82 FR 57469 - 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 232 (December 5, 2017)
Federal Register Volume 82, Issue 232 (December 5, 2017)
| Page Range | 57469-57478 | |
| FR Document | 2017-25901 |
[Federal Register Volume 82, Number 232 (Tuesday, December 5, 2017)] [Notices] [Pages 57469-57478] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-25901] [[Page 57469]] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0225] 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 November 7, 2017, to November 17, 2017. The last biweekly notice was published on November 21, 2017. DATES: Comments must be filed by January 4, 2018. A request for a hearing must be filed by February 5, 2018. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0225. 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: May Ma, Office of Administration, Mail Stop: OWFN-2-A13, 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: Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-2422, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2017-0225, 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-2017-0225. NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resour[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-2017-0225, 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. [[Page 57470]] 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. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not 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 [[Page 57471]] [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 by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly-available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP), Darlington County, South Carolina Date of amendment request: September 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17270A041. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to reflect the addition of a second qualified offsite power circuit. In addition, the proposed amendment requests approval to change the Updated Final Safety Analysis Report (UFSAR) to allow for the use of automatic load tap changers (LTCs) on the new (230 kilovolt (kV)) and the replacement (115kV) startup transformers. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO [Limiting Condition for Operation], Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431 [``Standard Technical Specifications-- Westinghouse Plants'']. The AC [alternating current] power systems are not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not increased. The consequences of an accident with the proposed LCO requiring two qualified offsite circuits between the offsite transmission network and the onsite emergency AC Electrical Power Distribution System to be operable are no different than the consequences of an accident in Modes 1, 2, 3, and 4 with the existing LCO that requires the single qualified offsite circuit to be operable. The additional 230kV startup transformer will improve the reliability and availability of offsite power to the emergency [[Page 57472]] buses by increasing the amount of available offsite power sources from one to two. The two qualified offsite circuits are designed to mitigate the consequences of previously evaluated accidents. The proposed change to TS 3.8.1 would not change any of the previously evaluated accidents in the UFSAR. The proposed change will also allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. The only accident previously evaluated where the probability of an accident is potentially affected by the proposed change is a loss of offsite power (LOOP). Failure of a LTC while in the automatic mode of operation that results in decreased voltage to the safety related buses could cause a LOOP if voltage decreased below the degraded grid voltage relay (DGVR) setpoint. The three postulated failure scenarios are: (1) Failure of a primary microcontroller that results in rapidly decreasing voltage supplied to the safety related buses; (2) failure of a primary microcontroller to respond to decreasing grid voltage; and (3) the backup microcontroller overrides the primary microcontroller when not required. For the first scenario, a backup microcontroller is provided for each LTC, which makes this failure unlikely. For the second scenario, operators would have ample time to address the condition utilizing identified procedures since grid voltage changes typically occur relatively slowly. In addition, the frequency of occurrence of all of these failure modes is small, based on the operating history of similar equipment at other plants. Furthermore, in all of the above potential failure modes, operators can take manual control of the LTC to mitigate the effects of the failure. Thus, the probability of a LOOP will not be significantly increased by operation of the LTCs in the automatic mode. The proposed change to allow operation of the LTCs in automatic mode has no effect on the consequences of a LOOP, since the emergency diesel generators (EDGs) provide power to safety related equipment following a LOOP. The design and function of the EDGs are not affected by the proposed change. The LTCs are each equipped with a backup microcontroller, which inhibits gross improper action of the LTC in the event of primary microcontroller failure. Additionally, the operator has procedurally identified actions available to prevent a sustained high voltage condition from occurring. Damage due to overvoltage is time-dependent, requiring a sustained high voltage condition. Therefore, damage to safety related equipment is unlikely, and the consequences of previously evaluated accidents are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO, Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431. The proposed change also will allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. All aspects of the proposed change involve electrical transformers that provide offsite power to safety-related equipment for accident mitigation. The proposed change does not alter the design, physical configuration or mode of operation of any other plant structure, system or component. No physical changes are being made to any other portion of the plant, so no new accident causal mechanisms are being introduced. The proposed change also does not result in any new mechanisms that could initiate damage to the reactor or its principal safety barriers (i.e., fuel cladding, reactor coolant system or primary containment). Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises TS 3.8.1 to reflect the addition of a second qualified offsite circuit at HBRSEP. The proposed change modifies the TS 3.8.1 LCO, Conditions, Required Actions and Completion Times to be more consistent with NUREG-1431. The new 230kV startup transformer will improve the reliability and availability of offsite power to the emergency buses by increasing the amount of available offsite power sources from one to two. Another improvement to the HBRSEP electrical system configuration as a result of the proposed change is that each emergency bus will be normally aligned to independent startup sources and will not require a fast bus transfer on a unit trip. This reduces the risk of loss of power to the emergency buses caused by power transfer and/or equipment failures. The margin of safety is increased with the proposed change to revise TS 3.8.1 to reflect the additional qualified offsite circuit. The proposed change will also allow operation of the LTCs on the 115kV and 230kV startup transformers in automatic mode. The inputs or assumptions of any of the analyses that demonstrate the integrity of the fuel cladding, reactor coolant system or containment during accident conditions are unaffected by this proposed change. The allowable values for the degraded voltage protection function are unchanged and will continue to ensure that the degraded voltage protection function actuates when required, but does not actuate prematurely to unnecessarily transfer safety related loads from offsite power to the EDGs. Automatic operation of the LTCs increases the margin of safety by reducing the potential for transferring loads to the EDGs during an undervoltage or overvoltage event on the offsite power sources. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202. NRC Branch Chief: Undine Shoop. Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP), Van Buren County, Michigan Date of amendment request: November 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17306A086. Description of amendment request: The proposed amendment would revise the PNP renewed facility operating license (RFOL) to change the full compliance implementation date for the fire protection program transition license condition. Specifically, the licensee is requesting additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c), ``National Fire Protection Association Standard NFPA 805.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to the PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents, and have no impact on the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of [[Page 57473]] accident from any accident previously evaluated? Response: No. The proposed changes to the PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes to the PNP RFOL to change the full compliance implementation date for the fire protection program transition license condition to allow additional time for completion of the required modifications necessary to achieve full compliance with 10 CFR 50.48(c) is administrative in nature. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William Glew, Associate General Counsel Nuclear, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 10601. NRC Branch Chief: David J. Wrona. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1 (ANO-1), Pope County, Arkansas Date of amendment request: October 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17275A910. Description of amendment request: The amendment would revise the ANO-1 Technical Specification (TS) 3.7.5, ``Emergency Feedwater (EFW) System,'' Bases to stipulate the conditions in which the TS 3.7.5, Condition A, 7-day Completion Time should apply to the ANO-1 turbine- driven EFW pump steam supply valves. 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 EFW system is not an initiator of any design basis accident or event and, therefore, the proposed change does not increase the probability of any accident previously evaluated. The proposed change to clarify the conditions in which the current 7-day Completion Time for an inoperable steam supply path to turbine- driven EFW pump does not change the response of the plant to any accidents, since single failure criterion is not applicable when complying with associated TS Actions. The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed change does not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed change does not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/ public radiation exposures. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not result in a change in the manner in which the EFW system provides plant protection. Absent a single failure (which is not assumed while in compliance with TS Actions), the EFW system will continue to supply water to the Steam Generators (SGs) to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the SGs, as required. There are no design changes associated with the proposed change. The change to the associated TS Bases does not change any existing accident scenarios, nor create any new or different accident scenarios. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the change clarifies the application of the current 7-day Completion Time for an inoperable steam supply path to the turbine-driven EFW pump and does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions, which does not assume an EFW system single failure when complying with TS Actions, and current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. The proposed change will not result in plant operation in a configuration outside the design basis. The associated TS will continue to limit the time in which one steam supply path to the turbine-driven EFW pump may be inoperable. Therefore, this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. 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: July 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17209A755. Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix A, plant-specific Technical Specifications (TS) to make them consistent with the remainder of the design licensing basis and the TS. Specifically, the requested amendment proposes changes to COL Appendix A, the Technical Specification updates for reactivity controls and other [[Page 57474]] miscellaneous changes, and Updated Final Safety Analysis Report (UFSAR) information in various locations. 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 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 change does not involve a physical alteration of the plant or a change in the methods governing normal plant operations. The change applies to a Diverse Actuation System (DAS) Manual Controls Mode 6 note for operability of the Automatic Depressurization System (ADS) Stage 4 valves that involves revising the note from reactor internals in place to upper internals in place. In accordance with Limiting Condition for Operation (LCO) 3.4.13 ADS--Shutdown, Reactor Coolant System (RCS) Open Applicability and TS 3.3.9, Engineered Safeguards Actuation System Instrumentation, Function 7, the ADS Stage 4 valves are not required to be operable in MODE 6 with the upper internals removed. However, the reactor internals would still be present. The change involves clarification of the note (with no change in required system or device function), such that the appropriate configuration in Mode 6 would be in place and would not conflict with TS 3.4.13 or TS 3.3.9. The revised note previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident as a result of the revised note and associated requirements and actions are no different than the consequences of the same accident during the existing ones. As a result, the consequences are not affected by this change. The proposed change does not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change involves revising the existing LCO 3.1.4 operability to be applicable to Rod Cluster Control Assemblies (RCCAs)with accompanying changes in actions and surveillance requirements (with no change in required system or device function), such that more appropriate, albeit less restrictive, actions would be applied. The proposed change does not involve a physical alteration of the plant as described in the UFSAR. No new equipment is being introduced, and equipment is not being operated in a new or different manner. There are no set points, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR as a result of this change. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change will not reduce a margin of safety because it has no effect on any assumption of the safety analyses. While the LCO 3.1.4 for Rod Group Alignment Limits is made less restrictive by eliminating the worth of the [Gray Rod Cluster Assemblies (GRCAs)] in MODES 1 and 2 with k eff >=1, no credit is taken in the current design basis for including their trip reactivity worth. As such, there is no significant reduction in a margin of safety. Therefore, the requested 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. Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50- 425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: September 12, 2017. A publicly-available version is in ADAMS under Accession No. ML17257A177. Description of amendment request: The amendments would revise Technical Specification (TS) 5.5.17, ``Containment Leakage Rate Testing Program,'' for the Vogtle Electric Generating Plant, Units 1 and 2, to (1) increase the existing Type A integrated leakage rate test interval from 10 to 15 years, (2) extend the Type C containment isolation valve leaking testing to a 75-month frequency, (3) adopt the use of American National Standards Institute/American Nuclear Society 56.8-2002, ``Containment System Leakage Testing Requirements,'' and (4) adopt a more conservative grace interval of 9 months for Type A, B, and C tests in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J.'' 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 the revision of Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Technical Specification (TS) Section 5.5.17, ``Primary Containment Leakage Rate Testing Program,'' to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years, and the extension of the Type C local leakage rate test (LLRT) 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 current 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 extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events (IE) probabilistic risk analysis (PRA) is 1.79E-03 person-rem/year for Unit 1 and Unit 2. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A states that a very small population is defined as an increase of <=1.0 person-rem per year or <=1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94- 01 and EPRI Report No. 1009325. Moreover, the risk [[Page 57475]] impact when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. In addition, as documented in NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' dated September 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The VEGP 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. The LLRT 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 exceptions previously granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS 5.5.17, Containment Leakage Rate Testing Program, involves the extension of the VEGP 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 do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. The proposed amendment also deletes exceptions previously granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.17 involves the extension of the VEGP 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 leaktightness 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 only the extension of the interval between Type A containment leak rate tests and Type C tests for VEGP. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusions 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 XI 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 under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to allow one-time extensions of the ILRT test frequency for VEGP. This exception was for an activity that would have already taken place by the time this amendment is approved; therefore, the deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any margin of safety as a result of this administrative change. 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35242. NRC Branch Chief: Michael T. Markley. STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499, South Texas Project (STP), Units 1 and 2, Matagorda County, Texas Date of amendment request: September 18, 2017. A publicly-available version is in ADAMS under Accession No. ML17261B272. Description of amendment request: The amendment would relocate the defined core plane regions where the radial peaking factor limits are not applicable, from Technical Specification (TS) 4.2.2.2.f to the Core Operating Limits Reports (COLR) for STP Units 1 and 2. The amendment would also revise the COLR Administrative Controls TS to add exclusion zones to the list of limits found in the COLRs, and to revise the description of the methodology used to determine the values. In addition, the proposed amendment requests administrative changes to the TSs. 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 relocation of the Fxy exclusion zones to the COLRs has no impact on the accidents analyzed in the STPNOC UFSAR [Updated Final Safety Analysis Report] and is not an accident initiator. Since the change does not impact any conditions that would initiate an accident, the probability or consequences of previously analyzed events is not increased. The proposed amendment does not change the actions to be taken if a core operating limit is exceeded and there are no physical changes associated with this proposed amendment. For each core reload, each accident analysis addressed in the STP UFSAR will continue to be examined with respect to changes in the cycle-dependent parameters, which are obtained from the use of NRC-approved reload design methodologies, to [[Page 57476]] ensure that the transient evaluation of new reloads are bounded by previously accepted analyses. This examination, which will be conducted per the requirements of 10 CFR 50.59, will ensure that future core reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated. Therefore, there is no impact to the probability or consequences of an accident previously evaluated due to the proposed change. [The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.] 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The relocation of the Fxy exclusion zone details from the Technical Specifications to the COLRs will not create the possibility of a new or different kind of accident from any accident previously evaluated. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. No new operator actions are created as a result of the proposed change. The cycle-specific variables are determined using the NRC approved methods and the COLRs are submitted to the NRC to allow the staff to continue to trend the values of these limits. The Technical Specifications will continue to require operation within the core operating limits and appropriate actions will be required if these limits are exceeded. The relocation of the Fxy exclusion zones to the COLRs has no impact on the accidents analyzed in the STPNOC Updated Final Safety Analysis Report (UFSAR) and is not an accident initiator. Since this change does not impact any conditions that would initiate an accident, there is no possibility of a new or different kind of accident resulting from this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. [The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.] 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The relocation of the Fxy exclusion zone details from the Technical Specifications to the COLRs will not affect the margin of safety. The margin of safety presently provided by the Technical Specifications remains unchanged. They will be incorporated into the COLR which is submitted to the NRC, therefore appropriate measures exist to control the values of these limits. The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. STPNOC will continue to confirm all safety analysis limits remain bounding on a cycle- specific basis using an NRC-approved methodology. Each core reload will involve a Reload Safety Evaluation to assure that operation of the unit within the cycle specific limits will not involve a significant reduction in the margin of safety. The proposed amendment does not affect the design of the facility or system operating parameters, does not physically alter safety-related systems and does not affect the method in which safety-related systems perform their functions. Therefore, the proposed change does not impact margin of safety. [The licensee stated that the administrative changes proposed to the TSs do not impact the operation of the facility in a manner that involves significant hazards considerations.] The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Attorney for licensee: Kym Harshaw, General Counsel, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, TX, 77483. NRC Branch Chief: Robert J. Pascarelli. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of amendment request: October 26, 2016, as supplemented by letters dated February 16, July 17, August 8, September 27, October 3, and November 8, 2017. Brief description of amendments: The amendments revised Technical Specification (TS) 5.5.13, ``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, to change the documents used by LSCS to implement the performance-based leakage testing program, and to delete the information regarding the performance of the next LSCS Type A tests to be performed. Additionally, the amendments deleted Conditions 2.D.(e) and 2.D.(c), respectively, of the LSCS Unit 1 and Unit 2 Renewed Facility Operating Licenses regarding conducting the third Type A test of each 10-year service period when the plant is shut down for the 10-year inservice inspection. Date of issuance: November 16, 2017. Effective date: As of the date of its issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 226 (Unit 1) and 212 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17283A085; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the TSs and Renewed Facility Operating Licenses. Date of initial notice in Federal Register: February 14, 2017 (82 FR 10597). The supplemental letters dated February 16, July 17, August 8, September 27, October 3, and November 8, 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. [[Page 57477]] The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 16, 2017. No significant hazards consideration comments received: No. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: February 17, 2017, as supplemented by letters dated March 20, 2017; July 13, 2017; August 8, 2017; August 30, 2017; and September 15, 2017. Brief description of amendments: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications to implement a measurement uncertainty recapture power uprate. Specifically, the amendments authorized an increase in the maximum licensed thermal power level from 3,951 megawatts thermal to 4,016 megawatts thermal, which is an increase of approximately 1.66 percent. Date of issuance: November 15, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendments Nos.: 316 (Unit 2) and 319 (Unit 3). A publicly- available version is in ADAMS under Accession No. ML17286A013; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: May 2, 2017 (82 FR 20497). The supplemental letters dated March 20, 2017; July 13, 2017; August 8, 2017; August 30, 2017; and September 15, 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 November 15, 2017. No significant hazards consideration comments received: No. Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1 (FCS), Washington County, Nebraska Date of amendment request: October 25, 2016, as supplemented by letter dated September 25, 2017. Brief description of amendment: The amendment revised the FCS Updated Safety Analysis Report to change the structural design methodology for the Auxiliary Building at FCS. Specifically, the amendment made the following changes: (1) Use of the ultimate strength design method from the industry standard American Concrete Institute (ACI) 318-63, ``Publication SP-10, Commentary on Building Code Requirements for Reinforced Concrete,'' for normal operating/service conditions for future designs and evaluations; (2) use higher concrete compressive strength values for Class B concrete, based on original strength test data; (3) use higher reinforcing steel yield strength values, based on original strength test data; and (4) make minor clarifications, including adding a definition of control fluids to the dead load section of the Updated Safety Analysis Report. Date of issuance: November 17, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 293. A publicly-available version is in ADAMS under Accession No. ML17278A607; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-40: The amendment revised the Emergency Plan and Emergency Action Level Scheme. Date of initial notice in Federal Register: January 17, 2017 (82 FR 4930). The supplemental letter dated September 25, 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 November 17, 2017. No significant hazards consideration comments received: No. PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of amendment request: November 17, 2016, as supplemented by letters dated August 7, 2017, and October 18, 2017. Brief description of amendments: The amendments revised Technical Specification requirements regarding accident monitoring instrumentation. Specifically, the amendments modified the list of instruments required to be operable based on implementation of Regulatory Guide 1.97, Revision 2, ``Instrumentation for Light-Water- Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.'' In addition, allowed outage times and required actions for inoperable accident monitoring instrumentation channels have been revised to be consistent with NUREG-1431, Revision 4.0, ``Standard Technical Specifications--Westinghouse Plants.'' Date of issuance: November 14, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 320 (Unit 1) and 301(Unit 2). A publicly-available version is in ADAMS under Accession No. ML17227A016; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: January 17, 2017 (82 FR 4931). The supplemental letters dated August 7, 2017, and October 18, 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 November 14, 2017. No significant hazards consideration comments received: No. PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: October 7, 2017, as supplemented by letters dated March 27, 2017, and July 13, 2017. [[Page 57478]] Brief description of amendment: The amendment modified Hope Creek Generating Station Technical Specification 6.8.4.f, ``Primary Containment Leakage Rate Testing Program,'' to extend the Type A reactor containment pressure test interval from one test in 10 years to one test in 15 years, and extend the Type C test interval up to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergency conditions. Date of issuance: November 8, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 207. A publicly-available version is in ADAMS under Accession No. ML17291A209; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-57: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: December 20, 2016 (81 FR 92869). The supplemental letters dated March 27, 2017, and July 13, 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 November 8, 2017. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama Date of amendment request: January 17, 2017, as supplemented by letter dated June 29, 2017. Brief description of amendments: The amendments change technical specifications (TSs) consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF- 545, Revision 3, ``TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,'' and TSTF-299, Revision 0, ``Administrative Controls Program 5.5.2.b Test Interval and Exception.'' Date of issuance: November 8, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 301 (Unit 1), 325 (Unit 2), and 285 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17277A207; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: April 25, 2017 (82 FR 19106). The supplemental letter dated June 29, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 8, 2017. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee Date of amendment request: March 28, 2017. Brief description of amendment: The amendment revised the completion date for License Condition 2.C.(5) for Watts Bar Nuclear Plant, Unit 2, regarding the completion of action to resolve the issues identified in Bulletin 2012-01, ``Design Vulnerability in Electric Power System'' (ADAMS Accession No. ML12074A115), from December 31, 2017, to December 31, 2018, to align with the remainder of the Tennessee Valley Authority fleet and with the nuclear industry. Date of issuance: November 6, 2017. Effective date: As of the date of issuance and shall be implemented within 15 days of issuance. Amendment No.: 17. A publicly-available version is in ADAMS under Accession No. ML17258A328; documents related to this amendment is listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-96: Amendment revised the Facility Operating License. Date of initial notice in Federal Register: July 5, 2017 (82 FR 31103). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 6, 2017. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 27th day of November 2017. For the Nuclear Regulatory Commission. Kathryn M. Brock, Acting Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2017-25901 Filed 12-4-17; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices 57469
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request for a hearing from any person. please contact the NRC’s Public
This biweekly notice includes all amendment request is shown below.
Document Room (PDR) reference staff at
notices of amendments issued, or 1–800–397–4209, 301–415–4737, or by The Commission is seeking public
proposed to be issued, from November email to pdr.resource@nrc.gov. The comments on this proposed
7, 2017, to November 17, 2017. The last ADAMS accession number for each determination. Any comments received
biweekly notice was published on document referenced (if it is available in within 30 days after the date of
November 21, 2017. ADAMS) is provided the first time that publication of this notice will be
DATES: Comments must be filed by it is mentioned in this document. considered in making any final
January 4, 2018. A request for a hearing • NRC’s PDR: You may examine and determination.
must be filed by February 5, 2018. purchase copies of public documents at Normally, the Commission will not
ADDRESSES: You may submit comments the NRC’s PDR, Room O1–F21, One issue the amendment until the
by any of the following methods (unless White Flint North, 11555 Rockville expiration of 60 days after the date of
this document describes a different Pike, Rockville, Maryland 20852. publication of this notice. The
method for submitting comments on a Commission may issue the license
B. Submitting Comments
specific subject): amendment before expiration of the 60-
• Federal Rulemaking Web site: Go to Please include Docket ID NRC–2017–
day period provided that its final
http://www.regulations.gov and search 0225, facility name, unit number(s),
determination is that the amendment
for Docket ID NRC–2017–0225. Address plant docket number, application date,
involves no significant hazards
questions about NRC dockets to Carol and subject in your comment
submission. consideration. In addition, the
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For The NRC cautions you not to include Commission may issue the amendment
technical questions, contact the identifying or contact information that prior to the expiration of the 30-day
individual listed in the FOR FURTHER you do not want to be publicly comment period if circumstances
INFORMATION CONTACT section of this disclosed in your comment submission. change during the 30-day comment
document. The NRC will post all comment period such that failure to act in a
• Mail comments to: May Ma, Office submissions at http:// timely way would result, for example in
of Administration, Mail Stop: OWFN–2– www.regulations.gov as well as enter the derating or shutdown of the facility. If
A13, U.S. Nuclear Regulatory comment submissions into ADAMS. the Commission takes action prior to the
Commission, Washington, DC 20555– The NRC does not routinely edit expiration of either the comment period
0001. comment submissions to remove or the notice period, it will publish in
For additional direction on obtaining identifying or contact information. the Federal Register a notice of
sradovich on DSK3GMQ082PROD with NOTICES
information and submitting comments, If you are requesting or aggregating issuance. If the Commission makes a
see ‘‘Obtaining Information and comments from other persons for final no significant hazards
Submitting Comments’’ in the submission to the NRC, then you should consideration determination, any
SUPPLEMENTARY INFORMATION section of inform those persons not to include hearing will take place after issuance.
this document. identifying or contact information that The Commission expects that the need
FOR FURTHER INFORMATION CONTACT: they do not want to be publicly to take this action will occur very
Paula Blechman, Office of Nuclear disclosed in their comment submission. infrequently.
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![Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices 57471
hearing.docket@nrc.gov, or by telephone Electronic Filing Help Desk is available participants are requested not to include
at 301–415–1677, to (1) request a digital between 9 a.m. and 6 p.m., Eastern copyrighted materials in their
identification (ID) certificate, which Time, Monday through Friday, submission.
allows the participant (or its counsel or excluding government holidays. For further details with respect to
representative) to digitally sign Participants who believe that they these license amendment applications,
submissions and access the E-Filing have a good cause for not submitting see the application for amendment
system for any proceeding in which it documents electronically must file an which is available for public inspection
is participating; and (2) advise the exemption request, in accordance with in ADAMS and at the NRC’s PDR. For
Secretary that the participant will be 10 CFR 2.302(g), with their initial paper additional direction on accessing
submitting a petition or other filing stating why there is good cause for information related to this document,
adjudicatory document (even in not filing electronically and requesting see the ‘‘Obtaining Information and
instances in which the participant, or its authorization to continue to submit Submitting Comments’’ section of this
counsel or representative, already holds documents in paper format. Such filings document.
an NRC-issued digital ID certificate). must be submitted by: (1) First class
Based upon this information, the mail addressed to the Office of the Duke Energy Progress, LLC, Docket No.
Secretary will establish an electronic Secretary of the Commission, U.S. 50–261, H. B. Robinson Steam Electric
docket for the hearing in this proceeding Nuclear Regulatory Commission, Plant, Unit No. 2 (HBRSEP), Darlington
if the Secretary has not already Washington, DC 20555–0001, Attention: County, South Carolina
established an electronic docket. Rulemaking and Adjudications Staff; or Date of amendment request:
Information about applying for a (2) courier, express mail, or expedited September 27, 2017. A publicly-
digital ID certificate is available on the delivery service to the Office of the available version is in ADAMS under
NRC’s public Web site at http:// Secretary, 11555 Rockville Pike, Accession No. ML17270A041.
www.nrc.gov/site-help/e-submittals/ Rockville, Maryland, 20852, Attention: Description of amendment request:
getting-started.html. Once a participant Rulemaking and Adjudications Staff. The proposed amendment would revise
has obtained a digital ID certificate and Participants filing adjudicatory the Technical Specifications (TSs) to
a docket has been created, the documents in this manner are reflect the addition of a second qualified
participant can then submit responsible for serving the document on offsite power circuit. In addition, the
adjudicatory documents. Submissions all other participants. Filing is proposed amendment requests approval
must be in Portable Document Format considered complete by first-class mail to change the Updated Final Safety
(PDF). Additional guidance on PDF as of the time of deposit in the mail, or Analysis Report (UFSAR) to allow for
submissions is available on the NRC’s by courier, express mail, or expedited the use of automatic load tap changers
public Web site at http://www.nrc.gov/ delivery service upon depositing the (LTCs) on the new (230 kilovolt (kV))
site-help/electronic-sub-ref-mat.html. A document with the provider of the and the replacement (115kV) startup
filing is considered complete at the time service. A presiding officer, having transformers.
the document is submitted through the granted an exemption request from Basis for proposed no significant
NRC’s E-Filing system. To be timely, an using E-Filing, may require a participant hazards consideration determination:
electronic filing must be submitted to or party to use E-Filing if the presiding As required by 10 CFR 50.91(a), the
the E-Filing system no later than 11:59 officer subsequently determines that the licensee has provided its analysis of the
p.m. Eastern Time on the due date. reason for granting the exemption from issue of no significant hazards
Upon receipt of a transmission, the E- use of E-Filing no longer exists. consideration, which is presented
Filing system time-stamps the document Documents submitted in adjudicatory below:
and sends the submitter an email notice proceedings will appear in the NRC’s
electronic hearing docket which is 1. Does the proposed change involve a
confirming receipt of the document. The
significant increase in the probability or
E-Filing system also distributes an email available to the public at https://
consequences of an accident previously
notice that provides access to the adams.nrc.gov/ehd, unless excluded evaluated?
document to the NRC’s Office of the pursuant to an order of the Commission Response: No.
General Counsel and any others who or the presiding officer. If you do not The proposed change revises TS 3.8.1 to
have advised the Office of the Secretary have an NRC-issued digital ID certificate reflect the addition of a second qualified
that they wish to participate in the as described above, click cancel when offsite circuit at HBRSEP. The proposed
proceeding, so that the filer need not the link requests certificates and you change modifies the TS 3.8.1 LCO [Limiting
serve the document on those will be automatically directed to the Condition for Operation], Conditions,
participants separately. Therefore, NRC’s electronic hearing dockets where Required Actions and Completion Times to
you will be able to access any publicly- be more consistent with NUREG–1431
applicants and other participants (or
[‘‘Standard Technical Specifications—
their counsel or representative) must available documents in a particular
Westinghouse Plants’’]. The AC [alternating
apply for and receive a digital ID hearing docket. Participants are current] power systems are not an initiator of
certificate before adjudicatory requested not to include personal any accident previously evaluated. As a
documents are filed so that they can privacy information, such as social result, the probability of an accident
obtain access to the documents via the security numbers, home addresses, or previously evaluated is not increased. The
E-Filing system. personal phone numbers in their filings, consequences of an accident with the
A person filing electronically using unless an NRC regulation or other law proposed LCO requiring two qualified offsite
the NRC’s adjudicatory E-Filing system requires submission of such circuits between the offsite transmission
network and the onsite emergency AC
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may seek assistance by contacting the information. For example, in some
NRC’s Electronic Filing Help Desk instances, individuals provide home Electrical Power Distribution System to be
operable are no different than the
through the ‘‘Contact Us’’ link located addresses in order to demonstrate
consequences of an accident in Modes 1, 2,
on the NRC’s public Web site at http:// proximity to a facility or site. With 3, and 4 with the existing LCO that requires
www.nrc.gov/site-help/e- respect to copyrighted works, except for the single qualified offsite circuit to be
submittals.html, by email to limited excerpts that serve the purpose operable. The additional 230kV startup
MSHD.Resource@nrc.gov, or by a toll- of the adjudicatory filings and would transformer will improve the reliability and
free call at 1–866–672–7640. The NRC constitute a Fair Use application, availability of offsite power to the emergency
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![57474 Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices
miscellaneous changes, and Updated physical alteration of the plant as described Society 56.8–2002, ‘‘Containment
Final Safety Analysis Report (UFSAR) in the UFSAR. No new equipment is being System Leakage Testing Requirements,’’
information in various locations. introduced, and equipment is not being and (4) adopt a more conservative grace
Basis for proposed no significant operated in a new or different manner. There
are no set points, at which protective or
interval of 9 months for Type A, B, and
hazards consideration determination: mitigative actions are initiated, affected by C tests in accordance with Nuclear
As required by 10 CFR 50.91(a), the this change. This change will not alter the Energy Institute (NEI) 94–01, Revision
licensee has provided its analysis of the manner in which equipment operation is 3–A, ‘‘Industry Guideline for
issue of no significant hazards initiated, nor will the function demands on Implementing Performance-Based
consideration, which is presented below credited equipment be changed. No change is Option of 10 CFR part 50, Appendix J.’’
with NRC staff edits in square brackets: being made to the procedures relied upon to Basis for proposed no significant
respond to an off-normal event as described hazards consideration determination:
1. Does the proposed amendment involve in the UFSAR as a result of this change. As
a significant increase in the probability or As required by 10 CFR 50.91(a), the
such, no new failure modes are being
consequences of an accident previously licensee has provided its analysis of the
introduced. The change does not alter
evaluated? assumptions made in the safety analysis and issue of no significant hazards
Response: No. consideration, which is presented
licensing basis. Therefore, this change does
The proposed change does not involve a below:
not create the possibility of a new or different
physical alteration of the plant or a change
kind of accident from any accident 1. Does the proposed change involve a
in the methods governing normal plant
previously evaluated. significant increase in the probability or
operations. The change applies to a Diverse
3. Does the proposed amendment involve consequences of an accident previously
Actuation System (DAS) Manual Controls
a significant reduction in a margin of safety? evaluated?
Mode 6 note for operability of the Automatic
Response: No. Response: No.
Depressurization System (ADS) Stage 4
The proposed change will not reduce a The proposed activity involves the revision
valves that involves revising the note from
margin of safety because it has no effect on of Vogtle Electric Generating Plant (VEGP),
reactor internals in place to upper internals
in place. In accordance with Limiting any assumption of the safety analyses. While Units 1 and 2, Technical Specification (TS)
Condition for Operation (LCO) 3.4.13 ADS— the LCO 3.1.4 for Rod Group Alignment Section 5.5.17, ‘‘Primary Containment
Shutdown, Reactor Coolant System (RCS) Limits is made less restrictive by eliminating Leakage Rate Testing Program,’’ to allow the
Open Applicability and TS 3.3.9, Engineered the worth of the [Gray Rod Cluster extension of the Type A integrated leakage
Safeguards Actuation System Assemblies (GRCAs)] in MODES 1 and 2 with rate test (ILRT) containment test interval to
Instrumentation, Function 7, the ADS Stage keff ≥1, no credit is taken in the current 15 years, and the extension of the Type C
4 valves are not required to be operable in design basis for including their trip reactivity local leakage rate test (LLRT) interval to 75
MODE 6 with the upper internals removed. worth. As such, there is no significant months. The current Type A test interval of
However, the reactor internals would still be reduction in a margin of safety. Therefore, 120 months (10 years) would be extended on
present. The change involves clarification of the requested amendment does not involve a a permanent basis to no longer than 15 years
the note (with no change in required system significant reduction in a margin of safety. from the last Type A test. The current Type
or device function), such that the appropriate The NRC staff has reviewed the C test interval of 60 months for selected
configuration in Mode 6 would be in place components would be extended on a
licensee’s analysis and, based on this performance basis to no longer than 75
and would not conflict with TS 3.4.13 or TS review, it appears that the three
3.3.9. The revised note previously evaluated. months. Extensions of up to nine months
As a result, the probability of an accident standards of 10 CFR 50.92(c) are (total maximum interval of 84 months for
previously evaluated is not affected. satisfied. Therefore, the NRC staff Type C tests) are permissible only for non-
The consequences of an accident as a result proposes to determine that the routine emergent conditions.
of the revised note and associated amendment request involves no The proposed extensions do not involve
requirements and actions are no different significant hazards consideration. either a physical change to the plant or a
than the consequences of the same accident Attorney for licensee: M. Stanford change in the manner in which the plant is
during the existing ones. As a result, the operated or controlled. The containment is
Blanton, Balch & Bingham LLP, 1710
consequences are not affected by this change. designed to provide an essentially leak tight
Sixth Avenue North, Birmingham, AL barrier against the uncontrolled release of
The proposed change does not alter or
35203–2015. radioactivity to the environment for
prevent the ability of structures, systems, and
components from performing their intended
NRC Branch Chief: Jennifer Dixon- postulated accidents. As such, the
function to mitigate the consequences of an Herrity. containment and the testing requirements
initiating event within the assumed invoked to periodically demonstrate the
Southern Nuclear Operating Company, integrity of the containment exist to ensure
acceptance limits. The proposed change does Inc., Docket Nos. 50–424 and 50–425,
not affect the source term, containment the plant’s ability to mitigate the
isolation, or radiological release assumptions Vogtle Electric Generating Plant, Units 1 consequences of an accident, and do not
used in evaluating the radiological and 2, Burke County, Georgia involve the prevention or identification of
consequences of an accident previously Date of amendment request: any precursors of an accident.
evaluated. Therefore, this change does not September 12, 2017. A publicly- The change in Type A test frequency to
involve a significant increase in the once-per-fifteen years, measured as an
available version is in ADAMS under increase to the total integrated plant risk for
probability or consequences of an accident
previously evaluated.
Accession No. ML17257A177. those accident sequences influenced by Type
2. Does the proposed amendment create Description of amendment request: A testing, based on the internal events (IE)
the possibility of a new or different kind of The amendments would revise probabilistic risk analysis (PRA) is 1.79E–03
accident from any accident previously Technical Specification (TS) 5.5.17, person-rem/year for Unit 1 and Unit 2.
evaluated? ‘‘Containment Leakage Rate Testing Electric Power Research Institute (EPRI)
Response: No. Program,’’ for the Vogtle Electric Report No. 1009325, Revision 2–A states that
The proposed change involves revising the Generating Plant, Units 1 and 2, to (1) a very small population is defined as an
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existing LCO 3.1.4 operability to be increase the existing Type A integrated increase of ≤1.0 person-rem per year or ≤1%
applicable to Rod Cluster Control Assemblies of the total population dose, whichever is
leakage rate test interval from 10 to 15
(RCCAs)with accompanying changes in less restrictive for the risk impact assessment
actions and surveillance requirements (with years, (2) extend the Type C of the extended ILRT intervals. This is
no change in required system or device containment isolation valve leaking consistent with the Nuclear Regulatory
function), such that more appropriate, albeit testing to a 75-month frequency, (3) Commission (NRC) Final Safety Evaluation
less restrictive, actions would be applied. adopt the use of American National for Nuclear Energy Institute (NEI) 94–01 and
The proposed change does not involve a Standards Institute/American Nuclear EPRI Report No. 1009325. Moreover, the risk
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![Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices 57475
impact when compared to other severe Amendment Nos. 130 (VEGP–1) and 108 The NRC staff has reviewed the
accident risks is negligible. Therefore, this (VEGP–2), to allow one-time extensions of licensee’s analysis and, based on this
proposed extension does not involve a the ILRT test frequency for VEGP. These review, it appears that the three
significant increase in the probability of an exceptions were for activities that would
standards of 10 CFR 50.92(c) are
accident previously evaluated. have already taken place by the time this
In addition, as documented in NUREG– amendment is approved; therefore, their satisfied. Therefore, the NRC staff
1493, ‘‘Performance-Based Containment deletion is solely an administrative action proposes to determine that the
Leak-Test Program,’’ dated September 1995, that does not result in any change in how the amendment request involves no
Types B and C tests have identified a very unit is operated. significant hazards consideration.
large percentage of containment leakage Therefore, the proposed change does not Attorney for licensee: Jennifer M.
paths, and the percentage of containment create the possibility of a new or different Buettner, Associate General Counsel,
leakage paths that are detected only by Type kind of accident from any previously Southern Nuclear Operating Company,
A testing is very small. The VEGP Type A evaluated. 40 Inverness Center Parkway,
test history supports this conclusion. 3. Does the proposed change involve a
The integrity of the containment is subject significant reduction in a margin of safety?
Birmingham, AL 35242.
to two types of failure mechanisms that can Response: No. NRC Branch Chief: Michael T.
be categorized as: (1) Activity based, and (2) The proposed amendment to TS 5.5.17 Markley.
time based. Activity-based failure involves the extension of the VEGP Type A STP Nuclear Operating Company
mechanisms are defined as degradation due containment test interval to 15 years and the
(STPNOC), Docket Nos. 50–498 and 50–
to system and/or component modifications or extension of the Type C test interval to 75
maintenance. The LLRT requirements and months for selected components. This 499, South Texas Project (STP), Units 1
administrative controls such as configuration amendment does not alter the manner in and 2, Matagorda County, Texas
management and procedural requirements for which safety limits, limiting safety system set Date of amendment request:
system restoration ensure that containment points, or limiting conditions for operation September 18, 2017. A publicly-
integrity is not degraded by plant are determined. The specific requirements available version is in ADAMS under
modifications or maintenance activities. The and conditions of the TS Containment Leak
Rate Testing Program exist to ensure that the
Accession No. ML17261B272.
design and construction requirements of the
containment combined with the containment degree of containment structural integrity
Description of amendment request:
inspections performed in accordance with and leaktightness that is considered in the The amendment would relocate the
American Society of Mechanical Engineers plant safety analysis is maintained. The defined core plane regions where the
(ASME) Section XI, and TS requirements overall containment leak rate limit specified radial peaking factor limits are not
serve to provide a high degree of assurance by TS is maintained. applicable, from Technical Specification
that the containment would not degrade in a The proposed change involves only the (TS) 4.2.2.2.f to the Core Operating
manner that is detectable only by a Type A extension of the interval between Type A Limits Reports (COLR) for STP Units 1
test. Based on the above, the proposed test containment leak rate tests and Type C tests and 2. The amendment would also
interval extensions do not significantly for VEGP. The proposed surveillance interval revise the COLR Administrative
increase the consequences of an accident extension is bounded by the 15-year ILRT
previously evaluated. interval and the 75-month Type C test
Controls TS to add exclusion zones to
The proposed amendment also deletes interval currently authorized within NEI 94– the list of limits found in the COLRs,
exceptions previously granted under TS 01, Revision 3–A. Industry experience and to revise the description of the
Amendment Nos. 130 (VEGP–1) and 108 supports the conclusions that Types B and C methodology used to determine the
(VEGP–2), to allow one-time extensions of testing detects a large percentage of values. In addition, the proposed
the ILRT test frequency for VEGP. These containment leakage paths and that the amendment requests administrative
exceptions were for activities that would percentage of containment leakage paths that changes to the TSs.
have already taken place by the time this are detected only by Type A testing is small. Basis for proposed no significant
amendment is approved; therefore, their The containment inspections performed in hazards consideration determination:
deletion is solely an administrative action accordance with ASME Section XI and TS As required by 10 CFR 50.91(a), the
that has no effect on any component and no serve to provide a high degree of assurance
impact on how the unit is operated. that the containment would not degrade in a
licensee has provided its analysis of the
Therefore, the proposed change does not manner that is detectable only by Type A issue of no significant hazards
result in a significant increase in the testing. The combination of these factors consideration, which is presented
probability or consequences of an accident ensures that the margin of safety in the plant below:
previously evaluated. safety analysis is maintained. The design, 1. Does the proposed amendment involve
2. Does the proposed change create the operation, testing methods and acceptance a significant increase in the probability or
possibility of a new or different kind of criteria for Types A, B, and C containment consequences of an accident previously
accident from any accident previously leakage tests specified in applicable codes evaluated?
evaluated? and standards would continue to be met, Response: No.
Response: No. with the acceptance of this proposed change, The relocation of the Fxy exclusion zones
The proposed amendment to the TS 5.5.17, since these are not affected by changes to the to the COLRs has no impact on the accidents
Containment Leakage Rate Testing Program, Type A and Type C test intervals. analyzed in the STPNOC UFSAR [Updated
involves the extension of the VEGP Type A The proposed amendment also deletes Final Safety Analysis Report] and is not an
containment test interval to 15 years and the exceptions previously granted under TS accident initiator. Since the change does not
extension of the Type C test interval to 75 Amendment Nos. 130 (VEGP–1) and 108 impact any conditions that would initiate an
months. The containment and the testing (VEGP–2), to allow one-time extensions of accident, the probability or consequences of
requirements to periodically demonstrate the the ILRT test frequency for VEGP. This previously analyzed events is not increased.
integrity of the containment exist to ensure exception was for an activity that would have The proposed amendment does not change
the plant’s ability to mitigate the already taken place by the time this the actions to be taken if a core operating
consequences of an accident do not involve amendment is approved; therefore, the limit is exceeded and there are no physical
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any accident precursors or initiators. The deletion is solely an administrative action changes associated with this proposed
proposed change does not involve a physical and does not change how the unit is operated amendment.
change to the plant (i.e., no new or different and maintained. Thus, there is no reduction For each core reload, each accident
type of equipment will be installed) or a in any margin of safety as a result of this analysis addressed in the STP UFSAR will
change to the manner in which the plant is administrative change. continue to be examined with respect to
operated or controlled. Therefore, the proposed change does not changes in the cycle-dependent parameters,
The proposed amendment also deletes involve a significant reduction in a margin of which are obtained from the use of NRC-
exceptions previously granted under TS safety. approved reload design methodologies, to
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![57476 Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices
ensure that the transient evaluation of new unit within the cycle specific limits will not For further details with respect to the
reloads are bounded by previously accepted involve a significant reduction in the margin action see (1) the applications for
analyses. This examination, which will be of safety. amendment, (2) the amendment, and (3)
conducted per the requirements of 10 CFR The proposed amendment does not affect the Commission’s related letter, Safety
50.59, will ensure that future core reloads the design of the facility or system operating
will not involve a significant increase in the parameters, does not physically alter safety- Evaluation and/or Environmental
probability or consequences of an accident related systems and does not affect the Assessment as indicated. All of these
previously evaluated. method in which safety-related systems items can be accessed as described in
Therefore, there is no impact to the perform their functions. the ‘‘Obtaining Information and
probability or consequences of an accident Therefore, the proposed change does not Submitting Comments’’ section of this
previously evaluated due to the proposed impact margin of safety. document.
change. [The licensee stated that the administrative
[The licensee stated that the administrative changes proposed to the TSs do not impact Exelon Generation Company, LLC,
changes proposed to the TSs do not impact the operation of the facility in a manner that Docket Nos. 50–373 and 50–374, LaSalle
the operation of the facility in a manner that involves significant hazards considerations.] County Station (LSCS), Units 1 and 2,
involves significant hazards considerations.] LaSalle County, Illinois
2. Does the proposed amendment create
The NRC staff has reviewed the
the possibility of a new or different kind of licensee’s analysis and, based on this Date of amendment request: October
accident from any accident previously review, it appears that the standards of 26, 2016, as supplemented by letters
evaluated? 10 CFR 50.92(c) are satisfied. Therefore, dated February 16, July 17, August 8,
Response: No. the NRC staff proposes to determine that September 27, October 3, and November
The relocation of the Fxy exclusion zone the request for amendments involves no 8, 2017.
details from the Technical Specifications to significant hazards consideration. Brief description of amendments: The
the COLRs will not create the possibility of Attorney for licensee: Kym Harshaw, amendments revised Technical
a new or different kind of accident from any General Counsel, STP Nuclear Specification (TS) 5.5.13, ‘‘Primary
accident previously evaluated. No safety-
Operating Company, P.O. Box 289, Containment Leakage Rate Testing
related equipment, safety function, or plant
operation will be altered as a result of this Wadsworth, TX, 77483. Program,’’ to allow for the permanent
proposed change. No new operator actions NRC Branch Chief: Robert J. extension of the Type A Integrated Leak
are created as a result of the proposed Pascarelli. Rate Testing and Type C Leak Rate
change. The cycle-specific variables are III. Notice of Issuance of Amendments Testing frequencies, to change the
determined using the NRC approved methods
to Facility Operating Licenses and documents used by LSCS to implement
and the COLRs are submitted to the NRC to the performance-based leakage testing
allow the staff to continue to trend the values Combined Licenses
program, and to delete the information
of these limits. The Technical Specifications During the period since publication of regarding the performance of the next
will continue to require operation within the the last biweekly notice, the
core operating limits and appropriate actions
LSCS Type A tests to be performed.
Commission has issued the following Additionally, the amendments
will be required if these limits are exceeded. amendments. The Commission has
The relocation of the Fxy exclusion zones deleted Conditions 2.D.(e) and 2.D.(c),
to the COLRs has no impact on the accidents
determined for each of these respectively, of the LSCS Unit 1 and
analyzed in the STPNOC Updated Final amendments that the application Unit 2 Renewed Facility Operating
Safety Analysis Report (UFSAR) and is not complies with the standards and Licenses regarding conducting the third
an accident initiator. Since this change does requirements of the Atomic Energy Act Type A test of each 10-year service
not impact any conditions that would initiate of 1954, as amended (the Act), and the period when the plant is shut down for
an accident, there is no possibility of a new Commission’s rules and regulations. the 10-year inservice inspection.
or different kind of accident resulting from The Commission has made appropriate Date of issuance: November 16, 2017.
this change. findings as required by the Act and the Effective date: As of the date of its
Therefore, the proposed change does not Commission’s rules and regulations in issuance and shall be implemented
create the possibility of a new or different
kind of accident from any accident
10 CFR chapter I, which are set forth in within 60 days from the date of
previously evaluated. the license amendment. issuance.
[The licensee stated that the administrative A notice of consideration of issuance Amendment Nos.: 226 (Unit 1) and
changes proposed to the TSs do not impact of amendment to facility operating 212 (Unit 2). A publicly-available
the operation of the facility in a manner that license or combined license, as version is in ADAMS under Accession
involves significant hazards considerations.] applicable, proposed no significant No. ML17283A085; documents related
3. Does the proposed amendment involve hazards consideration determination, to these amendments are listed in the
a significant reduction in a margin of safety? and opportunity for a hearing in Safety Evaluation enclosed with the
Response: No. connection with these actions, was amendments.
The relocation of the Fxy exclusion zone published in the Federal Register as Renewed Facility Operating License
details from the Technical Specifications to
the COLRs will not affect the margin of
indicated. Nos. NPF–11 and NPF–18: The
safety. The margin of safety presently
Unless otherwise indicated, the amendments revised the TSs and
provided by the Technical Specifications Commission has determined that these Renewed Facility Operating Licenses.
remains unchanged. They will be amendments satisfy the criteria for Date of initial notice in Federal
incorporated into the COLR which is categorical exclusion in accordance Register: February 14, 2017 (82 FR
submitted to the NRC, therefore appropriate with 10 CFR 51.22. Therefore, pursuant 10597). The supplemental letters dated
measures exist to control the values of these to 10 CFR 51.22(b), no environmental February 16, July 17, August 8,
limits. The development of the limits for impact statement or environmental September 27, October 3, and November
sradovich on DSK3GMQ082PROD with NOTICES
future reloads will continue to conform to assessment need be prepared for these 8, 2017, provided additional
those methods described in NRC-approved amendments. If the Commission has information that clarified the
documentation. STPNOC will continue to
confirm all safety analysis limits remain
prepared an environmental assessment application, did not expand the scope of
bounding on a cycle-specific basis using an under the special circumstances the application as originally noticed,
NRC-approved methodology. Each core provision in 10 CFR 51.22(b) and has and did not change the NRC staff’s
reload will involve a Reload Safety made a determination based on that original proposed no significant hazards
Evaluation to assure that operation of the assessment, it is so indicated. consideration determination.
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![57478 Federal Register / Vol. 82, No. 232 / Tuesday, December 5, 2017 / Notices
Brief description of amendment: The Amendment Nos.: 301 (Unit 1), 325 Dated at Rockville, Maryland, this 27th day
amendment modified Hope Creek (Unit 2), and 285 (Unit 3). A publicly- of November 2017.
Generating Station Technical available version is in ADAMS under For the Nuclear Regulatory Commission.
Specification 6.8.4.f, ‘‘Primary Accession No. ML17277A207; Kathryn M. Brock,
Containment Leakage Rate Testing documents related to these amendments Acting Director, Division of Operating Reactor
Program,’’ to extend the Type A reactor are listed in the Safety Evaluation Licensing, Office of Nuclear Reactor
containment pressure test interval from enclosed with the amendments. Regulation.
one test in 10 years to one test in 15 Renewed Facility Operating License [FR Doc. 2017–25901 Filed 12–4–17; 8:45 am]
years, and extend the Type C test Nos. DPR–33, DPR–52, and DPR–68: BILLING CODE 7590–01–P
interval up to 75 months, with a Amendments revised the Renewed
permissible extension period of 9 Facility Operating Licenses and TSs.
months (total of 84 months) for non- Date of initial notice in Federal NUCLEAR REGULATORY
routine emergency conditions. Register: April 25, 2017 (82 FR 19106). COMMISSION
Date of issuance: November 8, 2017. The supplemental letter dated June 29, [NRC–2017–0219]
Effective date: As of the date of 2017, provided additional information
issuance and shall be implemented that clarified the application, did not Applications and Amendments to
within 60 days of issuance. expand the scope of the application as Facility Operating Licenses and
Amendment No.: 207. A publicly- originally noticed, and did not change Combined Licenses Involving
available version is in ADAMS under the staff’s original proposed no Proposed No Significant Hazards
Accession No. ML17291A209; significant hazards consideration Considerations and Containing
documents related to this amendment determination as published in the Sensitive Unclassified Non-Safeguards
are listed in the Safety Evaluation Federal Register. Information and Order Imposing
enclosed with the amendment. The Commission’s related evaluation Procedures for Access to Sensitive
Renewed Facility Operating License of the amendment is contained in a Unclassified Non-Safeguards
No. NPF–57: Amendment revised the Safety Evaluation dated November 8, Information
Renewed Facility Operating License and 2017.
Technical Specifications. No significant hazards consideration AGENCY: Nuclear Regulatory
Date of initial notice in Federal Commission.
comments received: No.
Register: December 20, 2016 (81 FR ACTION: License amendment request;
92869). The supplemental letters dated Tennessee Valley Authority, Docket No. notice of opportunity to comment,
March 27, 2017, and July 13, 2017, 50–391, Watts Bar Nuclear Plant, Unit 2, request a hearing, and petition for leave
provided additional information that Rhea County, Tennessee to intervene; order imposing
clarified the application, did not expand Date of amendment request: March procedures.
the scope of the application as originally 28, 2017.
noticed, and did not change the NRC Brief description of amendment: The SUMMARY: The U.S. Nuclear Regulatory
staff’s original proposed no significant amendment revised the completion date Commission (NRC) received and is
hazards consideration determination as for License Condition 2.C.(5) for Watts considering approval of two amendment
published in the Federal Register. Bar Nuclear Plant, Unit 2, regarding the requests. The amendment requests are
The Commission’s related evaluation for Shearon Harris Nuclear Power Plant,
completion of action to resolve the
of the amendment is contained in a Unit 1; and LaSalle County Station,
issues identified in Bulletin 2012–01,
Safety Evaluation dated November 8, Units 1 and 2. The NRC proposes to
‘‘Design Vulnerability in Electric Power
2017. determine that each amendment request
System’’ (ADAMS Accession No.
No significant hazards consideration involves no significant hazards
ML12074A115), from December 31,
comments received: No. consideration. Because each amendment
2017, to December 31, 2018, to align
Tennessee Valley Authority, Docket with the remainder of the Tennessee request contains sensitive unclassified
Nos. 50–259, 50–260, and 50–296, Valley Authority fleet and with the non-safeguards information (SUNSI), an
Browns Ferry Nuclear Plant, Units 1, 2 nuclear industry. order imposes procedures to obtain
and 3, Limestone County, Alabama Date of issuance: November 6, 2017. access to SUNSI for contention
Effective date: As of the date of preparation.
Date of amendment request: January
17, 2017, as supplemented by letter issuance and shall be implemented DATES: Comments must be filed by
dated June 29, 2017. within 15 days of issuance. January 4, 2018. A request for a hearing
Brief description of amendments: The Amendment No.: 17. A publicly- must be filed by February 5, 2018. Any
amendments change technical available version is in ADAMS under potential party as defined in § 2.4 of title
specifications (TSs) consistent with Accession No. ML17258A328; 10 of the Code of Federal Regulations
Technical Specifications Task Force documents related to this amendment is (10 CFR), who believes access to SUNSI
(TSTF) Standard Technical listed in the Safety Evaluation enclosed is necessary to respond to this notice
Specifications Change Traveler TSTF– with the amendment. must request document access by
545, Revision 3, ‘‘TS Inservice Testing Facility Operating License No. NPF– December 15, 2017.
Program Removal & Clarify SR 96: Amendment revised the Facility ADDRESSES: You may submit comments
[Surveillance Requirement] Usage Rule Operating License. by any of the following methods:
Date of initial notice in Federal • Federal Rulemaking Web site: Go to
sradovich on DSK3GMQ082PROD with NOTICES
Application to Section 5.5 Testing,’’ and
TSTF–299, Revision 0, ‘‘Administrative Register: July 5, 2017 (82 FR 31103). http://www.regulations.gov and search
Controls Program 5.5.2.b Test Interval The Commission’s related evaluation for Docket ID NRC–2017–0219. Address
and Exception.’’ of the amendment is contained in a questions about NRC dockets to Carol
Date of issuance: November 8, 2017. Safety Evaluation dated November 6, Gallagher; telephone: 301–415–3463;
Effective date: As of the date of 2017. email: Carol.Gallagher@nrc.gov. For
issuance and shall be implemented No significant hazards consideration technical questions, contact the
within 60 days of issuance. comments received: No. individual listed in the FOR FURTHER
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Document Created: 2017-12-04 23:45:02
Document Modified: 2017-12-04 23:45:02
| 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 January 4, 2018. A request for a hearing must be filed by February 5, 2018. | |
| Contact | Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-2422, email: [email protected] | |
| FR Citation | 82 FR 57469 |
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