82 FR 27882 - Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 82, Issue 116 (June 19, 2017)
Federal Register Volume 82, Issue 116 (June 19, 2017)
| Page Range | 27882-27895 | |
| FR Document | 2017-12732 |
[Federal Register Volume 82, Number 116 (Monday, June 19, 2017)] [Notices] [Pages 27882-27895] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-12732] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0140] Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission. ACTION: Biweekly notice. ----------------------------------------------------------------------- SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued, from May 23, 2017, to June 2, 2017. The last biweekly notice was published on June 6, 2017. DATES: Comments must be filed by July 19, 2017. A request for a hearing must be filed by August 18, 2017. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0140. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1927, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2017-0140 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0140. 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/ [[Page 27883]] adams.html. To begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [email protected]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document. NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2017-0140 facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment 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 Web site 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. [[Page 27884]] 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 by August 18, 2017. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally- recognized Indian Tribe, or agency thereof may participate as a non- party under 10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC Web site 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 Web site 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 Web site 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 Web site 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 [[Page 27885]] by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester County, New York Date of amendment request: December 14, 2016, as supplemented by letter dated April 19, 2017. Publicly available versions are in ADAMS under Package Accession No. ML16355A066 and Accession No. ML17114A467, respectively. Description of amendment request: The amendments would revise the Appendix C Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.1.2 for IP2 and IP3 and Appendix A TS LCO 3.7.13 for IP2. These LCOs ensure that the fuel to be loaded into the Shielded Transfer Canister (STC) meets the design basis for the STC and has an acceptable rack location in the IP2 spent fuel pit before the STC is loaded with fuel. The proposed changes to these LCOs would increase the population of IP3 fuel eligible for transfer to the IP2 spent fuel pit and maintain full core offload capability for IP3. 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, with NRC staff's edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment would modify the IP2 and IP3 Technical Specifications (TS) to incorporate the results of revised criticality, thermal, and shielding and dose analyses and evaluations. [For IP2,] the proposed amendment was evaluated for impact on the following previously evaluated events and accidents: STC Criticality Accidents, SFP Criticality Accidents, Boron Dilution Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool [SFP] Cooling, and Natural Events. [IP2] STC Criticality Accidents The STC criticality accident considered were: Abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, and misalignment between the active fuel region and the neutron absorber. The probability of an STC criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the STC in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of an STC criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for these accidents. [IP2] SFP Criticality Accidents The SFP criticality accident of record considered the following accidents (1) a dropped fuel assembly or an assembly placed alongside a rack, (2) a misloaded fuel assembly, and (3) abnormal heat loads. Because the IP2 and IP3 fuel assemblies are identical [with] regards [to] those parameters that are utilized in the design basis criticality analysis (DBA) to qualify fresh fuel these accidents are bounding for IP3 fuel. The probability of an SFP criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of an SFP criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for this accident. [IP2] STC Thermal Accidents The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI-TRAC annulus, fuel misload, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria. The probability of an STC thermal accident will not increase significantly because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of an STC thermal accident will not increase significantly because the thermal analysis demonstrates that the same thermal acceptance criteria and requirements continue to be met for this accident. [IP2] Boron Dilution Accident The probability of a boron dilution event remains the same because the proposed change does not alter the manner in which the IP2 spent fuel cooling system or any other plant system is operated, or otherwise increase the likelihood of adding significant quantities of unborated water into the spent fuel pit. The consequences of the boron dilution event remains the same. The reactivity of the STC filled with the most reactive combination of approved fuel assemblies in unborated water results in a k eff less than 0.95. Thus, even in the unlikely event of a complete dilution of the spent fuel pit water, the STC will remain safely subcritical. [IP2] Fuel Handling Accident The probability of an FHA will not increase significantly due to the proposed changes because the individual fuel assemblies will be moved between the STC and the spent fuel pit racks and the STC and HI-TRAC will be moved in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of the existing fuel handling accident remain bounding because the IP3 fuel assembly design is essentially the same as the IP2 design and the IP3 fuel assemblies to be transferred to IP2 will be cooled a minimum of 6 years. This compares with a cooling time of 84 hours used in the existing FHA radiological analysis. The 6-year cooling time results in a significant reduction in the radioactive source term available for release from a damaged fuel assembly compared to the source term considered in the design basis FHA radiological analysis. The consequences of the previously analyzed fuel assembly drop accident, therefore, continue to provide a [[Page 27886]] bounding estimate of offsite dose for this accident. [IP2] Loss of Spent Fuel Pool Cooling The probability of a loss of spent fuel pit cooling remains the same because the proposed change does not alter the manner in which the IP2 spent fuel cooling loop is operated, designed or maintained. The consequences of a loss of spent fuel pit cooling remains the same because the thermal design basis for the spent fuel pit cooling loop provides for all fuel pit rack locations to be filled at the end of a full core discharge and therefore the design basis heat load effectively includes any heat load associated with the assemblies within the STC. [IP2] Natural Events The natural events considered include the following accidents (1) a seismic event, (2) high winds, tornado and tornado missiles, (3) flooding and (4) a lightning strike. The probability of natural event will not increase due to the proposed changes because there are no elements of the proposed changes that influence the occurrence of any natural event. The consequences of a natural event will not increase due to the proposed changes because the structural analyses design limits continue to be met. A lightning strike may cause ignition of the VCT fuel but this event is addressed under STC thermal accidents. [For IP3,] the proposed amendment was evaluated for impact on the following previously evaluated events and accidents: STC Criticality Accidents, SFP Criticality Accidents, Boron Dilution Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool Cooling, and Natural Events. [IP3] STC Criticality Accidents The STC criticality accident considered were: Abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, and misalignment between the active fuel region and the neutron absorber. The probability of an STC criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the STC in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of an STC criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for these accidents. [IP3] STC Thermal Accidents The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria. The probability of an STC thermal accident will not increase significantly because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of an STC thermal accident will not increase significantly because the thermal analysis demonstrates that the same thermal acceptance criteria and requirements continue to be met for this accident. [IP3] Boron Dilution Accident The probability of a boron dilution event remains the same because the proposed change does not alter the manner in which the IP3 spent fuel cooling system or any other plant system is operated, or otherwise increase the likelihood of adding significant quantities of unborated water into the spent fuel pit. The consequences of the boron dilution event remains the same. The reactivity of the STC filled with the most reactive combination of approved fuel assemblies in unborated water results in a keff less than 0.95. Thus, even in the unlikely event of a complete dilution of the spent fuel pit water, the STC will remain safely subcritical. [IP3] Fuel Handling Accident The probability of an FHA will not increase significantly due to the proposed changes because the individual fuel assemblies will be moved between the STC and the spent fuel pit racks and the STC and HI-TRAC will be moved in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used. The consequences of the existing fuel handling accident remain bounding because only IP3 fuel is moved in the IP3 spent fuel pit. The IP3 fuel assemblies to be transferred to IP2 will be cooled a minimum of 6 years. This compares with a cooling time of 84 hours used in the existing FHA radiological analysis. The 6-year cooling time results in a significant reduction in the radioactive source term available for release from a damaged fuel assembly compared to the source term considered in the design basis FHA radiological analysis. The consequences of the previously analyzed fuel assembly drop accident, therefore, continue to provide a bounding estimate of offsite dose for this accident. [IP3] Loss of Spent Fuel Pool Cooling The probability of a loss of spent fuel pit cooling remains the same because the proposed change does not alter the manner in which the IP3 spent fuel cooling loop is operated, designed or maintained. The consequences of a loss of spent fuel pit cooling remains the same because the thermal design basis for the spent fuel pit cooling loop provides for all fuel pit rack locations to be filled at the end of a full core discharge and therefore the design basis heat load effectively includes any heat load associated with the assemblies within the STC. [IP3] Natural Events The natural events considered include the following accidents (1) a seismic event, (2) high winds, tornado and tornado missiles, (3) flooding and (4) a lightning strike. The probability of natural event will not increase due to the proposed changes because there are no elements of the proposed changes that influence the occurrence of any natural event. The consequences of a natural event will not increase due to the proposed changes because the structural analyses design limits continue to be met. A lightning strike may cause ignition of the VCT fuel but this event is addressed under STC thermal accidents. 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 TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. 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 amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment would modify the TS to incorporate the results of revised criticality, thermal and shield and dose analyses. The margin of safety required by 10 CFR 50.58(b)(4) remains unchanged. New criticality evaluations for both the STC [and the IP2 SFP] confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins. The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI- TRAC annulus, fuel misload, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria without a significant loss of safety margin. The shielding and dose analyses demonstrate that the shielding and radiation protection requirements continue to be met without a significant loss of safety margin. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: James G. Danna. [[Page 27887]] Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland Date of amendment request: March 28, 2017. A publicly available version is in ADAMS under Accession No. ML17087A374. Description of amendment request: The amendments would revise the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications (TSs) to change the low level of the refueling water tank (RWT) to reflect a needed increase in the required borated water volume and change the allowable value of the RWT level-low function. 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 any accident previously evaluated? Response: No. The proposed amendment increases the required volume of water in the RWT to maintain the existing design requirements. The increase is necessary due to an increase in the RWT Level--Low RAS [recirculation actuation signal] setpoint, which allows more water to stay in the tank following a LOCA [loss-of-coolant accident]. The modification to the allowable value of the RWT level-low (function 5a) resolves a non-conservative TS per the guidance of Administrative Letter 98-10 ``Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety.'' The RWT is not an accident initiator. The RWT is required to supply adequate borated water to perform its mitigation function as assumed in the accident analyses. With the proposed increase in the minimum required water volume, the RWT maintains its design margin for supplying the required amount of borated water to the reactor core and the containment sump. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed amendment increases the required volume of water in the RWT to maintain the existing design requirements. The increase is necessary due to an increase in the RWT Level--Low RAS setpoint, which allows more water to stay in the tank following a LOCA. The modification to the allowable value of the RWT level-low (function 5a) resolves a non-conservative TS per the guidance of Administrative Letter 98-10 ``Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety.'' The proposed amendment does not impose any new or different requirements. The change does not alter assumptions made in the safety analyses. The proposed change is consistent with the safety analyses 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 previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment increases the required volume of water in the RWT to maintain the existing design requirements. The increase is necessary due to an increase in the RWT Level--Low RAS setpoint, which allows more water to stay in the tank following a loss-of- coolant accident. The modification to the allowable value of the RWT level-low (function 5a) resolves a non-conservative TS per the guidance of Administrative Letter 98-10 ``Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety.'' The proposed amendment does not affect the design, operation, and testing methods for systems, structures and components specified in applicable codes and standards (or alternatives approved for use by the NRC). With the proposed increase in the minimum required water volume, the RWT maintains its design margin for supplying the required amount of borated water to the reactor core and the containment sump. The RWT will continue to meet all of its requirements as described in the plant licensing basis (including the Updated Final Safety Analysis Report and the TS Bases). Similarly, there is no impact to Safety Analysis acceptance criteria as described in the plant licensing basis. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: James G. Danna. Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of amendment request: April 5, 2017. A publicly available version is in ADAMS under Accession No. ML17095A081. Description of amendment request: The amendment would revise the Nine Mile Point Nuclear Station, Unit 2, Technical Specifications to allow greater flexibility in performing surveillance testing in Modes 1, 2, or 3 of emergency diesel generators and Class 1E batteries. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction Notes.'' 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 modify Mode restriction Notes to allow performance of the Surveillance in whole or in part to reestablish Emergency Diesel Generator (EDG) Operability, and to allow the crediting of unplanned events that satisfy the Surveillances. The EDGs and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. To manage any increase in risk, the proposed changes require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the current prohibited Modes. The radiological consequences of an accident previously evaluated during the period that the EDG is being tested to reestablish operability are no different from the radiological consequences of an accident previously evaluated while the EDG is inoperable. As a result, the consequences of any accident previously evaluated are not increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. 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 amendment involve a significant reduction in a margin of safety? Response: No. The purpose of Surveillances is to verify that equipment is capable of performing its assumed safety function. The proposed changes will only allow the performance of the Surveillances to reestablish Operability, [[Page 27888]] and the proposed changes may not be used to remove an EDG from service. In addition, the proposed changes will potentially shorten the time that an EDG is unavailable because testing to reestablish Operability can be performed without a plant shutdown. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the normally prohibited Modes. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: James G. Danna. 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: April 27, 2017. A publicly available version is in ADAMS under Accession No. ML17121A449. Description of amendment request: The proposed amendments would revise Technical Specification 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to allow for the permanent extension of the Type A integrated leak rate testing and Type C leak rate testing frequencies, and would also delete a one-time exception. 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 activity involves revision of the Quad Cities Nuclear Power Station (QCNPS) Technical Specification (TS) 5.5.12, Primary Containment Leakage Rate Testing Program, to allow the extension of the QCNPS, Units 1 and 2, Type A containment integrated leakage rate test interval to 15 years, and the extension of the Type C local leakage rate test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The existing Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The 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 dose risk for changing the Type A Integrated Leak Rate Test (ILRT) interval from three-per-ten years to once-per- fifteen-years, measured as an increase to the total integrated dose risk for all internal events accident sequences for QCNPS, is 1.0E- 02 person-rem/yr (0.31%) using the Electric Power Research Institute (EPRI) guidance with the base case corrosion included. The change in dose risk drops to 2.7E-03 person-rem/yr (0.08%) when using the EPRI Expert Elicitation methodology. The values calculated per the EPRI guidance are all lower than the acceptance criteria of less than or equal to 1.0 person-rem/yr or less than 1.0% person-rem/yr defined in Section 1.3 of Attachment 3 to this LAR. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' dated January 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 QCNPS, Units 1 and 2 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) Section XI, 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. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes an exception previously granted in amendments 220 and 214 to allow one-time extensions of the ILRT test frequency for QCNPS, Units 1 and 2, respectively. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how QCNPS, Units 1 and 2 are operated. 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 amendment to TS 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' involves the extension of the QCNPS, Units 1 and 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. 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. 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. The proposed amendment also deletes an exception previously granted under TS Amendments 220 and 214 to allow the one-time extension of the ILRT test frequency for QCNPS, Units 1 and 2, respectively. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how the QCNPS units are operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.12 involves the extension of the QCNPS, Units 1 and 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. 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 and Type C tests for QCNPS, Units 1 and 2. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized [[Page 27889]] within NEI 94-01, Revision 3-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 ASME Section Xl and 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 and Type C test intervals. The proposed amendment also deletes exceptions previously granted to allow one-time extensions of the ILRT test frequency for QCNPS, Units 1 and 2. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how QCNPS is operated and maintained. Thus, there is no reduction in any 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: Tamra Domeyer, Associate General Counsel, Exelon Nuclear Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Florida Power & Light Company, Docket Nos. 50-250 and 251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida Date of amendment request: April 9, 2017. A publicly available version is in ADAMS under Accession No. ML17101A637. Description of amendment request: The amendments would modify the Technical Specifications (TSs) to remove various reporting requirements. Specifically, the amendments would remove the requirements to prepare various special reports, the Startup Report, and the Annual Report. In addition, the amendments would revise the TSs to remove the completion time for restoring spent fuel pool water level to address inoperability of one of the two parallel flow paths in the residual heat removal or safety injection headers for the Emergency Core Cooling Systems and to make other administrative changes, including updating plant staff and responsibilities and correcting a misspelling. 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 actions, surveillance requirements, and administrative controls associated with the proposed changes to the technical specifications (TS) are not initiators of any accidents previously evaluated, so the probability of accidents previously evaluated is unaffected by the proposed changes. The proposed changes do not alter the design, function, operation, or configuration of any plant structure, system, or component (SSC). The capability of any operable TS-required SSC to perform its specified safety function is not impacted by the proposed changes. 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. 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 amendment involve a significant reduction in a margin of safety? Response: No. The ability of any operable SSC to perform its 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 changes do not adversely impact 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: William S. Blair, Managing Attorney-- Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, FL 33408-0420. NRC Branch Chief: Undine S. Shoop. NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: April 20, 2017. A publicly available version is in ADAMS under Accession No. ML17111A631. Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) Section 3.1.2, ``Reactivity Anomalies,'' with a change to the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance. The proposed change would allow performance of the reactivity anomaly surveillance on a comparison of monitored to predicted core reactivity. The reactivity anomaly verification is currently determined by a comparison of monitored versus predicted control rod density. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The proposed change would only modify how the reactivity anomaly surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the TS requirements such that, rather than performing the surveillance by comparing monitored to predicted control rod density, the surveillance is performed by a direct comparison of core keff . Present day on-line core monitoring systems, such as 3D MONICORE and ACUMEN, are capable of performing the direct measurement of reactivity. Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed change does not involve a significant increase in the probability or consequence of a previously evaluated accident. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. [[Page 27890]] The proposed change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the Reactivity Anomalies TS will only provide a new, more efficient method of detecting an unexpected change in core reactivity. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change is to modify the method for performing the reactivity anomaly surveillance from a comparison of monitored to predicted control rod density to a comparison of monitored to predicted core keff . The direct comparison of keff provides a technically superior method of calculating any differences in the expected core reactivity. The reactivity anomaly surveillance will continue to be performed at the same frequency as is currently required by the TS, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified. The proposed change has no impact to 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: William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant (Point Beach), Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: March 31, 2017. A publicly available version is in ADAMS under Accession No. ML17090A511. Description of amendment request: The amendments would document a risk-informed resolution strategy to resolve low risk, legacy design code non-conformances associated with construction trusses in the containment buildings of Point Beach, Units 1 and 2. The proposed license amendment request (LAR) is a risk-informed licensing basis change. The proposed change is acceptance of the final configuration of the construction trusses, including the attached containment spray piping and ventilation ductwork, and the containment liners/walls adjacent to the trusses, using a risk-informed resolution. Accordingly, the proposed change meets the criteria set forth in Regulatory Guide (RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment [PRA] in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,'' and the generic guidance in RG 1.200, ``An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities.'' 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 probability of an accident previously evaluated is not changed. The containment structures and the containment spray piping and ventilation ducts attached to the construction trusses are accident mitigation equipment. They are not accident initiators. The acceptance of the final configuration of Point Beach Units 1 and 2 results in a change in core damage frequency and large early release frequency that is within acceptance guidelines and does not involve a significant reduction in the margin of safety. Although failures are postulated in the PRA analysis, the engineering calculations in support of the LAR conclude that the construction trusses and the associated structures/components remain structurally sound in the event of a design basis seismic or thermal event and there is no adverse impact or change to any station SSC's [structure, system, and components] design function and there is no change to accident mitigation response. This change has no impact on station fire risk caused by a seismic event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not install any new or different type of equipment in the plant. The proposed change does not create any new failure modes for existing equipment or any new limiting single failures. Engineering calculations conclude the construction trusses, equipment supported by the trusses, and containment liners remain capable of withstanding design basis seismic and thermal events and remain capable of performing their designated design functions. Additionally, the proposed change does not involve a change in the methods governing normal plant operation, and all safety functions will continue to perform as previously assumed in the accident analyses. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems and components important to safety. There are no new accidents identified associated with acceptance of the final modified configuration of Unit 1 and the current configuration of Unit 2. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The effects of the change, [Delta]CDF [core damage frequency] and [Delta]LERF, [large early release frequency] are within the acceptance guidelines shown in Figures 4 and 5 of Regulatory Guide 1.174. Consequently, the change does not result in a significant reduction in the margin of safety. The containment structures and liners, construction trusses, and equipment supported by the trusses remain fully capable of performing their specified design functions as concluded by supporting engineering calculations. Modifications associated with implementation of NFPA [National Fire Protection Association] 805 are planned that will provide protection of the reactor coolant system feed and bleed capability and result in additional safety margin. The proposed change does not affect the margin of safety associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change does not alter any safety analyses assumptions, safety limits, limiting safety system settings, or methods of operating the plant. The changes do not adversely impact the reliability of equipment credited in the safety analyses. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The station will implement new seismic and thermal event limits to ensure the construction trusses and associated equipment are inspected and/or analyzed for any event exceeding elastic stress limits to determine their capability to withstand a subsequent design basis event prior to Unit restart. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Blair, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420. NRC Branch Chief: David J. Wrona. [[Page 27891]] Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: April 27, 2017. A publicly available version is in ADAMS under Accession No. ML17118A049. Description of amendment request: The requested amendments propose changes to combined license (COL) Appendix C (and plant-specific Tier 1) Table 2.7.2-2 to revise the minimum chilled water flow rates to the supply air handling units serving the Main Control Room and the Class 1E electrical rooms, and the unit coolers serving the normal residual heat removal system and chemical and volume control system pump rooms. The proposed COL Appendix C (and plant-specific Design Control Document (Tier 1) changes require additional changes to corresponding Tier 2 component data information in Updated Final Safety Analysis Report (UFSAR) Chapter 9. Because this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document, the licensee also requested an exemption from the requirements of the Generic Design Control Document Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, Updated Final Safety Analysis Report (UFSAR) Table 9.2.7-1, and associated UFSAR design information to identify the revised equipment parameters for the nuclear island nonradioactive ventilation system (VBS) air (VAS) unit coolers and reduced chilled water system (VWS) cooling coil flow rates do not adversely impact the plant response to any accidents which are previously evaluated. The function of the cooling coils to provide chilled water to the VBS AHUs and VAS unit coolers is not credited in the safety analysis. No safety-related structure, system, component (SSC) or function is adversely affected by this change. The VWS safety-related function of containment isolation is not affected by this change. The change does not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the plant-specific UFSAR are not affected. The proposed changes do not involve a change to the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected. The proposed changes do not increase the probability or consequences of an accident previously evaluated as the VWS, VBS and VAS do not provide safety-related functions and the functions of each system to support required room environments are not changed. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design information to identify the revised equipment parameters for VBS AHUs and VAS unit coolers and reduced VWS cooling coil flow rates do not affect any safety-related equipment, and do not add any new interfaces to safety-related SSCs. The VWS function to provide chilled water is not adversely impacted. The function of the VAS to provide ventilation and cooling to maintain the environment of the serviced areas within the design temperature range is not adversely impacted by this change. No system or design function or equipment qualification is affected by these changes as the change does not modify the operation of any SSCs. The changes do not introduce a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment. Revised equipment parameters, including the reduced cooling coil flow rates, do not adversely impact the function of associated components. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The changes to COL Appendix C (and plant-specific Tier 1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design information do not affect any other safety-related equipment or fission product barriers. The requested changes will not adversely affect compliance with any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes as previously evaluated accidents are not impacted. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, and 72- 052, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, and Independent Spent Fuel Storage Installation (ISFSI), Limestone County, Alabama Tennessee Valley Authority, Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant (SQN), Units 1 and 2, and ISFSI, Hamilton County, Tennessee Tennessee Valley Authority (TVA), Docket Nos. 50-390, 50-391, and 72- 1048, Watts Bar Nuclear Plant (WBN), Units 1 and 2, and ISFSI, Rhea County, Tennessee Date of amendment request: January 4, 2017. A publicly available version is in ADAMS under Accession No. ML17004A340. Description of amendment request: The amendments would modify the Emergency Plans for BFN, Units 1, 2, and 3, and its ISFSI; SQN, Units 1 and 2, and its ISFSI; and WBN, Units 1 and 2, and its ISFSI, to adopt the Emergency Action Level (EAL) schemes based on Nuclear Energy Institute (NEI) 99-01, Revision 6, which has been endorsed by the NRC as documented in a letter dated March 28, 2013 (ADAMS Accession No. ML12346A463). The proposed changes to TVA's EAL schemes to adopt the guidance in NEI 99-01, Revision 6, do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR part 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of TVA's Emergency Response Organization (ERO) to respond in mitigating the consequences of accidents. The TVA ERO functions will continue to be performed as required. 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 changes to TVA's EAL schemes to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, ``Development of Emergency Action Levels [[Page 27892]] for Non-Passive Reactors,'' do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR [Part] 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of TVA's ERO to respond in mitigating the consequences of any design basis accident. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facilities or the manner in which the plants are operated and maintained. The proposed change does not adversely affect the ability of structures, systems, and components (SSC) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptable limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to TVA's EAL schemes to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, 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. All TVA 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. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes to TVA's EAL schemes to adopt the NRC- endorsed guidance in NEI 99-01, Revision 6, 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 adopt the NEI 99-01, Revision 6, EAL scheme guidance. The applicable requirements of 10 CFR 50.47 and 10 CFR [Part] 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 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: Benjamin G. Beasley. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson), Darlington County, South Carolina Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1 (Harris), Wake and Chatham Counties, North Carolina Date of amendment request: August 19, 2015, as supplemented by letters dated May 4, October 3, and November 17, 2016. Brief description of amendments: The amendments revised the Robinson Technical Specification (TS) 5.6.5.b and the Harris TS 6.9.1.6.2 to adopt the methodology reports DPC-NE-1008-P, Revision 0, ``Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors''; DPC-NF-2010, Revision 3, ``Nuclear Physics Methodology for Reload Design''; and DPC-NE-2011-P, Revision 2, ``Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors,'' for application specific to Robinson and Harris. Date of issuance: May 18, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 253 (Robinson) and 157 (Harris). A publicly available version is in ADAMS under Accession No. ML17102A923; documents related to these amendments are listed in the Safety Evaluations enclosed with the amendments. Renewed Facility Operating License Nos. DPR-23 and NPF-63: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: February 2, 2016 (81 FR 5492). The supplemental letter dated May 4, 2016, provided additional information that expanded the scope of the application as originally noticed, and changed the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. Accordingly, the NRC published a second proposed no significant hazards consideration determination in the Federal Register on August 2, 2016 (81 FR 50746). This notice superseded the original notice in its entirety. The supplemental letters dated October 3 and November 17, 2016, provided additional information that clarified the application, did not expand the scope beyond the second notice, and did not change the NRC staff's proposed no significant hazards consideration determination as published in the Federal Register. [[Page 27893]] The Commission's related evaluations of the amendments are contained in the Safety Evaluations dated May 18, 2017. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendment request: December 21, 2015, as supplemented by letters dated June 29, July 13, August 15, November 1, November 17, 2016, and February 27, 2017. Brief description of amendments: The amendments adopted the approved changes to Standard Technical Specifications for General Electric (BWR/4) [Boiling Water Reactor] Plants, NUREG-1433, Revision 4, to allow relocation of specific technical specification surveillance frequencies to a licensee-controlled program. The changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee Control-- RITSTF Initiative 5b'' (ADAMS Package Accession No. ML090850642), and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996). Date of issuance: May 24, 2017. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment Nos.: 276 (Unit 1) and 304 (Unit 2). A publicly available version is in ADAMS under Accession No. ML17096A129; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-71 and DPR-62: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: March 29, 2016 (81 FR 17504). The supplemental letters dated June 29, July 13, August 15, November 1, November 17, 2016, and February 27, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 24, 2017. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: May 26, 2016, as supplemented by letter dated December 19, 2016. Brief description of amendment: The amendment revised the Technical Specifications (TSs) by adding a new Administrative Controls section to establish, implement, and maintain a Diesel Fuel Oil Testing Program. It also relocated to this program the current TS surveillance requirements (SRs) for evaluating diesel fuel oil, along with the SRs for draining, sediment removal, and cleaning of each main fuel oil storage tank at least once every 10 years. In addition, the licensee took an exception to NRC Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems for Standby Diesel Generators,'' to allow for the ability to perform sampling of new fuel oil offsite prior to its addition to the fuel oil storage tanks. Date of issuance: May 24, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment No.: 158. A publicly available version is in ADAMS under Accession No. ML17048A184; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-63: Amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: October 11, 2016 (81 FR 70178). The supplemental letter dated December 19, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2017. No significant hazards consideration comments received: No. Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: June 29, 2016, as supplemented by letter dated November 4, 2016. Brief description of amendment: The amendment revised the Shearon Harris Nuclear Power Plant, Unit 1, Technical Specification (TS) 3/ 4.11.1.4, ``Liquid Holdup Tanks''; TS 3/4.11.2.5, ``Explosive Gas Mixture''; and TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring Program.'' The amendment deleted TS Definition 1.16, ``GASEOUS RADWASTE TREATMENT SYSTEM''; TS 3/4.11.1.4, ``Liquid Holdup Tanks''; and TS 3/ 4.11.2.5, ``Explosive Gas Mixture.'' The amendment relocated the deleted requirements for these TSs to licensee control under TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring Program.'' The description for TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring Program,'' was modified to include the controls for potentially explosive gas mixtures contained in the Gaseous Waste Processing System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The amendment relocated requirements associated with TS 3/4.11.1.4 and TS 3/4.11.2.5 to the licensee-controlled Plant Programs Procedure PLP-114, ``Relocated Technical Specifications and Design Basis Requirements.'' Date of issuance: May 25, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No.: 159. A publicly available version is in ADAMS under Accession No. ML17074A672; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-63: The amendment revised the Facility Operating License and TSs. Date of initial notice in Federal Register: October 25, 2016 (81 FR 73433). The supplemental letter dated November 4, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 25, 2017. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant, Van Buren County, Michigan Date of amendment request: July 11, 2016. Brief description of amendment: The amendment approved adoption of NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-545, Revision 3, ``TS [Technical Specification] Inservice Testing Program [[Page 27894]] Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,'' dated October 21, 2015. Specifically, the amendment deleted Palisades Nuclear Plant TS 5.5.7, ``Inservice Testing Program,'' and added a new defined term, ``INSERVICE TESTING PROGRAM,'' to the TSs. All existing references to the ``Inservice Testing Program,'' in the Palisades Nuclear Plant TS SRs are replaced with ``INSERVICE TESTING PROGRAM'' so that the SRs refer to the new definition in lieu of the deleted program. Date of issuance: May 30, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 262. A publicly available version is in ADAMS under Accession No. ML17082A465; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-20: Amendment revised the Renewed Facility Operating License and TSs. Date of initial notice in Federal Register: August 30, 2016 (81 FR 59663). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 30, 2017. 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 Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 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. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 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 Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Exelon Generation Company, LLC, Docket No. 50-244, R. E. Ginna Nuclear Power Plant, Wayne County, New York Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania Date of amendment request: July 26, 2016, as supplemented by letter dated October 6, 2016. Brief description of amendments: The amendments revised the Inservice Testing Program requirements in each plant's technical specifications (TSs). The changes included deleting the current TS requirements for the Inservice Testing Program, adding a new defined term, ``INSERVICE TESTING PROGRAM,'' to the TSs, and revising other TSs to reference this new defined term instead of the deleted program. Date of issuance: May 26, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 191, 192, 197, 197, 320, 298, 212, 254, 247, 223, 209, 227, 161, 313, 317, 266, 261, 124, and 290. A publicly available version is in ADAMS under Accession No. ML17073A067. Documents related to these amendments are listed in the Safety Evaluations enclosed with the amendments. Facility Operating License Nos.: NPF-72, NPF-77, NPF-37, NPF-66, DPR-53, DPR-69, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-63, NPF-69, DPR-44, DPR-56, DPR-29, DPR-30, DPR-18, and DPR-50. Amendments revised the Facility Operating Licenses and TSs. Date of initial notice in Federal Register: November 8, 2016 (81 FR 78648). The Commission's related evaluations of the amendments are contained in Safety Evaluations dated May 26, 2017. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania Date of amendment request: June 24, 2016, as supplemented by letters dated September 13, 2016; December 15, 2016; and March 16, 2017. Brief description of amendment: The amendment modified the Renewed Facility Operating License to reflect the direct transfer of Toledo Edison Company's 18.26 percent leased interest in Beaver Valley Power Station, Unit 2, and Ohio Edison Company's 21.66 percent leased interest in Beaver Valley Power Station, Unit 2, from FirstEnergy Nuclear Operating Company to FirstEnergy Nuclear Generation, LLC. Date of issuance: May 30, 2017. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 187. A publicly available version is in ADAMS under Accession No. ML17115A123. Renewed Facility Operating License No. NPF-73: Amendment revised the Renewed Facility Operating License. Date of initial notice in Federal Register: January 23, 2017 (82 FR 7880). The supplemental letter dated March 16, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 14, 2017. 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: July 21, 2016, as supplemented by letter dated September 26, 2016. Brief description of amendments: The amendments revised the Donald C. Cook Nuclear Plant, Units 1 and 2, Technical Specification (TS) Surveillance Requirements (SRs), consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.'' Specifically, the change revised the TSs to eliminate Section 5.5.6, ``Inservice Testing Program.'' A new defined term, ``INSERVICE TESTING PROGRAM,'' was added to the TS Definitions section. TS SRs that previously referred to the Inservice Testing Program from Section 5.5.6 were revised to refer to the new defined term, ``INSERVICE TESTING PROGRAM.'' Date of issuance: May 24, 2017. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 335 (Unit 1) and 317 (Unit 2). A publicly available version is in ADAMS under Accession No. ML17103A106; documents related [[Page 27895]] to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: September 27, 2016 (81 FR 66307). The supplemental letter dated September 26, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 24, 2017. No significant hazards consideration comments received: No. Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia Date of amendment request: May 18, 2016, as supplemented by letters dated February 10, 2017; March 1, 2017; and March 10, 2017. Brief description of amendments: The amendments revised Technical Specification 3.14 ``Circulating and Service Water Systems,'' to extend the Allowed Outage Time for only one operable Service Water flow path to the Changing Pump Service Water subsystem and to the Main Control Room/Emergency Switchgear Room air conditioning subsystem. Date of issuance: May 31, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 289 (Unit 1) and 289 (Unit 2). A publicly available version is in ADAMS under Accession No. ML17100A253; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: October 25, 2016 (81 FR 73443). The supplemental letters dated February 10, 2017; March 1, 2017; and March 10, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 31, 2017. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 6th day of June 2017. For the Nuclear Regulatory Commission. Eric J. Benner, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2017-12732 Filed 6-16-17; 8:45 am] BILLING CODE 7590-01-P
![27882 Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices
confirming receipt of the document. The electronic hearing docket which is Commission that such amendment
E-Filing system also distributes an email available to the public at https:// involves no significant hazards
notice that provides access to the adams.nrc.gov/ehd, unless excluded consideration, notwithstanding the
document to the NRC’s Office of the pursuant to an order of the Commission pendency before the Commission of a
General Counsel and any others who or the presiding officer. If you do not request for a hearing from any person.
have advised the Office of the Secretary have an NRC-issued digital ID certificate This biweekly notice includes all
that they wish to participate in the as described above, click cancel when notices of amendments issued, or
proceeding, so that the filer need not the link requests certificates and you proposed to be issued, from May 23,
serve the document on those will be automatically directed to the 2017, to June 2, 2017. The last biweekly
participants separately. Therefore, NRC’s electronic hearing dockets where notice was published on June 6, 2017.
applicants and other participants (or you will be able to access any publicly DATES: Comments must be filed by July
their counsel or representative) must available documents in a particular 19, 2017. A request for a hearing must
apply for and receive a digital ID hearing docket. Participants are be filed by August 18, 2017.
certificate before adjudicatory requested not to include personal
ADDRESSES: You may submit comments
documents are filed so that they can privacy information, such as social
security numbers, home addresses, or by any of the following methods (unless
obtain access to the documents via the
E-Filing system. personal phone numbers in their filings, this document describes a different
A person filing electronically using unless an NRC regulation or other law method for submitting comments on a
the NRC’s adjudicatory E-Filing system requires submission of such specific subject):
may seek assistance by contacting the information. For example, in some • Federal Rulemaking Web site: Go to
NRC’s Electronic Filing Help Desk instances, individuals provide home http://www.regulations.gov and search
through the ‘‘Contact Us’’ link located addresses in order to demonstrate for Docket ID NRC–2017–0140. Address
on the NRC’s public Web site at http:// proximity to a facility or site. With questions about NRC dockets to Carol
www.nrc.gov/site-help/e- respect to copyrighted works, except for Gallagher; telephone: 301–415–3463;
submittals.html, by email to limited excerpts that serve the purpose email: Carol.Gallagher@nrc.gov. For
MSHD.Resource@nrc.gov, or by a toll- of the adjudicatory filings and would technical questions, contact the
free call at 1–866–672–7640. The NRC constitute a Fair Use application, individual listed in the FOR FURTHER
Electronic Filing Help Desk is available participants are requested not to include INFORMATION CONTACT section of this
between 9 a.m. and 6 p.m., Eastern copyrighted materials in their document.
Time, Monday through Friday, submission. • Mail comments to: Cindy Bladey,
excluding government holidays. Office of Administration, Mail Stop:
Dated at Rockville, Maryland, this 12th day
Participants who believe that they TWFN–8–D36M, U.S. Nuclear
of June 2017.
have a good cause for not submitting Regulatory Commission, Washington,
For the Nuclear Regulatory Commission.
documents electronically must file an DC 20555–0001.
Jacob Zimmerman, For additional direction on obtaining
exemption request, in accordance with
10 CFR 2.302(g), with their initial paper Chief, Enrichment and Conversion Branch, information and submitting comments,
Division of Fuel Cycle Safety, Safeguards, see ‘‘Obtaining Information and
filing stating why there is good cause for and Environmental Review, Office of Nuclear
not filing electronically and requesting Material Safety and Safeguards. Submitting Comments’’ in the
authorization to continue to submit SUPPLEMENTARY INFORMATION section of
[FR Doc. 2017–12696 Filed 6–16–17; 8:45 am]
documents in paper format. Such filings this document.
BILLING CODE 7590–01–P
must be submitted by: (1) First class FOR FURTHER INFORMATION CONTACT:
mail addressed to the Office of the Lynn Ronewicz, Office of Nuclear
Secretary of the Commission, U.S. NUCLEAR REGULATORY Reactor Regulation, U.S. Nuclear
Nuclear Regulatory Commission, COMMISSION Regulatory Commission, Washington,
Washington, DC 20555–0001, Attention: DC 20555–0001; telephone: 301–415–
Rulemaking and Adjudications Staff; or [NRC–2017–0140] 1927, email: Lynn.Ronewicz@nrc.gov.
(2) courier, express mail, or expedited SUPPLEMENTARY INFORMATION:
delivery service to the Office of the Biweekly Notice: Applications and
Secretary, 11555 Rockville Pike, Amendments to Facility Operating I. Obtaining Information and
Rockville, Maryland, 20852, Attention: Licenses and Combined Licenses Submitting Comments
Rulemaking and Adjudications Staff. Involving No Significant Hazards
Considerations A. Obtaining Information
Participants filing adjudicatory
documents in this manner are AGENCY: Nuclear Regulatory Please refer to Docket ID NRC–2017–
responsible for serving the document on Commission. 0140 facility name, unit number(s),
all other participants. Filing is ACTION: Biweekly notice. plant docket number, application date,
considered complete by first-class mail and subject when contacting the NRC
as of the time of deposit in the mail, or SUMMARY: Pursuant to Section 189a.(2) about the availability of information for
by courier, express mail, or expedited of the Atomic Energy Act of 1954, as this action. You may obtain publicly
delivery service upon depositing the amended (the Act), the U.S. Nuclear available information related to this
document with the provider of the Regulatory Commission (NRC) is action by any of the following methods:
• Federal Rulemaking Web site: Go to
asabaliauskas on DSKBBXCHB2PROD with NOTICES
service. A presiding officer, having publishing this regular biweekly notice.
granted an exemption request from The Act requires the Commission to http://www.regulations.gov and search
using E-Filing, may require a participant publish notice of any amendments for Docket ID NRC–2017–0140.
or party to use E-Filing if the presiding issued, or proposed to be issued, and • NRC’s Agencywide Documents
officer subsequently determines that the grants the Commission the authority to Access and Management System
reason for granting the exemption from issue and make immediately effective (ADAMS): You may obtain publicly
use of E-Filing no longer exists. any amendment to an operating license available documents online in the
Documents submitted in adjudicatory or combined license, as applicable, ADAMS Public Documents collection at
proceedings will appear in the NRC’s upon a determination by the http://www.nrc.gov/reading-rm/
VerDate Sep<11>2014 17:09 Jun 16, 2017 Jkt 241001 PO 00000 Frm 00101 Fmt 4703 Sfmt 4703 E:\FR\FM\19JNN1.SGM 19JNN1](https://thefederalregister.org/doc/82_FR_27998/page/1.svg)
![Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices 27885
by courier, express mail, or expedited (LCO) 3.1.2 for IP2 and IP3 and in the same manner, using the same
delivery service upon depositing the Appendix A TS LCO 3.7.13 for IP2. equipment, procedures, and other
document with the provider of the These LCOs ensure that the fuel to be administrative controls (i.e. fuel move sheets)
that are currently used.
service. A presiding officer, having loaded into the Shielded Transfer
The consequences of an SFP criticality
granted an exemption request from Canister (STC) meets the design basis accident are not changed because the
using E-Filing, may require a participant for the STC and has an acceptable rack reactivity analysis demonstrates that the
or party to use E-Filing if the presiding location in the IP2 spent fuel pit before same subcriticality criteria and requirements
officer subsequently determines that the the STC is loaded with fuel. The continue to be met for this accident.
reason for granting the exemption from proposed changes to these LCOs would [IP2] STC Thermal Accidents
use of E-Filing no longer exists. increase the population of IP3 fuel The thermal analyses demonstrate that the
Documents submitted in adjudicatory eligible for transfer to the IP2 spent fuel postulated accidents (rupture of the HI–
proceedings will appear in the NRC’s pit and maintain full core offload TRAC water jacket, 50-gallon transported fuel
electronic hearing docket which is capability for IP3. tank rupture and fire, simultaneous loss of
available to the public at https:// Basis for proposed no significant water from the water jacket and HI–TRAC
adams.nrc.gov/ehd, unless excluded hazards consideration determination: annulus, fuel misload, hypothetical tipover,
pursuant to an order of the Commission As required by 10 CFR 50.91(a), the and crane malfunction) continue to meet
or the presiding officer. If you do not licensee has provided its analysis of the their acceptance criteria.
have an NRC-issued digital ID certificate The probability of an STC thermal accident
issue of no significant hazards
as described above, click cancel when will not increase significantly because the
consideration, which is presented individual fuel assemblies will be loaded
the link requests certificates and you below, with NRC staff’s edits in square into the SFP in the same manner, using the
will be automatically directed to the brackets: same equipment, procedures, and other
NRC’s electronic hearing dockets where administrative controls (i.e. fuel move sheets)
1. Does the proposed amendment involve
you will be able to access any publicly a significant increase in the probability or that are currently used.
available documents in a particular consequences of an accident previously The consequences of an STC thermal
hearing docket. Participants are evaluated? accident will not increase significantly
requested not to include personal Response: No. because the thermal analysis demonstrates
privacy information, such as social The proposed amendment would modify that the same thermal acceptance criteria and
security numbers, home addresses, or the IP2 and IP3 Technical Specifications (TS) requirements continue to be met for this
personal phone numbers in their filings, to incorporate the results of revised accident.
unless an NRC regulation or other law criticality, thermal, and shielding and dose [IP2] Boron Dilution Accident
analyses and evaluations.
requires submission of such The probability of a boron dilution event
[For IP2,] the proposed amendment was
information. For example, in some evaluated for impact on the following
remains the same because the proposed
instances, individuals provide home change does not alter the manner in which
previously evaluated events and accidents:
addresses in order to demonstrate STC Criticality Accidents, SFP Criticality the IP2 spent fuel cooling system or any other
proximity to a facility or site. With Accidents, Boron Dilution Accidents, Fuel plant system is operated, or otherwise
respect to copyrighted works, except for Handling Accidents, Loss of Spent Fuel Pool increase the likelihood of adding significant
[SFP] Cooling, and Natural Events. quantities of unborated water into the spent
limited excerpts that serve the purpose fuel pit.
of the adjudicatory filings and would [IP2] STC Criticality Accidents The consequences of the boron dilution
constitute a Fair Use application, The STC criticality accident considered event remains the same. The reactivity of the
participants are requested not to include were: Abnormal temperature, dropped, STC filled with the most reactive
copyrighted materials in their mislocated, and misloaded fuel assemblies, combination of approved fuel assemblies in
submission. and misalignment between the active fuel unborated water results in a keff less than
For further details with respect to region and the neutron absorber. 0.95. Thus, even in the unlikely event of a
these license amendment applications, The probability of an STC criticality complete dilution of the spent fuel pit water,
see the application for amendment accident will not increase significantly due to the STC will remain safely subcritical.
which is available for public inspection the proposed changes because the individual [IP2] Fuel Handling Accident
fuel assemblies will be loaded into the STC
in ADAMS and at the NRC’s PDR. For in the same manner, using the same The probability of an FHA will not
additional direction on accessing equipment, procedures, and other increase significantly due to the proposed
information related to this document, administrative controls (i.e. fuel move sheets) changes because the individual fuel
see the ‘‘Obtaining Information and that are currently used. assemblies will be moved between the STC
Submitting Comments’’ section of this The consequences of an STC criticality and the spent fuel pit racks and the STC and
document. accident are not changed because the HI–TRAC will be moved in the same manner,
reactivity analysis demonstrates that the using the same equipment, procedures, and
Entergy Nuclear Operations, Inc., same subcriticality criteria and requirements other administrative controls (i.e. fuel move
Docket Nos. 50–247 and 50–286, Indian continue to be met for these accidents. sheets) that are currently used.
Point Nuclear Generating Unit Nos. 2 The consequences of the existing fuel
[IP2] SFP Criticality Accidents handling accident remain bounding because
and 3 (IP2 and IP3), Westchester
The SFP criticality accident of record the IP3 fuel assembly design is essentially the
County, New York considered the following accidents (1) a same as the IP2 design and the IP3 fuel
Date of amendment request: dropped fuel assembly or an assembly placed assemblies to be transferred to IP2 will be
December 14, 2016, as supplemented by alongside a rack, (2) a misloaded fuel cooled a minimum of 6 years. This compares
assembly, and (3) abnormal heat loads.
asabaliauskas on DSKBBXCHB2PROD with NOTICES
letter dated April 19, 2017. Publicly with a cooling time of 84 hours used in the
available versions are in ADAMS under Because the IP2 and IP3 fuel assemblies are existing FHA radiological analysis. The 6-
Package Accession No. ML16355A066 identical [with] regards [to] those parameters year cooling time results in a significant
that are utilized in the design basis criticality reduction in the radioactive source term
and Accession No. ML17114A467, analysis (DBA) to qualify fresh fuel these available for release from a damaged fuel
respectively. accidents are bounding for IP3 fuel. assembly compared to the source term
Description of amendment request: The probability of an SFP criticality considered in the design basis FHA
The amendments would revise the accident will not increase significantly due to radiological analysis. The consequences of
Appendix C Technical Specifications the proposed changes because the individual the previously analyzed fuel assembly drop
(TS) Limiting Condition for Operation fuel assemblies will be loaded into the SFP accident, therefore, continue to provide a
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![27886 Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices
bounding estimate of offsite dose for this The consequences of an STC thermal proposed changes that influence the
accident. accident will not increase significantly occurrence of any natural event.
[IP2] Loss of Spent Fuel Pool Cooling because the thermal analysis demonstrates The consequences of a natural event will
that the same thermal acceptance criteria and not increase due to the proposed changes
The probability of a loss of spent fuel pit requirements continue to be met for this
cooling remains the same because the because the structural analyses design limits
accident. continue to be met. A lightning strike may
proposed change does not alter the manner
in which the IP2 spent fuel cooling loop is [IP3] Boron Dilution Accident cause ignition of the VCT fuel but this event
operated, designed or maintained. The probability of a boron dilution event is addressed under STC thermal accidents.
The consequences of a loss of spent fuel pit remains the same because the proposed Therefore, the proposed change does not
cooling remains the same because the change does not alter the manner in which involve a significant increase in the
thermal design basis for the spent fuel pit the IP3 spent fuel cooling system or any other probability or consequences of an accident
cooling loop provides for all fuel pit rack plant system is operated, or otherwise previously evaluated.
locations to be filled at the end of a full core increase the likelihood of adding significant 2. Does the proposed amendment create
discharge and therefore the design basis heat quantities of unborated water into the spent the possibility of a new or different kind of
load effectively includes any heat load fuel pit. accident from any accident previously
associated with the assemblies within the The consequences of the boron dilution evaluated?
STC. event remains the same. The reactivity of the Response: No.
[IP2] Natural Events STC filled with the most reactive The proposed TS changes do not create the
The natural events considered include the combination of approved fuel assemblies in possibility of a new or different kind of
following accidents (1) a seismic event, (2) unborated water results in a keff less than accident from any accident previously
high winds, tornado and tornado missiles, (3) 0.95. Thus, even in the unlikely event of a evaluated. No new modes of operation are
flooding and (4) a lightning strike. complete dilution of the spent fuel pit water, introduced by the proposed changes. The
The probability of natural event will not the STC will remain safely subcritical. proposed changes will not create any failure
increase due to the proposed changes [IP3] Fuel Handling Accident mode not bounded by previously evaluated
because there are no elements of the accidents.
The probability of an FHA will not
proposed changes that influence the Therefore, the proposed changes do not
increase significantly due to the proposed
occurrence of any natural event.
changes because the individual fuel create the possibility of a new or different
The consequences of a natural event will
assemblies will be moved between the STC kind of accident, from any accident
not increase due to the proposed changes
and the spent fuel pit racks and the STC and previously evaluated.
because the structural analyses design limits
HI–TRAC will be moved in the same manner, 3. Does the proposed amendment involve
continue to be met. A lightning strike may
using the same equipment, procedures, and a significant reduction in a margin of safety?
cause ignition of the VCT fuel but this event
other administrative controls (i.e. fuel move Response: No.
is addressed under STC thermal accidents.
[For IP3,] the proposed amendment was sheets) that are currently used. The proposed amendment would modify
evaluated for impact on the following The consequences of the existing fuel the TS to incorporate the results of revised
previously evaluated events and accidents: handling accident remain bounding because criticality, thermal and shield and dose
STC Criticality Accidents, SFP Criticality only IP3 fuel is moved in the IP3 spent fuel analyses. The margin of safety required by 10
Accidents, Boron Dilution Accidents, Fuel pit. The IP3 fuel assemblies to be transferred
CFR 50.58(b)(4) remains unchanged. New
Handling Accidents, Loss of Spent Fuel Pool to IP2 will be cooled a minimum of 6 years.
criticality evaluations for both the STC [and
Cooling, and Natural Events. This compares with a cooling time of 84
hours used in the existing FHA radiological the IP2 SFP] confirm that operation in
[IP3] STC Criticality Accidents analysis. The 6-year cooling time results in accordance with the proposed amendment
The STC criticality accident considered a significant reduction in the radioactive continues to meet the required subcriticality
were: Abnormal temperature, dropped, source term available for release from a margins. The thermal analyses demonstrate
mislocated, and misloaded fuel assemblies, damaged fuel assembly compared to the that the postulated accidents (rupture of the
and misalignment between the active fuel source term considered in the design basis HI–TRAC water jacket, 50-gallon transported
region and the neutron absorber. FHA radiological analysis. The consequences fuel tank rupture and fire, simultaneous loss
The probability of an STC criticality of the previously analyzed fuel assembly of water from the water jacket and HI–TRAC
accident will not increase significantly due to drop accident, therefore, continue to provide annulus, fuel misload, hypothetical tipover,
the proposed changes because the individual a bounding estimate of offsite dose for this and crane malfunction) continue to meet
fuel assemblies will be loaded into the STC accident. their acceptance criteria without a significant
in the same manner, using the same [IP3] Loss of Spent Fuel Pool Cooling loss of safety margin. The shielding and dose
equipment, procedures, and other analyses demonstrate that the shielding and
administrative controls (i.e. fuel move sheets) The probability of a loss of spent fuel pit
radiation protection requirements continue to
that are currently used. cooling remains the same because the
be met without a significant loss of safety
The consequences of an STC criticality proposed change does not alter the manner
in which the IP3 spent fuel cooling loop is margin.
accident are not changed because the Therefore, the proposed change does not
reactivity analysis demonstrates that the operated, designed or maintained.
The consequences of a loss of spent fuel pit involve a significant reduction in a margin of
same subcriticality criteria and requirements
cooling remains the same because the safety.
continue to be met for these accidents.
thermal design basis for the spent fuel pit
[IP3] STC Thermal Accidents The NRC staff has reviewed the
cooling loop provides for all fuel pit rack
The thermal analyses demonstrate that the locations to be filled at the end of a full core licensee’s analysis and, based on this
postulated accidents (rupture of the HI– discharge and therefore the design basis heat review, it appears that the three
TRAC water jacket, 50-gallon transported fuel load effectively includes any heat load standards of 10 CFR 50.92(c) are
tank rupture and fire, simultaneous loss of associated with the assemblies within the satisfied. Therefore, the NRC staff
asabaliauskas on DSKBBXCHB2PROD with NOTICES
water from the water jacket and HI–TRAC STC. proposes to determine that the
annulus, fuel mislead, hypothetical tipover,
and crane malfunction) continue to meet [IP3] Natural Events amendment request involves no
their acceptance criteria. The probability of The natural events considered include the significant hazards consideration.
an STC thermal accident will not increase following accidents (1) a seismic event, (2) Attorney for licensee: Jeanne Cho,
significantly because the individual fuel high winds, tornado and tornado missiles, (3) Assistant General Counsel, Entergy
assemblies will be loaded into the SFP in the flooding and (4) a lightning strike. Nuclear Operations, Inc., 440 Hamilton
same manner, using the same equipment, The probability of natural event will not
procedures, and other administrative controls increase due to the proposed changes
Avenue, White Plains, NY 10601.
(i.e. fuel move sheets) that are currently used. because there are no elements of the NRC Branch Chief: James G. Danna.
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![Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices 27887
Exelon Generation Company, LLC, The proposed amendment does not impose greater flexibility in performing
Docket Nos. 50–317 and 50–318, Calvert any new or different requirements. The surveillance testing in Modes 1, 2, or 3
Cliffs Nuclear Power Plant, Units 1 and change does not alter assumptions made in of emergency diesel generators and
the safety analyses. The proposed change is
2, Calvert County, Maryland Class 1E batteries. The proposed
consistent with the safety analyses
Date of amendment request: March assumptions and current plant operating changes are based on Technical
28, 2017. A publicly available version is practice. Specifications Task Force (TSTF)
in ADAMS under Accession No. Therefore, the proposed change does not Traveler TSTF–283–A, Revision 3,
ML17087A374. create the possibility of a new or different ‘‘Modify Section 3.8 Mode Restriction
Description of amendment request: kind of accident from any previously Notes.’’
The amendments would revise the evaluated. Basis for proposed no significant
3. Does the proposed amendment involve hazards consideration determination:
Calvert Cliffs Nuclear Power Plant, a significant reduction in a margin of safety?
Units 1 and 2, Technical Specifications As required by 10 CFR 50.91(a), the
Response: No. licensee has provided its analysis of the
(TSs) to change the low level of the The proposed amendment increases the
refueling water tank (RWT) to reflect a issue of no significant hazards
required volume of water in the RWT to
needed increase in the required borated maintain the existing design requirements. consideration, which is presented
water volume and change the allowable The increase is necessary due to an increase below:
value of the RWT level-low function. in the RWT Level—Low RAS setpoint, which 1. Does the proposed amendment involve
Basis for proposed no significant allows more water to stay in the tank a significant increase in the probability or
hazards consideration determination: following a loss-of-coolant accident. The consequences of an accident previously
As required by 10 CFR 50.91(a), the modification to the allowable value of the evaluated?
RWT level-low (function 5a) resolves a non- Response: No.
licensee has provided its analysis of the conservative TS per the guidance of The proposed changes modify Mode
issue of no significant hazards Administrative Letter 98–10 ‘‘Dispositioning restriction Notes to allow performance of the
consideration, which is presented of Technical Specifications That Are Surveillance in whole or in part to
below: Insufficient to Assure Plant Safety.’’ reestablish Emergency Diesel Generator
1. Does the proposed amendment involve The proposed amendment does not affect (EDG) Operability, and to allow the crediting
a significant increase in the probability or the design, operation, and testing methods of unplanned events that satisfy the
consequences of any accident previously for systems, structures and components Surveillances. The EDGs and their associated
evaluated? specified in applicable codes and standards emergency loads are accident mitigating
Response: No. (or alternatives approved for use by the NRC). features, and are not an initiator of any
The proposed amendment increases the With the proposed increase in the minimum accident previously evaluated. As a result,
required volume of water in the RWT to required water volume, the RWT maintains the probability of any accident previously
maintain the existing design requirements. its design margin for supplying the required evaluated is not significantly increased. To
The increase is necessary due to an increase amount of borated water to the reactor core manage any increase in risk, the proposed
in the RWT Level—Low RAS [recirculation and the containment sump. The RWT will changes require an assessment to verify that
actuation signal] setpoint, which allows more continue to meet all of its requirements as plant safety will be maintained or enhanced
water to stay in the tank following a LOCA described in the plant licensing basis by performance of the Surveillance in the
[loss-of-coolant accident]. The modification (including the Updated Final Safety Analysis current prohibited Modes. The radiological
to the allowable value of the RWT level-low Report and the TS Bases). Similarly, there is consequences of an accident previously
(function 5a) resolves a non-conservative TS no impact to Safety Analysis acceptance evaluated during the period that the EDG is
per the guidance of Administrative Letter 98– criteria as described in the plant licensing being tested to reestablish operability are no
10 ‘‘Dispositioning of Technical basis. different from the radiological consequences
Specifications That Are Insufficient to Assure Therefore, the proposed amendment does of an accident previously evaluated while the
Plant Safety.’’ not involve a significant reduction in a EDG is inoperable. As a result, the
The RWT is not an accident initiator. The margin of safety. consequences of any accident previously
RWT is required to supply adequate borated evaluated are not increased.
The NRC staff has reviewed the Therefore, the proposed changes do not
water to perform its mitigation function as
assumed in the accident analyses. With the
licensee’s analysis and, based on this involve a significant increase in the
proposed increase in the minimum required review, it appears that the three probability or consequences of any accident
water volume, the RWT maintains its design standards of 10 CFR 50.92(c) are previously evaluated.
margin for supplying the required amount of satisfied. Therefore, the NRC staff 2. Does the proposed amendment create
borated water to the reactor core and the proposes to determine that the the possibility of a new or different kind of
containment sump. amendment request involves no accident from any accident previously
Therefore, the proposed amendment does evaluated?
significant hazards consideration.
not involve a significant increase in the Response: No.
Attorney for licensee: Tamra Domeyer, The proposed changes do not involve a
probability or consequences of an accident
Associate General Counsel, Exelon physical alteration to the plant (i.e., no new
previously evaluated.
2. Does the proposed amendment create Generation Company, LLC, 4300 or different type of equipment will be
the possibility of a new or different kind of Winfield Road, Warrenville, IL 60555. installed) or a change to the methods
accident from any previously evaluated? NRC Branch Chief: James G. Danna. governing normal plant operation. The
Response: No. changes do not alter the assumptions made
Exelon Generation Company, LLC, in the safety analysis.
The proposed amendment increases the
required volume of water in the RWT to
Docket No. 50–410, Nine Mile Point Therefore, the proposed changes do not
maintain the existing design requirements. Nuclear Station, Unit 2, Oswego County, create the possibility of a new or different
New York kind of accident from any accident
asabaliauskas on DSKBBXCHB2PROD with NOTICES
The increase is necessary due to an increase
in the RWT Level—Low RAS setpoint, which Date of amendment request: April 5, previously evaluated.
allows more water to stay in the tank 2017. A publicly available version is in 3. Does the proposed amendment involve
following a LOCA. The modification to the a significant reduction in a margin of safety?
allowable value of the RWT level-low
ADAMS under Accession No. Response: No.
(function 5a) resolves a non-conservative TS ML17095A081. The purpose of Surveillances is to verify
per the guidance of Administrative Letter 98– Description of amendment request: that equipment is capable of performing its
10 ‘‘Dispositioning of Technical The amendment would revise the Nine assumed safety function. The proposed
Specifications That Are Insufficient to Assure Mile Point Nuclear Station, Unit 2, changes will only allow the performance of
Plant Safety.’’ Technical Specifications to allow the Surveillances to reestablish Operability,
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![27890 Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices
The proposed change does not involve any trusses, using a risk-informed adversely affect the design function or
changes to the operation, testing, or resolution. Accordingly, the proposed operation of any structures, systems and
maintenance of any safety-related, or change meets the criteria set forth in components important to safety.
otherwise important to safety systems. All There are no new accidents identified
Regulatory Guide (RG) 1.174, ‘‘An
systems important to safety will continue to associated with acceptance of the final
be operated and maintained within their Approach for Using Probabilistic Risk modified configuration of Unit 1 and the
design bases. The proposed changes to the Assessment [PRA] in Risk-Informed current configuration of Unit 2.
Reactivity Anomalies TS will only provide a Decisions on Plant-Specific Changes to Therefore, the proposed change does not
new, more efficient method of detecting an the Licensing Basis,’’ and the generic create the possibility of a new or different
unexpected change in core reactivity. guidance in RG 1.200, ‘‘An Approach kind of accident from any accident
Therefore, the proposed change does not for Determining the Technical previously evaluated.
create the possibility of a new or different Adequacy of Probabilistic Risk 3. Does the proposed change involve a
kind of accident from any accident significant reduction in a margin of safety?
Assessment Results for Risk-Informed
previously evaluated. Response: No.
3. Does the proposed change involve a Activities.’’
The effects of the change, DCDF [core
significant reduction in a margin of safety? Basis for proposed no significant
damage frequency] and DLERF, [large early
Response: No. hazards consideration determination: release frequency] are within the acceptance
The proposed change is to modify the As required by 10 CFR 50.91(a), the guidelines shown in Figures 4 and 5 of
method for performing the reactivity anomaly licensee has provided its analysis of the Regulatory Guide 1.174. Consequently, the
surveillance from a comparison of monitored issue of no significant hazards change does not result in a significant
to predicted control rod density to a consideration which is presented below: reduction in the margin of safety.
comparison of monitored to predicted core The containment structures and liners,
keff. The direct comparison of keff provides a 1. Does the proposed change involve a
significant increase in the probability or construction trusses, and equipment
technically superior method of calculating supported by the trusses remain fully capable
any differences in the expected core consequences of an accident previously
evaluated? of performing their specified design
reactivity. The reactivity anomaly functions as concluded by supporting
surveillance will continue to be performed at Response: No.
The probability of an accident previously engineering calculations.
the same frequency as is currently required Modifications associated with
by the TS, only the method of performing the evaluated is not changed. The containment
structures and the containment spray piping implementation of NFPA [National Fire
surveillance will be changed. Consequently, Protection Association] 805 are planned that
core reactivity assumptions made in safety and ventilation ducts attached to the
construction trusses are accident mitigation will provide protection of the reactor coolant
analyses will continue to be adequately system feed and bleed capability and result
verified. The proposed change has no impact equipment. They are not accident initiators.
The acceptance of the final configuration of in additional safety margin.
to the margin of safety. The proposed change does not affect the
Point Beach Units 1 and 2 results in a change
The NRC staff has reviewed the in core damage frequency and large early margin of safety associated with confidence
licensee’s analysis and, based on this release frequency that is within acceptance in the ability of the fission product barriers
review, it appears that the three guidelines and does not involve a significant (i.e., fuel cladding, reactor coolant system
reduction in the margin of safety. Although pressure boundary, and containment
standards of 10 CFR 50.92(c) are
failures are postulated in the PRA analysis, structure) to limit the level of radiation dose
satisfied. Therefore, the NRC staff to the public. The proposed change does not
the engineering calculations in support of the
proposes to determine that the LAR conclude that the construction trusses alter any safety analyses assumptions, safety
amendment request involves no and the associated structures/components limits, limiting safety system settings, or
significant hazards consideration. remain structurally sound in the event of a methods of operating the plant. The changes
Attorney for licensee: William Blair, design basis seismic or thermal event and do not adversely impact the reliability of
P.O. Box 14000, Juno Beach, FL 33408– there is no adverse impact or change to any equipment credited in the safety analyses.
0420. station SSC’s [structure, system, and The proposed change does not adversely
NRC Branch Chief: David J. Wrona. components] design function and there is no affect systems that respond to safely
change to accident mitigation response. shutdown the plant and to maintain the plant
NextEra Energy Point Beach, LLC, This change has no impact on station fire in a safe shutdown condition.
Docket Nos. 50–266 and 50–301, Point risk caused by a seismic event. The station will implement new seismic
Beach Nuclear Plant (Point Beach), Therefore, the proposed change does not and thermal event limits to ensure the
Units 1 and 2, Town of Two Creeks, involve a significant increase in the construction trusses and associated
Manitowoc County, Wisconsin probability or consequences of an accident equipment are inspected and/or analyzed for
previously evaluated. any event exceeding elastic stress limits to
Date of amendment request: March 2. Does the proposed change create the determine their capability to withstand a
31, 2017. A publicly available version is possibility of a new or different kind of subsequent design basis event prior to Unit
in ADAMS under Accession No. accident from any accident previously restart.
ML17090A511. evaluated? Therefore, the proposed change does not
Description of amendment request: Response: No. involve a significant reduction in a margin of
The amendments would document a The proposed change does not install any safety.
risk-informed resolution strategy to new or different type of equipment in the
plant. The proposed change does not create The NRC staff has reviewed the
resolve low risk, legacy design code any new failure modes for existing licensee’s analysis and, based on this
non-conformances associated with equipment or any new limiting single review, it appears that the three
construction trusses in the containment failures. Engineering calculations conclude standards of 10 CFR 50.92(c) are
buildings of Point Beach, Units 1 and 2. the construction trusses, equipment satisfied. Therefore, the NRC staff
asabaliauskas on DSKBBXCHB2PROD with NOTICES
The proposed license amendment supported by the trusses, and containment proposes to determine that the
request (LAR) is a risk-informed liners remain capable of withstanding design amendment request involves no
licensing basis change. The proposed basis seismic and thermal events and remain significant hazards consideration.
change is acceptance of the final capable of performing their designated design
Attorney for licensee: William Blair,
functions. Additionally, the proposed change
configuration of the construction does not involve a change in the methods Managing Attorney—Nuclear, Florida
trusses, including the attached governing normal plant operation, and all Power & Light Company, P.O. Box
containment spray piping and safety functions will continue to perform as 14000, 700 Universe Boulevard, Juno
ventilation ductwork, and the previously assumed in the accident analyses. Beach, FL 33408–0420.
containment liners/walls adjacent to the Thus, the proposed change does not NRC Branch Chief: David J. Wrona.
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![27892 Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices
for Non-Passive Reactors,’’ do not reduce the Therefore, the proposed changes do not Duke Energy Progress Inc., Docket No.
capability to meet the emergency planning involve any reduction in a margin of safety. 50–261, H. B. Robinson Steam Electric
requirements established in 10 CFR 50.47 Plant, Unit No. 2 (Robinson), Darlington
and 10 CFR [Part] 50, Appendix E. The The NRC staff has reviewed the
licensee’s analysis and, based on this County, South Carolina
proposed changes do not reduce the
functionality, performance, or capability of review, it appears that the three Duke Energy Progress, LLC, Docket No.
TVA’s ERO to respond in mitigating the standards of 10 CFR 50.92(c) are 50–400, Shearon Harris Nuclear Power
consequences of any design basis accident. satisfied. Therefore, the NRC staff Plant, Unit 1 (Harris), Wake and
The proposed changes do not adversely proposes to determine that the Chatham Counties, North Carolina
affect accident initiators or precursors nor amendment request involves no
alter the design assumptions, conditions, and Date of amendment request: August
significant hazards consideration.
configuration of the facilities or the manner 19, 2015, as supplemented by letters
in which the plants are operated and Attorney for licensee: General dated May 4, October 3, and November
maintained. The proposed change does not Counsel, Tennessee Valley Authority, 17, 2016.
adversely affect the ability of structures, 400 West Summit Hill Drive, 6A West Brief description of amendments: The
systems, and components (SSC) to perform Tower, Knoxville, TN 37902. amendments revised the Robinson
their intended safety function to mitigate the NRC Branch Chief: Benjamin G.
consequences of an initiating event within
Technical Specification (TS) 5.6.5.b and
Beasley. the Harris TS 6.9.1.6.2 to adopt the
the assumed acceptable limits. The proposed
changes do not affect the source term, III. Notice of Issuance of Amendments methodology reports DPC–NE–1008–P,
containment isolation, or radiological release to Facility Operating Licenses and Revision 0, ‘‘Nuclear Design
assumptions used in evaluating the Combined Licenses Methodology Using CASMO–5/
radiological consequences of any accident SIMULATE–3 for Westinghouse
previously evaluated. Further, the proposed During the period since publication of Reactors’’; DPC–NF–2010, Revision 3,
changes do not increase the types and the last biweekly notice, the ‘‘Nuclear Physics Methodology for
amounts of radioactive effluent that may be Commission has issued the following Reload Design’’; and DPC–NE–2011–P,
released offsite, nor significantly increase amendments. The Commission has
individual or cumulative occupational/
Revision 2, ‘‘Nuclear Design
determined for each of these Methodology Report for Core Operating
public radiation exposure. amendments that the application
Therefore, the proposed changes do not Limits of Westinghouse Reactors,’’ for
complies with the standards and application specific to Robinson and
involve a significant increase in the
probability or consequences of an accident
requirements of the Atomic Energy Act Harris.
previously evaluated. of 1954, as amended (the Act), and the Date of issuance: May 18, 2017.
2. Does the proposed amendment create Commission’s rules and regulations. Effective date: As of the date of
the possibility of a new or different kind of The Commission has made appropriate issuance and shall be implemented
accident from any accident previously findings as required by the Act and the within 120 days of issuance.
evaluated? Commission’s rules and regulations in Amendment Nos.: 253 (Robinson) and
Response: No. 10 CFR Chapter I, which are set forth in 157 (Harris). A publicly available
The proposed changes to TVA’s EAL the license amendment.
schemes to adopt the NRC-endorsed
version is in ADAMS under Accession
A notice of consideration of issuance No. ML17102A923; documents related
guidance in NEI 99–01, Revision 6, do not
involve any physical changes to plant
of amendment to facility operating to these amendments are listed in the
systems or equipment. The proposed changes license or combined license, as Safety Evaluations enclosed with the
do not involve the addition of any new plant applicable, proposed no significant amendments.
equipment. The proposed changes will not hazards consideration determination, Renewed Facility Operating License
alter the design configuration, or method of and opportunity for a hearing in Nos. DPR–23 and NPF–63: Amendments
operation of plant equipment beyond its connection with these actions, was revised the Renewed Facility Operating
normal functional capabilities. All TVA ERO published in the Federal Register as Licenses and TSs.
functions will continue to be performed as indicated. Date of initial notice in Federal
required. The proposed changes do not create Unless otherwise indicated, the Register: February 2, 2016 (81 FR
any new credible failure mechanisms,
malfunctions, or accident initiators.
Commission has determined that these 5492). The supplemental letter dated
Therefore, the proposed changes do not amendments satisfy the criteria for May 4, 2016, provided additional
create the possibility of a new or different categorical exclusion in accordance information that expanded the scope of
kind of accident from those that have been with 10 CFR 51.22. Therefore, pursuant the application as originally noticed,
previously evaluated. to 10 CFR 51.22(b), no environmental and changed the NRC staff’s original
3. Does the proposed amendment involve impact statement or environmental proposed no significant hazards
a significant reduction in a margin of safety? assessment need be prepared for these consideration determination as
Response: No. amendments. If the Commission has published in the Federal Register.
The proposed changes to TVA’s EAL prepared an environmental assessment Accordingly, the NRC published a
schemes to adopt the NRC-endorsed
guidance in NEI 99–01, Revision 6, do not
under the special circumstances second proposed no significant hazards
alter or exceed a design basis or safety limit. provision in 10 CFR 51.22(b) and has consideration determination in the
There is no change being made to safety made a determination based on that Federal Register on August 2, 2016 (81
analysis assumptions, safety limits, or assessment, it is so indicated. FR 50746). This notice superseded the
limiting safety system settings that would For further details with respect to the original notice in its entirety. The
adversely affect plant safety as a result of the
asabaliauskas on DSKBBXCHB2PROD with NOTICES
action see (1) the applications for supplemental letters dated October 3
proposed changes. There are no changes to amendment, (2) the amendment, and (3) and November 17, 2016, provided
setpoints or environmental conditions of any the Commission’s related letter, Safety additional information that clarified the
SSC or the manner in which any SSC is application, did not expand the scope
Evaluation, and/or Environmental
operated. Margins of safety are unaffected by
the proposed changes to adopt the NEI 99– Assessment as indicated. All of these beyond the second notice, and did not
01, Revision 6, EAL scheme guidance. The items can be accessed as described in change the NRC staff’s proposed no
applicable requirements of 10 CFR 50.47 and the ‘‘Obtaining Information and significant hazards consideration
10 CFR [Part] 50, Appendix E will continue Submitting Comments’’ section of this determination as published in the
to be met. document. Federal Register.
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![Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices 27893
The Commission’s related evaluations Duke Energy Progress, LLC, Docket No. Holdup Tanks’’; TS 3/4.11.2.5,
of the amendments are contained in the 50–400, Shearon Harris Nuclear Power ‘‘Explosive Gas Mixture’’; and TS 6.8.4.j,
Safety Evaluations dated May 18, 2017. Plant, Unit 1, Wake and Chatham ‘‘Gas Storage Tank Radioactivity
No significant hazards consideration Counties, North Carolina Monitoring Program.’’ The amendment
comments received: No. Date of amendment request: May 26, deleted TS Definition 1.16, ‘‘GASEOUS
2016, as supplemented by letter dated RADWASTE TREATMENT SYSTEM’’;
Duke Energy Progress, LLC, Docket Nos. TS 3/4.11.1.4, ‘‘Liquid Holdup Tanks’’;
50–325 and 50–324, Brunswick Steam December 19, 2016.
Brief description of amendment: The and TS 3/4.11.2.5, ‘‘Explosive Gas
Electric Plant, Units 1 and 2, Brunswick Mixture.’’ The amendment relocated the
County, North Carolina amendment revised the Technical
Specifications (TSs) by adding a new deleted requirements for these TSs to
Date of amendment request: Administrative Controls section to licensee control under TS 6.8.4.j, ‘‘Gas
December 21, 2015, as supplemented by establish, implement, and maintain a Storage Tank Radioactivity Monitoring
letters dated June 29, July 13, August 15, Diesel Fuel Oil Testing Program. It also Program.’’ The description for TS 6.8.4.j,
November 1, November 17, 2016, and relocated to this program the current TS ‘‘Gas Storage Tank Radioactivity
February 27, 2017. surveillance requirements (SRs) for Monitoring Program,’’ was modified to
Brief description of amendments: The evaluating diesel fuel oil, along with the include the controls for potentially
amendments adopted the approved SRs for draining, sediment removal, and explosive gas mixtures contained in the
changes to Standard Technical cleaning of each main fuel oil storage Gaseous Waste Processing System and
Specifications for General Electric tank at least once every 10 years. In the quantity of radioactivity contained
(BWR/4) [Boiling Water Reactor] Plants, addition, the licensee took an exception in unprotected outdoor liquid storage
NUREG–1433, Revision 4, to allow to NRC Regulatory Guide 1.137, tanks. The amendment relocated
relocation of specific technical Revision 1, ‘‘Fuel-Oil Systems for requirements associated with TS 3/
specification surveillance frequencies to Standby Diesel Generators,’’ to allow for 4.11.1.4 and TS 3/4.11.2.5 to the
a licensee-controlled program. The the ability to perform sampling of new licensee-controlled Plant Programs
changes are described in Technical fuel oil offsite prior to its addition to the Procedure PLP–114, ‘‘Relocated
Specification Task Force (TSTF) fuel oil storage tanks. Technical Specifications and Design
Traveler, TSTF–425, Revision 3, Date of issuance: May 24, 2017. Basis Requirements.’’
‘‘Relocate Surveillance Frequencies to Effective date: As of the date of Date of issuance: May 25, 2017.
Effective date: As of the date of
Licensee Control—RITSTF Initiative 5b’’ issuance and shall be implemented
issuance and shall be implemented
(ADAMS Package Accession No. within 120 days of issuance.
within 90 days of issuance.
ML090850642), and are described in the Amendment No.: 158. A publicly Amendment No.: 159. A publicly
Notice of Availability published in the available version is in ADAMS under available version is in ADAMS under
Federal Register on July 6, 2009 (74 FR Accession No. ML17048A184; Accession No. ML17074A672;
31996). documents related to this amendment documents related to this amendment
Date of issuance: May 24, 2017. are listed in the Safety Evaluation are listed in the Safety Evaluation
Effective date: As of the date of enclosed with the amendment. enclosed with the amendment.
issuance and shall be implemented Renewed Facility Operating License Renewed Facility Operating License
within 180 days of issuance. No. NPF–63: Amendment revised the No. NPF–63: The amendment revised
Amendment Nos.: 276 (Unit 1) and Renewed Facility Operating License and the Facility Operating License and TSs.
304 (Unit 2). A publicly available TSs. Date of initial notice in Federal
version is in ADAMS under Accession Date of initial notice in Federal Register: October 25, 2016 (81 FR
No. ML17096A129; documents related Register: October 11, 2016 (81 FR 73433). The supplemental letter dated
to these amendments are listed in the 70178). The supplemental letter dated November 4, 2016, provided additional
Safety Evaluation enclosed with the December 19, 2016, provided additional information that clarified the
amendments. information that clarified the application, did not expand the scope of
application, did not expand the scope of the application as originally noticed,
Facility Operating License Nos. DPR–
the application as originally noticed, and did not change the NRC staff’s
71 and DPR–62: Amendments revised
and did not change the NRC staff’s original proposed no significant hazards
the Facility Operating Licenses and
original proposed no significant hazards consideration determination as
Technical Specifications.
consideration determination as published in the Federal Register.
Date of initial notice in Federal published in the Federal Register.
Register: March 29, 2016 (81 FR The Commission’s related evaluation
The Commission’s related evaluation of the amendment is contained in a
17504). The supplemental letters dated of the amendment is contained in a
June 29, July 13, August 15, November Safety Evaluation dated May 25, 2017.
Safety Evaluation dated May 24, 2017. No significant hazards consideration
1, November 17, 2016, and February 27, No significant hazards consideration comments received: No.
2017, provided additional information comments received: No.
that clarified the application, did not Entergy Nuclear Operations, Inc.,
expand the scope of the application as Duke Energy Progress, LLC, Docket No. Docket No. 50–255, Palisades Nuclear
originally noticed, and did not change 50–400, Shearon Harris Nuclear Power Plant, Van Buren County, Michigan
the NRC staff’s original proposed no Plant, Unit 1, Wake and Chatham
asabaliauskas on DSKBBXCHB2PROD with NOTICES
Date of amendment request: July 11,
significant hazards consideration Counties, North Carolina
2016.
determination as published in the Date of amendment request: June 29, Brief description of amendment: The
Federal Register. 2016, as supplemented by letter dated amendment approved adoption of NRC-
The Commission’s related evaluation November 4, 2016. approved Technical Specifications Task
of the amendments is contained in a Brief description of amendment: The Force (TSTF) Standard Technical
Safety Evaluation dated May 24, 2017. amendment revised the Shearon Harris Specifications Change Traveler TSTF–
No significant hazards consideration Nuclear Power Plant, Unit 1, Technical 545, Revision 3, ‘‘TS [Technical
comments received: No. Specification (TS) 3/4.11.1.4, ‘‘Liquid Specification] Inservice Testing Program
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![27894 Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices
Removal & Clarify SR [Surveillance and 50–278, Peach Bottom Atomic Facility Operating License to reflect the
Requirement] Usage Rule Application to Power Station, Units 2 and 3, York and direct transfer of Toledo Edison
Section 5.5 Testing,’’ dated October 21, Lancaster Counties, Pennsylvania Company’s 18.26 percent leased interest
2015. Specifically, the amendment Exelon Generation Company, LLC, in Beaver Valley Power Station, Unit 2,
deleted Palisades Nuclear Plant TS Docket Nos. 50–254 and 50–265, Quad and Ohio Edison Company’s 21.66
5.5.7, ‘‘Inservice Testing Program,’’ and Cities Nuclear Power Station, Units 1 percent leased interest in Beaver Valley
added a new defined term, ‘‘INSERVICE and 2, Rock Island County, Illinois Power Station, Unit 2, from FirstEnergy
TESTING PROGRAM,’’ to the TSs. All Nuclear Operating Company to
existing references to the ‘‘Inservice Exelon Generation Company, LLC, FirstEnergy Nuclear Generation, LLC.
Testing Program,’’ in the Palisades Docket No. 50–244, R. E. Ginna Nuclear Date of issuance: May 30, 2017.
Nuclear Plant TS SRs are replaced with Power Plant, Wayne County, New York Effective date: As of the date of
‘‘INSERVICE TESTING PROGRAM’’ so Exelon Generation Company, LLC, issuance and shall be implemented
that the SRs refer to the new definition Docket No. 50–289, Three Mile Island within 30 days of issuance.
in lieu of the deleted program. Nuclear Station, Unit 1, Dauphin Amendment No.: 187. A publicly
Date of issuance: May 30, 2017. County, Pennsylvania available version is in ADAMS under
Effective date: As of the date of Date of amendment request: July 26, Accession No. ML17115A123.
issuance and shall be implemented Renewed Facility Operating License
2016, as supplemented by letter dated
within 60 days. October 6, 2016. No. NPF–73: Amendment revised the
Amendment No.: 262. A publicly Brief description of amendments: The Renewed Facility Operating License.
available version is in ADAMS under amendments revised the Inservice Date of initial notice in Federal
Accession No. ML17082A465; Testing Program requirements in each Register: January 23, 2017 (82 FR
documents related to this amendment plant’s technical specifications (TSs). 7880). The supplemental letter dated
are listed in the Safety Evaluation The changes included deleting the March 16, 2017, provided additional
enclosed with the amendment. current TS requirements for the information that clarified the
Renewed Facility Operating License Inservice Testing Program, adding a new application, did not expand the scope of
No. DPR–20: Amendment revised the defined term, ‘‘INSERVICE TESTING the application as originally noticed,
Renewed Facility Operating License and PROGRAM,’’ to the TSs, and revising and did not change the NRC staff’s
TSs. other TSs to reference this new defined original proposed no significant hazards
Date of initial notice in Federal term instead of the deleted program. consideration determination as
Register: August 30, 2016 (81 FR Date of issuance: May 26, 2017. published in the Federal Register.
59663). Effective date: As of the date of The Commission’s related evaluation
The Commission’s related evaluation issuance and shall be implemented of the amendment is contained in a
of the amendment is contained in a within 90 days of issuance. Safety Evaluation dated April 14, 2017.
Safety Evaluation dated May 30, 2017. Amendment Nos.: 191, 192, 197, 197,
No significant hazards consideration Indiana Michigan Power Company,
320, 298, 212, 254, 247, 223, 209, 227, Docket Nos. 50–315 and 50–316, Donald
comments received: No. 161, 313, 317, 266, 261, 124, and 290. C. Cook Nuclear Plant, Units 1 and 2,
Exelon Generation Company, LLC, A publicly available version is in Berrien County, Michigan
Docket Nos. STN 50–456 and STN 50– ADAMS under Accession No.
457, Braidwood Station, Units 1 and 2, ML17073A067. Documents related to Date of amendment request: July 21,
Will County, Illinois these amendments are listed in the 2016, as supplemented by letter dated
Safety Evaluations enclosed with the September 26, 2016.
Exelon Generation Company, LLC, Brief description of amendments: The
Docket Nos. STN 50–454 and STN 50– amendments.
Facility Operating License Nos.: NPF– amendments revised the Donald C. Cook
455, Byron Station, Unit Nos. 1 and 2, Nuclear Plant, Units 1 and 2, Technical
72, NPF–77, NPF–37, NPF–66, DPR–53,
Ogle County, Illinois Specification (TS) Surveillance
DPR–69, NPF–62, DPR–19, DPR–25,
Exelon Generation Company, LLC, NPF–11, NPF–18, DPR–63, NPF–69, Requirements (SRs), consistent with the
Docket Nos. 50–317 and 50–318, Calvert DPR–44, DPR–56, DPR–29, DPR–30, NRC-approved Technical Specifications
Cliffs Nuclear Power Plant, Units 1 and DPR–18, and DPR–50. Amendments Task Force (TSTF) Traveler, TSTF–545,
2, Calvert County, Maryland revised the Facility Operating Licenses Revision 3, ‘‘TS Inservice Testing
and TSs. Program Removal & Clarify SR Usage
Exelon Generation Company, LLC,
Date of initial notice in Federal Rule Application to Section 5.5
Docket No. 50–461, Clinton Power
Register: November 8, 2016 (81 FR Testing.’’ Specifically, the change
Station, Unit No. 1, DeWitt County,
78648). revised the TSs to eliminate Section
Illinois
The Commission’s related evaluations 5.5.6, ‘‘Inservice Testing Program.’’ A
Exelon Generation Company, LLC, of the amendments are contained in new defined term, ‘‘INSERVICE
Docket Nos. 50–237 and 50–249, Safety Evaluations dated May 26, 2017. TESTING PROGRAM,’’ was added to
Dresden Nuclear Power Station, Units 2 No significant hazards consideration the TS Definitions section. TS SRs that
and 3, Grundy County, Illinois comments received: No. previously referred to the Inservice
Exelon Generation Company, LLC, Testing Program from Section 5.5.6 were
FirstEnergy Nuclear Operating revised to refer to the new defined term,
Docket Nos. 50–373 and 50–374, LaSalle Company, et al., Docket No. 50–412,
asabaliauskas on DSKBBXCHB2PROD with NOTICES
County Station, Units 1 and 2, LaSalle ‘‘INSERVICE TESTING PROGRAM.’’
Beaver Valley Power Station, Unit 2, Date of issuance: May 24, 2017.
County, Illinois Beaver County, Pennsylvania Effective date: As of the date of
Exelon Generation Company, LLC, Date of amendment request: June 24, issuance and shall be implemented
Docket Nos. 50–220 and 50–410, Nine 2016, as supplemented by letters dated within 120 days of issuance.
Mile Point Nuclear Station, Units 1 and September 13, 2016; December 15, 2016; Amendment Nos.: 335 (Unit 1) and
2, Oswego County, New York and March 16, 2017. 317 (Unit 2). A publicly available
Exelon Generation Company, LLC and Brief description of amendment: The version is in ADAMS under Accession
PSEG Nuclear LLC, Docket Nos. 50–277 amendment modified the Renewed No. ML17103A106; documents related
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![Federal Register / Vol. 82, No. 116 / Monday, June 19, 2017 / Notices 27895
to these amendments are listed in the Amendment Nos.: 289 (Unit 1) and ACTION: Notice.
Safety Evaluation enclosed with the 289 (Unit 2). A publicly available
amendments. version is in ADAMS under Accession SUMMARY: This notice identifies
Renewed Facility Operating License No. ML17100A253; documents related Schedule A, B, and C appointing
Nos. DPR–58 and DPR–74: Amendments to these amendments are listed in the authorities applicable to a single agency
revised the Renewed Facility Operating Safety Evaluation enclosed with the that were established or revoked from
Licenses and TSs. amendments. January 1, 2017 to January 31, 2017.
Date of initial notice in Federal Facility Operating License Nos. NPF–
Register: September 27, 2016 (81 FR 4 and NPF–7: Amendments revised the FOR FURTHER INFORMATION CONTACT:
66307). The supplemental letter dated Facility Operating Licenses and Senior Executive Resources Services,
September 26, 2016, provided Technical Specifications. Senior Executive Service and
additional information that clarified the Date of initial notice in Federal Performance Management, Employee
application, did not expand the scope of Register: October 25, 2016 (81 FR Services, 202–606–2246.
the application as originally noticed, 73443). The supplemental letters dated
SUPPLEMENTARY INFORMATION: In
and did not change the NRC staff’s February 10, 2017; March 1, 2017; and
March 10, 2017, provided additional accordance with 5 CFR 213.103,
original proposed no significant hazards
information that clarified the Schedule A, B, and C appointing
consideration determination as
application, did not expand the scope of authorities available for use by all
published in the Federal Register.
The Commission’s related evaluation the application as originally noticed, agencies are codified in the Code of
of the amendments is contained in a and did not change the NRC staff’s Federal Regulations (CFR). Schedule A,
Safety Evaluation dated May 24, 2017. original proposed no significant hazards B, and C appointing authorities
No significant hazards consideration consideration determination as applicable to a single agency are not
comments received: No. published in the Federal Register. codified in the CFR, but the Office of
The Commission’s related evaluation Personnel Management (OPM)
Virginia Electric and Power Company, publishes a notice of agency-specific
of the amendments is contained in a
Docket Nos. 50–280 and 50–281, Surry authorities established or revoked each
Safety Evaluation dated May 31, 2017.
Power Station, Units No. 1 and No. 2, No significant hazards consideration month in the Federal Register at
Surry County, Virginia comments received: No. www.gpo.gov/fdsys/. OPM also
Date of amendment request: May 18, publishes an annual notice of the
Dated at Rockville, Maryland, this 6th day
2016, as supplemented by letters dated of June 2017. consolidated listing of all Schedule A,
February 10, 2017; March 1, 2017; and For the Nuclear Regulatory Commission.
B, and C appointing authorities, current
March 10, 2017. as of June 30, in the Federal Register.
Eric J. Benner,
Brief description of amendments: The
Deputy Director, Division of Operating Schedule A
amendments revised Technical Reactor Licensing, Office of Nuclear Reactor
Specification 3.14 ‘‘Circulating and Regulation. No schedule A authorities to report
Service Water Systems,’’ to extend the [FR Doc. 2017–12732 Filed 6–16–17; 8:45 am] during January 2017.
Allowed Outage Time for only one
BILLING CODE 7590–01–P
operable Service Water flow path to the Schedule B
Changing Pump Service Water
No schedule B authorities to report
subsystem and to the Main Control
OFFICE OF PERSONNEL during January 2017.
Room/Emergency Switchgear Room air
conditioning subsystem. MANAGEMENT
Schedule C
Date of issuance: May 31, 2017. Excepted Service
Effective date: As of the date of The following Schedule C appointing
issuance and shall be implemented AGENCY:U.S. Office of Personnel authorities were approved during
within 60 days of issuance. Management (OPM). January 2017.
Authorization
Agency name Organization name Position title Effective date
no.
COMMISSION ON CIVIL RIGHTS Office of Commissioners .............. Special Assistant (2) ..................... CC170001 01/09/2017
CC170002 01/09/2017
DEPARTMENT OF DEFENSE ..... Office of the Secretary of Defense Special Assistant (Russia, DD170034 01/04/2017
Ukraine, & Eurasia).
DEPARTMENT OF TRANSPOR- Office of the Secretary (2) ............ Advisor (2) .................................... DT170028 01/06/2017
TATION. DT170029 01/06/2017
DEPARTMENT OF THE TREAS- Office of the Secretary ................. Special Assistant .......................... DY170038 01/04/2017
URY.
The following Schedule C appointing
asabaliauskas on DSKBBXCHB2PROD with NOTICES
authorities were revoked during January
2017.
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Document Created: 2017-06-17 01:48:40
Document Modified: 2017-06-17 01:48:40
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by July 19, 2017. A request for a hearing must be filed by August 18, 2017. | |
| Contact | Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1927, email: [email protected] | |
| FR Citation | 82 FR 27882 |
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