 50-244, R. E. Ginna Nuclear 
Power Plant, Wayne County, New York

    Date of amendment request: February 15, 2018. A publicly-available 
version is in ADAMS under Accession No. ML19045A282.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing 
Program,'' to reflect an increase to the existing Type A integrated 
leak rate test program test interval from 10 years to 15 years, in 
accordance with Nuclear Energy Institute (NEI) Report NEI 94-01, 
Revision 2-A, ``Industry Guideline for Implementing Performance-Based 
Option of 10 CFR part 50, Appendix J.'' The proposed change would also 
reflect adoption of both the use of American National Standards 
Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, ``Containment 
System Leakage Testing Requirements,'' and a more conservative 
allowable test interval extension of 9 months for Type A leakage tests 
in accordance with NEI 94-01, Revision 2-A. The amendment would also 
make an administrative change to remove the exception under TS 5.5.15 
for the one-time 15-year Type A test internal being performed after May 
31, 1996, and performed prior to May 31, 2011, as this has already 
occurred.
    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 activity involves the revision of the R. E. Ginna 
Nuclear Power Plant (GNPP) Technical Specification (TS) 5.5.15, 
``Primary Containment Leakage Rate Testing Program,'' to allow the 
extension of the Type A Integrated Leakage Rate Test (ILRT) 
containment test interval to 15 years. Per the guidance provided in 
Nuclear Energy Institute (NEI) 94-01, Industry Guideline for 
Implementing Performance-Based Option of 10 CFR 50, Appendix J, 
Revision 2-A, the current Type A test interval of 10 years would be 
extended on a permanent basis to no longer than 15 years from the 
last Type A test.
    The proposed interval extensions do not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The containment is designed to 
provide an essentially leak tight barrier against the uncontrolled 
release of radioactivity to the environment for postulated 
accidents. As such, the containment and the testing requirements 
invoked to periodically demonstrate the integrity of the containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident.
    The change in Type A test frequency to once-per-fifteen-years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, based on the 
probabilistic risk assessment (PRA) is 0.29 person-Roentgen 
equivalent man (rem)/year. Electric Power Research Institute (EPRI) 
Report No. 1009325, Revision 2A states that a very small population 
dose is defined as an increase of less than 1.0 person-rem per year 
or less than 1 percent of the total population dose, whichever is 
less restrictive for the risk impact assessment of the extended ILRT 
intervals. This is consistent with the Nuclear Regulatory Commission 
(NRC) Final Safety Evaluation which endorsed NEI 94-01 and EPRI 
Report No. 1009325, Revision 2A. Moreover, the risk impact when 
compared to other severe accident risks is negligible. Therefore, 
the proposed extension does not involve a significant increase in 
the probability of an accident previously evaluated.
    In addition, as documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' dated September 1995, Types B and C 
tests have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The GNPP Type A test 
history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel (B&PV) Code, Section XI, ``Rules for 
Inservice Inspection of Nuclear Power Plant Components,'' 
Containment Maintenance Rule Inspections, Containment Coatings 
Program and TS requirements serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
detectable only by a Type A test (ILRT). Based on the above, the 
proposed test interval extensions do not significantly increase the 
consequences of an accident previously evaluated.
    This proposed amendment also deletes the exception previously 
granted to allow one-time extension of the ILRT test frequency for 
GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires 
the first Type A test performed after May 31, 1996, to be performed 
by May 31, 2011. This exception was included in the TS for one-time 
testing activities that would have already taken place by the time 
this amendment is approved; therefore, deletion is solely an 
administrative action that has no effect on any component and no 
impact on how the unit is operated.

[[Page 14149]]

    Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the GNPP TS 5.5.15 involves the 
extension of the GNPP Type A containment test interval from 10 years 
to 15 years. The containment and the testing requirements to 
periodically demonstrate the integrity of the containment exist to 
ensure the plant's ability to mitigate the consequences of an 
accident; thereby, do not involve any accident precursors or 
initiators.
    The proposed change does not involve a physical modification to 
the plant (i.e., no new or different type of equipment will be 
installed) nor does it alter the design, configuration, or change 
the manner in which the plant is operated or controlled beyond the 
standard functional capabilities of the equipment.
    This proposed amendment also deletes the exception previously 
granted to allow one-time extension of the ILRT test frequency for 
GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires 
the first Type A test performed after May 31, 1996, to be performed 
by May 31, 2011. This exception was included in the TS for one-time 
testing activities that would have already taken place by the time 
this amendment is approved; therefore, deletion is solely an 
administrative action that has no effect on any component and no 
impact on how the unit is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.15 involves the extension of 
the GNPP Type A containment test interval to 15 years. This 
amendment does not alter the manner in which safety limits, limiting 
safety system set points, or limiting conditions for operation are 
determined. The specific requirements and conditions of the TS 
Containment Leak Rate Testing Program exist to ensure that the 
degree of containment structural integrity and leak-tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained.
    The proposed change involves the extension of the interval 
between Type A containment leak rate tests for GNPP. The proposed 
surveillance interval extension is bounded by the 15-year ILRT 
interval currently authorized within NEI 94-01, Revision 2-A. 
Industry experience supports the conclusion that Types B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with Option B to 10 CFR 50, Appendix J and 
the overlapping inspection activities performed as part of ASME 
Section Xl, and the TS serve to provide a high degree of assurance 
that the containment would not degrade in a manner that is 
detectable only by Type A testing. The combination of these factors 
ensures that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Types A, B, and C containment leakage tests specified 
in applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
changes to the Type A test intervals.
    In addition, this proposed amendment also deletes the exception 
previously granted to allow one-time extension of the ILRT test 
frequency for GNPP. Specifically, TS 5.5.15, item a. is deleted, as 
it requires the first Type A test performed after May 31, 1996, to 
be performed by May 31, 2011. This exception was included in the TS 
for one-time testing activities that would have already taken place 
by the time this amendment is approved; therefore, deletion is 
solely an administrative action that has no effect on any component 
and no impact on how the unit is operated.
    Therefore, the proposed change does not 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.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
346, Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), Ottawa 
County, Ohio

    Date of amendment request: February 5, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19036A523.
    Description of amendment request: By letter dated April 25, 2018 
(ADAMS Accession No. ML18115A007), FENOC notified the NRC that DBNPS 
will permanently cease power operations by May 31, 2020. The proposed 
amendment would revise the DBNPS renewed facility operating license 
(RFOL) and technical specifications (TSs) following the permanent 
cessation of power operations to reflect the post-shutdown and 
permanently defueled condition. The proposed amendment would eliminate 
TS requirements and license conditions which would not be applicable 
once DBNPS ceases power operations and can no longer place fuel in the 
reactor vessel. The proposed amendment would also eliminate obsolete 
license conditions. In addition, the proposed amendment would revise 
several license conditions and TS requirements, including limiting 
conditions for operation (LCOs), usage rules, definitions, surveillance 
requirements (SRs), and administrative controls. FENOC also proposed to 
revise the licensing bases for DBNPS, including the design bases 
accident (DBA) analysis.
    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 would not take effect until DBNPS has 
certified to the NRC that it has permanently ceased operation and 
entered a permanently defueled condition. Because the 10 CFR part 50 
license for DBNPS will no longer authorize operation of the reactor, 
or emplacement or retention of fuel into the reactor vessel with the 
certifications required by 10 CFR part 50.82(a)(1) submitted, as 
specified in 10 CFR part 50.82(a)(2), the occurrence of postulated 
accidents associated with reactor operation is no longer credible.
    The remaining [Updated Final Safety Analysis Report] UFSAR 
Chapter 15 postulated design basis accident (DBA) events that could 
potentially occur at a permanently defueled facility would be a fuel 
handling accident (FHA) in the spent fuel pool (SFP), the waste gas 
decay tank rupture (WGDTR), and external causes. The FHA analyses 
for DBNPS shows that, following 95 days of decay time after reactor 
shutdown and provided the SFP water level requirements of TS LCO 
3.7.14 are met, the dose consequences are acceptable without relying 
on structures, systems, and components (SSCs) to remain functional 
for accident mitigation during and following the event other than 
the passive SFP structure. The remaining DBAs that support the 
permanently shutdown and defueled condition do not rely on any 
active safety systems for mitigation.
    The probability of occurrence of previously evaluated accidents 
is not increased, since safe storage and handling of fuel will be 
the only operations performed, and therefore, bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation

[[Page 14150]]

will no longer be credible in a permanently defueled reactor. This 
significantly reduces the scope of applicable accidents.
    The deletion of TS definitions and rules of usage and 
application requirements that will not be applicable in a defueled 
condition has no impact on facility SSCs or the methods of operation 
of such SSCs. The deletion of design features and safety limits not 
applicable to the permanently shut down and defueled status of DBNPS 
has no impact on the remaining applicable DBAs.
    The removal of LCOs or SRs that are related only to the 
operation of the nuclear reactor or only to the prevention, 
diagnosis, or mitigation of reactor-related transients or accidents 
do not affect the applicable DBAs previously evaluated since these 
DBAs are no longer applicable in the permanently defueled condition.
    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 delete or modify certain DBNPS RFOL, TS, 
and current licensing bases (CLB) have no impact on facility SSCs 
affecting the safe storage of spent irradiated fuel, or on the 
methods of operation of such SSCs, or on the handling and storage of 
spent irradiated fuel itself. The removal of TS that are related 
only to the operation of the nuclear reactor, or only to the 
prevention, diagnosis, or mitigation of reactor related transients 
or accidents, cannot result in different or more adverse failure 
modes or accidents than previously evaluated because the reactor 
will be permanently shutdown and defueled.
    The proposed modification or deletion of requirements of the 
DBNPS RFOL, TS, and CLB do not affect systems credited in the 
accident analysis for the remaining credible DBAs at DBNPS. The 
proposed RFOL and PDTS [permanently defueled TSs] will continue to 
require proper control and monitoring of safety significant 
parameters and activities. The TS regarding SFP water level and 
spent fuel storage is retained to preserve the current requirements 
for safe storage of irradiated fuel. The proposed amendment does not 
result in any new mechanisms that could initiate damage to the 
remaining relevant safety barriers for defueled plants (fuel 
cladding, spent fuel racks, SFP integrity, and SFP water level). 
Since extended operation in a defueled condition and safe fuel 
handling will be the only operation allowed, and therefore bounded 
by the existing analyses, such a condition does not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes are to delete or modify certain RFOL, TS, 
and CLB once the DBNPS facility has been permanently shutdown and 
defueled. Because the 10 CFR part 50 license for DBNPS will no 
longer authorize operation of the reactor, or emplacement or 
retention of fuel into the reactor vessel, the occurrence of 
postulated accidents associated with reactor operation is no longer 
credible. The remaining postulated DBA events that could potentially 
occur at a permanently defueled facility would be a[n] FHA, WGDTR, 
and external causes. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis 
analyses.
    The proposed changes are limited to those portions of the RFOL, 
TS, and CLB that are not related to the safe storage of irradiated 
fuel. The requirements that are proposed to be revised or deleted 
from the RFOL, TS, and CLB are not credited in the updated 
applicable accident analysis for the remaining applicable postulated 
accidents, and as such, do not contribute to the margin of safety 
associated with the accident analysis. Postulated design basis 
accidents involving the reactor will no longer be possible because 
the reactor will be permanently shutdown and defueled, and DBNPS 
will no longer be authorized to operate the reactor.
    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: Rick Giannantonio, General Counsel, 
FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: David J. Wrona.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: February 26, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19060A060.
    Description of amendment request: The proposed amendment would 
expand the criteria within technical specification (TS) 3.2.1 
surveillance requirements to apply a revised penalty factor to measured 
transient FQ(Z) in response to Westinghouse Nuclear Safety 
Advisory Letter, NSAL-15-1, ``Heat Flux Hot Channel Factor Technical 
Specification Surveillance.''
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed amendment to add an additional surveillance 
requirement, to apply the penalty factor of 1.02 or a factor 
specified in the COLR [core operating limit report], whichever is 
greater, to the transient FQ(Z) calculation, ensures that 
the assumptions and inputs to the safety analyses remain valid and 
does not result in actions that would increase the probability or 
consequences of any accident previously evaluated.
    The design of the protection systems will be unaffected. The 
reactor protection system and engineered safety feature actuation 
system will continue to function in a manner consistent with the 
plant design basis. All design, material and construction standards 
that were applicable prior to the request are maintained.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident-
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Operation in accordance with the revised TS and its limits 
precludes new challenges to systems or structures that might 
introduce a new type of accident. All design and performance 
criteria will continue to be met and no new single failure 
mechanisms will be created. The proposed change for resolution of 
Westinghouse NSAL-15-1 does not involve the alteration of plant 
equipment or introduce unique operational modes or accident 
precursors. Therefore it does not create the potential for a 
different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or, different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation in accordance with the revised TS and its limits 
preserves the margins assumed in the safety analyses. This ensures 
that all design and performance criteria associated with the safety 
analysis will continue to be met and that the margin of safety is 
not affected.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.

[[Page 14151]]

    NRC Branch Chief: David J. Wrona.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: February 26, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19063A498.
    Description of amendment request: The proposed amendment would 
adopt Technical Specification Task Force (TSTF) Traveler TSTF-563, 
``Revise Instrument Testing Definitions to Incorporate the Surveillance 
Frequency Control Program.'' TSTF-563 revises the TS definitions of 
Channel Calibration, Channel Operational Test, and Trip Actuating 
Device Operational Test.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change revises the TS [technical specification] 
definitions of Channel Calibration, COT [channel operational test], 
and TADOT [trip actuating device operational test] to allow the 
frequency for testing the components or devices in each step to be 
determined in accordance with the TS Surveillance Frequency Control 
Program. All components in the channel continue to be tested. The 
frequency at which a channel test is performed is not an initiator 
of any accident previously evaluated, so the probability of an 
accident is not affected by the proposed change. The channels 
surveilled in accordance with the affected definitions continue to 
be required to be operable and the acceptance criteria of the 
surveillances are unchanged. As a result, any mitigating functions 
assumed in the accident analysis will continue to be performed.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident-
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, COT, and TADOT to allow the frequency for testing the 
components or devices in each step to be determined in accordance 
with the TS Surveillance Frequency Control Program. The design 
function or operation of the components involved are not affected 
and there is no physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). No credible new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases are introduced. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis 
assumptions.
    Therefore, the proposed changes do not create the possibility of 
a new or, different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, COT, and TADOT to allow the frequency for testing the 
components or devices in each step to be determined in accordance 
with the TS Surveillance Frequency Control Program. The Surveillance 
Frequency Control Program assures sufficient safety margins are 
maintained, and that design, operation, surveillance methods, and 
acceptance criteria specified in applicable codes and standards (or 
alternatives approved for use by the NRC) will continue to be met as 
described in the plants' licensing basis. The proposed change does 
not adversely affect existing plant safety margins, or the 
reliability of the equipment assumed to operate in the safety 
analysis. As such, there are no changes being made to safety 
analysis assumptions, safety limits, or limiting safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by the method of 
determining surveillance test intervals under an NRC-approved 
licensee-controlled program.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David J. Wrona.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham wCounty, New Hampshire

    Date of amendment request: October 4, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18277A377.
    Description of amendment request: The amendment would revise the 
Seabrook Station, Unit No. 1 (Seabrook), Technical Specifications (TSs) 
and Surveillance Requirements (SRs) associated with the control rods. 
The amendment would adopt changes provided in Technical Specifications 
Task Force (TSTF) Traveler TSTF-234, ``Add Action for More than One 
[D]RPI [Digital Rod Position Indicator] Inoperable,'' and TSTF-547, 
``Clarification of Rod Position Requirements,'' and make various other 
changes to align the Seabrook TSs more closely with NUREG-1431, 
``Standard Technical Specifications--Westinghouse Plants.'' In all, the 
amendment would revise SR 4.1.1.1.1, SR 4.1.1.2, TS 3.1.3.1, SR 
4.1.3.1.1, TS 3.1.3.2, SR 4.1.3.2, TS 3.1.3.3, SR 4.1.3.3, TS 3.1.3.5, 
SR 4.1.3.5, TS 3.1.3.6, SR 4.1.3.6, TS 3.10.5, SR 4.10.5, and TS 
6.8.1.6.b.
    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.
    Control and shutdown rods are assumed to insert into the core to 
shut down the reactor in evaluated accidents. Rod insertion limits 
ensure that adequate negative reactivity is available to provide the 
assumed shutdown margin (SDM). Rod alignment limits maintain an 
appropriate power distribution and reactivity insertion profile.
    Control and shutdown rods are initiators to several accidents 
previously evaluated, such as rod ejection. The proposed change does 
not change the limiting conditions for operation for the rods or 
make any technical changes to the surveillance requirements 
governing the rods. Therefore, the proposed change has no 
significant effect on the probability of any accident previously 
evaluated.
    Adding new TS Actions to provide a limited time to repair rod 
control system failures has no effect on the SDM assumed in the 
accident analysis as the proposed Actions require verification that 
SDM is maintained. The effects on power distribution will not cause 
a significant increase in the consequences of any accident 
previously evaluated as all TS requirements on power distribution 
continue to be applicable.
    The proposed change to resolve the conflicts in the TS ensures 
that the intended Actions are followed when equipment is inoperable. 
Actions taken with inoperable equipment are not assumptions in the 
accidents previously evaluated and have no significant effect on the 
consequences.
    The capability of any operable TS-required equipment to perform 
its specified safety function is not impacted by the proposed 
change. As a result, the outcomes of accidents previously evaluated 
are unaffected. Therefore, the proposed changes do not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?

[[Page 14152]]

    Response: No.
    The proposed change does 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 system 
or component. No physical changes are made to the plant, so no new 
causal mechanisms are introduced. Therefore, the proposed changes to 
the TS 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 ability of the control rods to perform their designated 
safety function is unaffected by the proposed changes. The proposed 
changes do not alter any safety analyses assumptions, safety limits, 
limiting safety system settings, or method of operating the plant. 
The proposed change to provide time to repair rods that are operable 
but immovable does not result in a significant reduction in the 
margin of safety because all rods must be verified to be operable, 
and all other banks must be within the insertion limits. The changes 
do not adversely affect plant operating margins or the reliability 
of equipment credited in the safety analyses. Therefore, the 
proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Debbie Hendell, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: February 27, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19058A221.
    Description of amendment request: The proposed change is consistent 
with Technical Specifications Task Force (TSTF) Traveler TSTF-546, 
Revision 0, ``Revise APRM [Average Power Range Monitor] Channel 
Adjustment Surveillance Requirement'' (ADAMS Accession No. 
ML17205A444). The amendment would alter Surveillance Requirement (SR) 
4.3.1.1 of Technical Specification 3.3.1, ``Reactor Protection System 
Instrumentation.'' The change would revise the SR to verify that 
calculated (i.e., calorimetric heat balance) power is no more than 2 
percent greater than the APRM channel output. The SR requires the APRM 
channel to be adjusted such that calculated power is no more than 2 
percent greater than the APRM indicated power when operating at >=24 
percent of rated thermal power. This change would revise the SR to 
distinguish between APRM indications that are consistent with the 
accident analyses and those that provide additional margin.
    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 APRM system and the RPS are not initiators of any accidents 
previously evaluated. As a result, the proposed change does not 
affect the probability of any accident previously evaluated. The 
APRM system and the Reactor Protection System (RPS) functions act to 
mitigate the consequences of accidents previously evaluated. The 
reliability of APRM system and the RPS is not significantly affected 
by removing the gain adjustment requirement on the APRM channels 
when the APRMs are calibrated conservatively with respect to the 
calculated heat balance. This is because the actual core thermal 
power at which the reactor will automatically trip is lower, thereby 
increasing the margin to the core thermal limits and the limiting 
safety system settings assumed in the safety analyses. The 
consequences of an accident during the adjustment of the APRM 
instrumentation are no different from those during the existing 
surveillance testing period or the existing time allowed to restore 
the instruments to operable status. As a result, the ability of the 
APRM system and the RPS to mitigate any accident previously 
evaluated is not significantly affected.
    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.
    The proposed change does not alter the protection system design, 
create new failure modes, or change any modes of operation. The 
proposed change does not involve a physical alteration of the plant; 
no new or different kind of equipment will be installed. 
Consequently, there are no new initiators that could result in a new 
or different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety provided by the APRM system and the RPS is 
to ensure that the reactor is shut down automatically when plant 
parameters exceed the setpoints for the system. Any reduction in the 
margin of safety resulting from the adjustment of the APRM channels 
while continuing operation is considered to be offset by delaying a 
plant shutdown (i.e., a transient) for a short time with the APRM 
system, the primary indication of core power and an input to the 
RPS, not calibrated. Additionally, the short time period required 
for adjustment is consistent with the time allowed by Technical 
Specifications to restore the core power distribution parameters to 
within limits and is acceptable based on the low probability of a 
transient or design basis accident occurring simultaneously with 
inaccurate APRM channels.
    The proposed change does not alter setpoints or limits 
established or assumed by the accident analyses. The Technical 
Specifications continue to require operability of the RPS functions, 
which provide core protection for postulated reactivity insertion 
events occurring during power operating conditions consistent with 
the plant safety analyses.
    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: Steven Fleischer, PSEG Services Corporation, 
80 Park Plaza, T-5, Newark, NJ 07102.
    NRC Branch Chief: James G. Danna.

Tennessee Valley Authority (TVA), Docket Nos. 50-390 and 50-391, Watts 
Bar Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: October 12, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18288A352.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) by the adoption, with administrative and 
technical variations, of Technical Specification Task Force (TSTF) 
Traveler TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to 
Licensee Control--Risk Informed Technical Specification Task Force 
(RITSTF) Initiative 5b.'' TSTF-425, Revision 3, provides for the 
relocation of specific surveillance frequencies to a licensee-
controlled program. Additionally, the change would add a new program, 
the Surveillance Frequency Control Program (SFCP), to TS Section 5.0, 
``Administrative Controls.''

[[Page 14153]]

    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 relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
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 for which the 
surveillance frequencies are relocated are still required to be 
operable, meet the acceptance criteria for the surveillance 
requirements, 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The change does 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 change does not impose any new or 
different requirements. The change does not alter assumptions made 
in the safety analysis. The proposed change is consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for structures, systems, [and] components, 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), 
because these are not affected by changes to the surveillance 
frequencies. Similarly, there is no effect to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, TVA will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC approved NEI [Nuclear Energy Institute] 04-10, Revision 1, in 
accordance with the TS SFCP. This 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 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 Branch Chief: Undine Shoop.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 1, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19032A632.
    Description of amendment request: The amendments would adopt 
Technical Specification Task Force Traveler TSTF-563, ``Revise 
Instrument Testing Definitions to Incorporate the Surveillance 
Frequency Control Program.''
    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 revises the TS [Technical Specification] 
definitions of Channel Calibration, COT [Channel Operational Test], 
and TADOT [Trip Actuation Device Operational Test] to allow the 
frequency for testing the components or devices in each step to be 
determined in accordance with the TS Surveillance Frequency Control 
Program. All components in the channel continue to be tested. The 
frequency at which a channel test is performed is not an initiator 
of any accident previously evaluated, so the probability of an 
accident is not affected by the proposed change. The channels 
surveilled in accordance with the affected definitions continue to 
be required to be operable and the acceptance criteria of the 
surveillances are unchanged. As a result, any mitigating functions 
assumed in the accident analysis will continue to be performed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, COT, and TADOT to allow the frequency for testing the 
components or devices in each step to be determined in accordance 
with the TS Surveillance Frequency Control Program. The design 
function or operation of the components involved are not affected 
and there is no physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). No credible new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases are introduced. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis 
assumptions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, COT, and TADOT to allow the frequency for testing the 
components or devices in each step to be determined in accordance 
with the TS Surveillance Frequency Control Program. The Surveillance 
Frequency Control Program assures sufficient safety margins are 
maintained, and that design, operation, surveillance methods, and 
acceptance criteria specified in applicable codes and standards (or 
alternatives approved for use by the Nuclear Regulatory Commission 
(NRC)) will continue to be met as described in the plants' licensing 
basis. The proposed change does not adversely affect existing plant 
safety margins, or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes being 
made to safety analysis assumptions, safety limits, or limiting 
safety system settings that would adversely affect plant safety as a 
result of the proposed change. Margins of safety are unaffected by 
method of determining surveillance test intervals under an NRC-
approved licensee-controlled program.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.

[[Page 14154]]

    NRC Branch Chief: Undine Shoop.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Unit 1, Coffey County, Kansas

    Date of amendment request: January 23, 2019, as supplemented by 
letter dated March 11, 2019. Publicly-available versions are in ADAMS 
under Accession Nos. ML19036A772 and ML19078A131, respectively.)
    Description of amendment request: The amendment would revise 
technical specification (TS) requirements in Section 1.3, ``Completion 
Times,'' and Section 3.0, ``Limiting Condition for Operation (LCO) 
Applicability,'' regarding LCO and surveillance requirement (SR) usage. 
The proposed changes are consistent with the NRC-approved Technical 
Specifications Task Force (TSTF) Traveler TSTF-529, Revision 4, 
``Clarify Use and Application Rules,'' using the consolidated line item 
improvement process (ADAMS Accession No. ML16062A271). The model safety 
evaluation was approved by the NRC in a letter dated April 21, 2016 
(ADAMS Package Accession No. ML16060A441).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Section 1.3 and LCO 3.0.4 have no effect 
on the requirement for systems to be Operable and have no effect on 
the application of TS actions. The proposed change to SR 3.0.3 
states that the allowance may only be used when there is a 
reasonable expectation the surveillance will be met when performed. 
Since the proposed change does not significantly affect system 
Operability, the proposed change will have no significant effect on 
the initiating events for accidents previously evaluated and will 
have no significant effect on the ability of the systems to mitigate 
accidents previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the TS usage rules does not affect the 
design or function of any plant systems. The proposed change does 
not change the Operability requirements for plant systems or the 
actions taken when plant systems are not operable.
    Therefore, it is concluded that this 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 clarifies the application of Section 1.3 and 
LCO 3.0.4 and does not result in changes in plant operation. SR 
3.0.3 is revised to allow application of SR 3.0.3 when an SR has not 
been previously performed if there is reasonable expectation that 
the SR will be met when performed. This expands the use of SR 3.0.3 
while ensuring the affected system is capable of performing its 
safety function. As a result, plant safety is either improved or 
unaffected.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 1200 17th Street NW, Washington, DC 20036.
    NRC Branch Chief: Robert J. Pascarelli.

IV. 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.

Dominion Energy Nuclear Connecticut, Inc., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: April 4, 2018, as supplemented by letter 
dated October 22, 2018.
    Brief description of amendment: The amendment revised ACTION 18 in 
Technical Specifications Table 3.3-3, Functional Unit 7.e, ``Control 
Building Inlet Ventilation Radiation,'' for Millstone Power Station, 
Unit No. 3, to allow continued fuel handling and reactor operation with 
inoperable inlet radiation monitoring instrumentation provided that one 
train of the control room emergency ventilation system is operating in 
the emergency mode. The technical specification change specifies that 
one train of the control room emergency ventilation system be placed in 
the emergency mode of operation within 7 days if one radiation monitor 
channel is inoperable, or immediately, if both radiation monitor 
channels are inoperable.
    Date of issuance: March 21, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 272. A publicly-available version is in ADAMS under 
Accession No. ML19042A277; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-49: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: July 17, 2018 (83 FR 
33266). The supplemental letter dated October 22, 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

[[Page 14155]]

consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2019.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003 and 50-247, Indian 
Point Nuclear Generating Unit Nos. 1 and 2 (Indian Point 1 and Indian 
Point 2), Westchester County, New York

    Date of amendment request: June 20, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18179A173.
    Brief description of amendments: The amendments deleted certain 
license conditions from the Indian Point 1 and Indian Point 2 Operating 
Licenses that impose specific requirements on the decommissioning trust 
agreement. With approval of these amendments, the provisions of 10 CFR 
50.75(h), which specify the regulatory requirements for decommissioning 
trust funds, apply to the licensee, Entergy Nuclear Operations, Inc., 
for Indian Point 1 and Indian Point 2.
    Date of issuance: March 21, 2019.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment Nos.: 61 (Unit No. 1) and 289 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML19065A101; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Provisional Operating License No. DPR-5 and Renewed Facility 
Operating License No. DPR-26: The amendments revised the Operating 
Licenses.
    Date of initial notice in Federal Register: September 11, 2018 (83 
FR 45984).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 26, 2018.
    Brief description of amendment: The amendment revised the Updated 
Final Safety Analysis Report descriptions for the replacement of the 
Turbine First Stage Pressure output signals with Power Range Neutron 
Monitoring System output signals.
    Date of issuance: March 12, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No: 217. A publicly-available version is in ADAMS under 
Accession No. ML18215A196; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-29: The amendment 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 5, 2018 (83 FR 
26115).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 27, 2018, as supplemented by 
letter dated October 10, 2018.
    Brief description of amendment: The amendment revised the GGNS 
Emergency Plan to adopt an Emergency Action Level scheme based on 
Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, 
``Development of Emergency Action Levels for Non-Passive Reactors,'' 
dated November 2012, which was endorsed by the NRC by letter dated 
March 28, 2013.
    Date of issuance: March 12, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 365 days of issuance.
    Amendment No: 216. A publicly-available version is in ADAMS under 
Accession No. ML19025A023; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-29: The amendment 
revised the GGNS Emergency Plan.
    Date of initial notice in Federal Register: June 5, 2018 (83 FR 
26104). The supplemental letter dated October 10, 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 March 12, 2019.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

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

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249, 
Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, 
Illinois

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: January 31, 2018, as supplemented by 
letters dated July 27 and November 29, 2018.
    Brief description of amendments: The amendments revise the 
emergency response organization positions identified in the emergency 
plan for each site.
    Date of issuance: March 21, 2019.
    Effective date: As of the date of issuance and shall be implemented 
on or before December 31, 2019.
    Amendment Nos.: Braidwood 201/201, Byron 206/206, Clinton 223, 
Dresden 46/261/254, LaSalle 236/222, and Quad Cities 274/269. A 
publicly-available version is in ADAMS under Accession No. ML19036A586. 
Documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, 
NPF-62, DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, and DPR-30: 
Amendments revised the emergency plans.
    Date of initial notice in Federal Register: April 10, 2018 (83 FR 
15417).
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated March 21, 2019.
    No significant hazards consideration comments received: No.

[[Page 14156]]

Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

    Date of amendment request: May 14, 2018, as supplemented by letter 
dated November 20, 2018.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to increase the minimum load required 
for the Emergency Diesel Generator (EDG) partial-load rejection 
Surveillance Requirement (SR). Additionally, the amendments modified 
the EDG voltage and frequency limits for the SR and established a 
recovery period for the EDG(s) to return to steady-state conditions.
    Date of issuance: March 18, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1, 285 and Unit 2, 279. A publicly-available 
version is in ADAMS under Accession No. ML18354A673; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: July 3, 2018 (83 FR 
31185). The supplemental letter dated November 20, 2018, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 18, 2019.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket No. 52-025, Vogtle Electric 
Generating Plant (VEGP), Unit 3, Burke County, Georgia

    Date of amendment request: October 19, 2018.
    Description of amendment: The amendment authorizes the Southern 
Nuclear Operating Company to depart from certified AP1000 Design 
Control Document (DCD) Tier 2* material that has been incorporated into 
the Updated Final Safety Analysis Report (UFSAR). Specifically, the 
proposed departure consists of changes to Tier 2* information in the 
UFSAR (which includes the plant-specific DCD information) to change the 
vertical reinforcement information provided in the VEGP Unit 3 column 
line 1 wall from elevation 135'-3'' to 137'-0'' .
    Date of issuance: March 13, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 156 for Unit 3. Publicly-available versions are in 
an ADAMS package under Accession No. ML19044A500 which includes the 
Safety Evaluation that references documents, located in that ADAMS 
package, related to this amendment.
    Facility Combined Licenses No. NPF-91: Amendment revised the 
Facility Combined License.
    Date of initial notice in Federal Register: November 20, 2018 (83 
FR 58607).
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated March 13, 2019.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, 
Unit 2, Rhea County, Tennessee

    Date of amendment request: March 5, 2018, as supplemented by 
letters dated April 27 and October 11, 2018.
    Brief description of amendment: The amendment revised License 
Condition 2.C.(4), concerning the use of the PAD4TCD computer program. 
While the current License Condition permits the use of PAD4TCD for Unit 
2, Cycles 1 and 2 only, the revision allows the use of PAD4TCD until 
the Unit 2 steam generators (SGs) are replaced with SGs equivalent to 
the existing SGs at Unit 1.
    Date of issuance: March 20, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 26. A publicly-available version is in ADAMS under 
Accession No. ML19046A286; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-96: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: December 4, 2018 (83 FR 
62623). The supplemental letters dated April 27 and October 11, 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 March 20, 2019.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant (Watts Bar), Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: August 1, 2018, as supplemented by 
letter dated March 4, 2019.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to adopt, with minor variation, Technical 
Specification Task Force (TSTF) Traveler TSTF-266-A, Revision 3, 
``Eliminate the Remote Shutdown System Table of Instrumentation and 
Controls.'' Specifically, the comparable TS Table 3.3.4-1, ``Remote 
Shutdown System Instrumentation and Controls,'' was deleted from Watts 
Bar, Units 1 and 2, TS 3.3.4, ``Remote Shutdown System.''
    Date of issuance: March 18, 2019.
    Effective date: As of the date of issuance and shall be implemented 
by March 24, 2019.
    Amendment Nos.: 124 and 25. A publicly-available version is in 
ADAMS under Accession No. ML19066A009; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. NPF-90 and NPF-96: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 12, 2019 (84 
FR 3510). The supplemental letter dated March 4, 2019, requested 
expedited completion of the NRC review of 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 (NSHC) determination as published in the Federal 
Register.
    The Commission's related evaluation of the amendments and final 
NSHC determination are contained in a Safety Evaluation dated March 18, 
2019.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 29th day of March 2019.
    For the Nuclear Regulatory Commission.
Craig G. Erlanger,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2019-06449 Filed 4-8-19; 8:45 am]
BILLING CODE 7590-01-P