| Page Range | 73428-73447 | |
| FR Document | 2016-25641 |
[Federal Register Volume 81, Number 206 (Tuesday, October 25, 2016)] [Notices] [Pages 73428-73447] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2016-25641] ======================================================================= ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2016-0214] 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 September 27, 2016, to October 7, 2016. The last biweekly notice was published on October 11, 2016. DATES: Comments must be filed by November 25, 2016. A request for a hearing must be filed by December 27, 2016. ADDRESSES: You may submit comments by any of the following methods.Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0214. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected]. I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0214, 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-2016-0214. NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [email protected]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document. NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2016-0214, 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:/ / [[Page 73429]] www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. I. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition 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, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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/. If a petition is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (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. The petition must also set forth the specific contentions which the petitioner seeks to have litigated at 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 shall 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 those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall 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 these requirements 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, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions consistent with the NRC's regulations, policies, and procedures. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day 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). 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 decide 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 held 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 any amendment unless the Commission finds an imminent danger to the health [[Page 73430]] or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by December 27, 2016. 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. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c). If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in 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 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, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene (hereinafter ``petition''), and documents filed by interested governmental entities participating 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. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [email protected], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition (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. System requirements for accessing the E-Submittal server are available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Electronic Filing Help Desk will not be able to offer assistance in using unlisted software. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a petition. Submissions should 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 documents are 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 documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically using the NRC's adjudicatory E- Filing system may seek assistance by contacting the NRC 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 7 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, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document 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 http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal [[Page 73431]] privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a petition will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. 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. The Commission will issue a notice or order granting or denying a hearing request or intervention petition, designating the issues for any hearing that will be held and designating the Presiding Officer. A notice granting a hearing will be published in the Federal Register and served on the parties to the hearing. 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. Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power Station (KPS), Carlton, Wisconsin Date of amendment request: September 14, 2015. A publicly available version is in ADAMS under Accession No. ML15261A238. Description of amendment request: The amendment would revise the KPS Permanently Defueled Emergency Plan (PDEP) and the Permanently Defueled Emergency Action Levels (EAL) Bases Document. DEK requests revisions of the PDEP and the EAL Bases Document that reflect DEK's plan to transfer all spent fuel to the independent spent fuel storage installation (ISFSI). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment would modify the KPS renewed facility operating license by revising the emergency plan and revising the EAL scheme. KPS has permanently ceased operation and is permanently defueled. The proposed amendment is conditioned on all spent nuclear fuel being removed from wet storage in the spent fuel pool and placed in dry storage within the ISFSI. Occurrence of postulated accidents associated with spent fuel stored in a spent fuel pool is no longer credible in a spent fuel pool devoid of such fuel. The proposed amendment has no effect on plant systems, structures, and components (SSCs) and no effect on the capability of any plant SSC to perform its design function. The proposed amendment would not increase the likelihood of the malfunction of any plant SSC. The proposed amendment would have no effect on any of the previously evaluated accidents in the KPS Updated Safety Analysis Report (USAR). Since KPS has permanently ceased operation, the generation of fission products has ceased and the remaining source term continues to decay. This continues to significantly reduce the consequences of previously postulated accidents. Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment constitutes a revision of the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at KPS. The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment as a result of the proposed amendment. Similarly, the proposed amendment would not physically change any SSCs involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures. The credible events for the ISFSI remain unchanged. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in the margin of safety? Response: No. Because the 10 CFR part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. With all nuclear spent fuel pool transferred out of wet storage from the spent fuel pool and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible. There are no longer credible events that would result in any releases beyond the site boundary exceeding the EPA PAG [Environmental Protection Agency protective action guideline] exposure levels, as detailed in the EPA's ``Protective Action Guide and Planning Guidance for Radiological Incidents,'' Draft for Interim Use and Public Comment dated March 2013 (PAG Manual). The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not affect either the way in which the plant structures, systems, and components perform their safety function or their design margins. Because there is no change to the physical design of the plant, there is no change to any of these margins. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Bruce A. Watson. Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power Station (KPS), Carlton, Wisconsin Date of amendment request: July 28, 2016. A publicly available version is in ADAMS under Accession No. ML16216A187. Description of amendment request: The amendment would revise the KPS Updated Safety Analysis Report (USAR) Section 9.5.2.2.4, ``Auxiliary Building Crane,'' to: (1) Add a description of a non-single failure proof intermediate lifting device that DEK intends to use during a specific spent fuel cask handling activity in the auxiliary building, and (2) incorporate a new load drop analysis applicable to the use of this intermediate lifting device. The amendment also includes (for information) a new Technical Requirements Manual section that governs the use of the non-single failure proof intermediate lifting device to ensure compliance with the required parameters in the load drop analysis. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. [[Page 73432]] The probability of a heavy load drop onto fuel is unchanged by this amendment since the intermediate lift device is not used for handling loaded or unloaded spent fuel canisters in or around the spent fuel pool. Heavy load lifts in and around the spent fuel pool will continue to be performed per the current licensing basis. The proposed amendment has no effect on the capability of any plant systems, structures, and components (SSCs) to perform their design functions. The spent fuel pool is unaffected by the proposed amendment. The design function of the auxiliary building crane is not changed. Other lifting devices and interfacing lifting points associated with spent fuel cask handling are designed in accordance with applicable NRC guidance pertaining to single failure proof lifting systems. Therefore, the proposed amendment would not increase the likelihood of the malfunction of any plant SSC. The proposed amendment would have no effect on any of the previously evaluated accidents in the KPS USAR. Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment does not affect cask handling activities in or around the KPS spent fuel pool. Drops of heavy loads will continue to be very improbable events. Use of a different type of equipment to lift spent fuel canisters does not involve any new or different kind of accident. The proposed amendment does not involve a physical alteration of the plant. Similarly, the proposed amendment would not physically change any SSCs involved in the mitigation of any postulated accidents. The physical structure of the spent fuel canisters is not altered by this amendment. The possibility of a heavy load drop onto fuel remains non- credible since the intermediate lift device is not used to handle spent fuel canisters in or around the spent fuel pool. Heavy load lifts in and around the spent fuel pool will continue to be performed per the current licensing basis. The proposed amendment does not impact safe shutdown equipment. The spent fuel pool, including its cooling and inventory makeup, is unaffected by the proposed amendment. The current licensing basis (USAR Section 14.2.1) includes evaluations of the consequences of a fuel handling accident involving failure of fuel cladding. Postulation of a canister load drop creates the possibility of a new initiator of this previously evaluated accident (failure of fuel cladding) caused by the postulated non-mechanistic single failure of the intermediate lift device. The analysis concludes that the postulated drop of a canister loaded with fuel assemblies would not result in failure of canister integrity (and therefore there would be no radiological release). The consequences of a canister drop are bounded by the current licensing scenario of a fuel handling accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in the margin of safety? Response: No. Heavy load handling will continue to be conducted in accordance with NRC approved methods. Analysis of a postulated load drop of a loaded spent fuel canister demonstrates satisfactory outcomes. The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not significantly affect either the way in which the plant structures, systems, and components perform their safety function or their design margins. Because there is no change to the physical design of the plant, there is likewise no significant change to any of these margins. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Bruce A. Watson. Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida Date of amendment request: August 31, 2016. A publicly available version is in ADAMS under Accession No. ML16243A259. Description of amendment request: The amendment would revise the Operating License and associated Permanently Defueled Technical Specifications (PDTS) to reflect removal of all CR-3 spent nuclear fuel from the spent fuel pools and its transfer to dry cask storage within the onsite Independent Spent Fuel Storage Installation (ISFSI). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment would modify the CR-3 facility operating license and PDTS by deleting the portions of the license and PDTS that are no longer applicable to a facility with no spent nuclear fuel stored in the spent fuel pools, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment will be implemented within 60 days following DEF's submittal of written notification to the NRC that all spent fuel assemblies have been transferred out of the spent fuel pools and placed in dry storage within the ISFSI. The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure: 1. The integrity of the reactor coolant boundary; 2. The capability to shutdown the reactor and maintain it in a safe shutdown condition; or 3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.43(a)(1) or 100.11. The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI, none of the SSCs at CR-3 are required to be relied on for accident mitigation. Therefore, none of the SSCs at CR-3 meet the definition of a safety- related SSC stated in 10 CFR 50.2. The proposed deletion of requirements in the PDTS does not affect systems credited in any accident analysis at CR-3. Section 14 of the CR-3 Final Safety Analysis Report (FSAR) described the design basis accidents (DBAs) related to the spent fuel pools. These postulated accidents are predicated on spent fuel being stored in the spent fuel pools. With the removal of the spent fuel from the spent fuel pools, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Manager, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences. The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? [[Page 73433]] Response: No. The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the spent fuel pools, and relocate certain administrative controls to the Quality Assurance Program Description or other licensee controlled document. After the removal of the spent fuel from the spent fuel pools and transfer to the ISFSI, there are no spent fuel assemblies that remain in the spent fuel pools. Coupled with a prohibition against storage of fuel in the spent fuel pools, the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The removal of all spent nuclear fuel from the spent fuel pools into storage in casks within an ISFSI, coupled with a prohibition against future storage of fuel within the spent fuel pools, removes the potential for fuel related accidents. The design basis and accident assumptions within the CR-3 FSAR and the PDTS relating to safe management and safety of spent fuel in the spent fuel pools are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities. The requirements for systems, structures, and components (SSCs) that have been removed from the CR-3 PDTS are not credited in the existing accident analysis for any applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara S. Nichols, 550 South Tryon Street, Charlotte, NC 28202. NRC Branch Chief: Bruce A. Watson. Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: June 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16182A387. Description of amendment request: The amendment would revise HNP Technical Specifications (TSs) to (1) delete the Gaseous Radwaste Treatment System definition from TSs, (2) relocate Explosive Gas Mixture TS requirements and Liquid Holdup Tanks TS requirements to a licensee-controlled program in the Procedures and Programs TSs section, and (3) modify the Gas Storage Tank Radioactivity Monitoring Program TSs into an Explosive Gas and Storage Tank Radioactivity Monitoring Program to include controls for potentially explosive gas mixtures and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. 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 do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative in nature and alter only the format and location of programmatic controls and procedural details relative to explosive gas monitoring and liquid holdup tanks. Existing TS containing procedural details are being relocated to licensee control. Compliance with applicable regulatory requirements will continue to be maintained. In addition, the proposed changes do not alter the conditions or assumptions in any of the previous accident analyses. Because the previous accident analyses remain bounding, the radiological consequences previously evaluated are not adversely affected by the proposed changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there will be no change in types or increase in the amounts of any effluents released offsite. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes do not involve a significant reduction in a margin of safety and are considered administrative in nature. The proposed changes do not involve any actual change in the methodology used in the monitoring of explosive gas mixtures contained in the Gaseous Waste Processing System. HNP does not currently utilize unprotected outdoor liquid storage tanks; therefore, there are no associated methodology changes with this request. These changes provide for the relocation of procedural details outside of the technical specifications with the addition of appropriate administrative controls to provide continued assurance of compliance to applicable regulatory requirements. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202. NRC Acting Branch Chief: Jeanne A. Dion. Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington Date of amendment request: July 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16194A515. Description of amendment request: The amendment would reduce the minimum reactor dome pressure associated with the critical power correlation from 785 pounds per square inch gauge (psig) to 685 psig in Technical Specification (TS) 2.1.1, ``Reactor Core SLs [Safety Limits],'' and associated bases. 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 change does not involve a modification of any plant hardware; the probability and consequence of the Pressure Regulator Failure Open (PRFO) transient are essentially unchanged. The reduction in the reactor dome pressure safety limit (SL) from 785 psig to 685 psig provides greater margin to accommodate the pressure reduction [[Page 73434]] during the transient within the revised TS limit. The proposed change will continue to support the validity range for the correlations and the calculation of Minimum Core Power Ratio (MCPR) as approved. The proposed TS revision involves no significant changes to the operation of any systems or components in normal, accident or transient operating conditions. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed reduction in the reactor dome pressure SL from 785 psig to 685 psig is a change based upon previously approved documents and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced. Therefore, the change does not introduce a new or different kind of accident from those previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. The proposed change in reactor dome pressure enhances the safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged. The available pressure range is expanded by the change, thus offering greater margin for pressure reduction during the transient. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817. NRC Branch Chief: Robert J. Pascarelli. Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant (JAF), Oswego County, New York Date of amendment request: January 15, 2016, as supplemented by letters dated June 3, 2016, and September 19, 2016. Publicly available versions are available in ADAMS under Accession Nos. ML16015A456, ML16155A326, and ML16263A237, respectively. Description of amendment request: The licensee has provided a formal notification to the NRC of the intention to permanently cease power operations of JAF at the end of the current operating cycle. Once certifications for permanent cessation of operation and permanent removal of fuel from the reactor are submitted to the NRC, certain staffing and training Technical Specifications (TSs) administrative controls will no longer be applicable or appropriate for the permanently defueled condition. Therefore, ENO is requesting approval of changes to the staffing and training requirements in Section 5.0, ``Administrative Controls,'' of the JAF TSs. Specifically, the amendment would revise and remove certain requirements in TS Sections 5.1, ``Responsibility''; 5.2, ``Organization''; and 5.3, ``Plant Staff Qualifications,'' and add additional definitions to TS Section 1.1, ``Definitions.'' The proposed amendment would not be effective until the certification of permanent cessation of operation and certification of permanent removal of fuel from the reactor vessel are submitted to the NRC. The license amendment request was originally noticed in the Federal Register on March 1, 2016 (81 FR 10678). The notice is being reissued in its entirety to include the revised scope and description of the amendment request. 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, with NRC staff revisions provided in [brackets], which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment would not take effect until JAF has permanently ceased operation and entered a permanently defueled condition. The proposed amendment would modify the JAF TS by deleting the portions of the TS that are no longer applicable to a permanently defueled facility, while modifying the other sections to correspond to the permanently defueled condition. The deletion and modification of provisions of the administrative controls do not directly affect the design of structures, systems, and components (SSCs) necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor. In a permanently defueled condition, the only credible accident is the fuel handling accident. The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The administrative removal of or modifications of the TS that are related only to administration of facility cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and JAF will no longer be authorized to operate the reactor. The proposed deletion of requirements of the JAF TS do not affect systems credited in the accident analysis for the fuel handling accident at JAF. The proposed TS will continue to require proper control and monitoring of safety significant parameters and activities. The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Because the 10 CFR part 50 license for JAF will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel once the certifications required by 10 CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible accident is a fuel handling accident (FHA). The proposed [[Page 73435]] amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA. The proposed changes are limited to those portions of the OL [operating license] and TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the JAF OL and TS are not credited in the existing accident analysis for the remaining applicable as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs [design-basis accidents] involving the reactor are no longer possible because the reactor will be permanently shutdown and defueled and JAF will no longer be authorized to operate the reactor. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: July 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16210A227. Description of amendment request: The amendments would revise technical specification (TS) requirements relating to: (1) The inservice inspection (ISI) program required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Code (Code), and (2) the inservice testing (IST) program required by the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The proposed changes are based, in part, on Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.'' 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 revises TS 4.0.5, Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 & 3 components, by revising the Inservice Testing Program and Inservice Inspection Program specification. Most requirements in the IST Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining requirements in the TS Section 4.0.5, IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, ``Inservice Testing Program,'' is added to the TS, which references the requirements of 10 CFR 50.55a(f). Similarly, the requirements in the ISI Program are revised, as they are [ ] duplicative of requirements in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. Performance of inservice testing or inservice inspection is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than two years may be extended by up to six months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing or inservice inspection performed. In most cases, the frequency of inservice testing and inservice inspection is unchanged. However, the frequency of testing or inspection would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change eliminates some provisions from the TS in lieu of provisions in the ASME Code, as modified by use of Code Case OMN-20 (IST) or ASME Boiler and Pressure Vessel Code (ISI). Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than two years to be extended by six months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS SR 4.0.2 allowance to perform a specified surveillance time interval with a maximum allowable extension not to exceed 25% of the surveillance interval, unless there is a specific SR referencing usage of the INSERVICE TESTING PROGRAM and TS SR 4.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. However, elimination of the statement will have no effect on plant operation or safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Douglas A. Broaddus. Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania Date of amendment request: July 15, 2016. A publicly-available version is in ADAMS under Package Accession No. ML16201A306. Description of amendment request: The amendment would revise the [[Page 73436]] Radiological Emergency Plan Annex for TMI-1. The proposed changes would decrease the radiation protection technician staffing from three to two technicians, remove two maintenance technicians currently assigned to the repair and corrective action function, and eliminate the on-shift Operations Support Center director position. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to the TMI Emergency Plan do not increase the probability or consequences of an accident. The proposed changes do not impact the function of plant Structures, Systems, or Components (SSCs). The proposed changes do not affect accident initiators or accident precursors, nor do the changes alter design assumptions. The proposed changes do not alter or prevent the ability of the onsite ERO [emergency response organization] to perform their intended functions to mitigate the consequences of an accident or event. The proposed changes remove onsite ERO positions no longer credited or considered necessary in support of Emergency Plan implementation. Therefore, the proposed changes to the Emergency Plan do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed changes have no impact on the design, function, or operation of any plant SSCs. The proposed changes do not affect plant equipment or accident analyses. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed changes do not introduce failure modes that could result in a new accident, and the proposed changes do not alter assumptions made in the safety analysis. The proposed changes remove onsite ERO positions no longer credited or considered necessary in support of Emergency Plan implementation. Therefore, the proposed changes to the Emergency Plan do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. Margins of safety are unaffected by the proposed changes to the ERO minimum on-shift staffing. The proposed changes are associated with the Emergency Plan staffing and do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by these proposed changes. The proposed changes to the Emergency Plan will continue to provide the necessary onsite ERO response staff. Therefore, the proposed changes to the Emergency Plan do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Douglas A. Broaddus. Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50- 306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of amendment request: August 31, 2016. A publicly-available version is in ADAMS under Accession No. ML16244A493. Brief description of amendment request: The amendments would revise the Required Actions and associated Completion Times to Technical Specification (TS) 3.8.7, ``Inverters--Operating.'' Specifically, Condition B would be deleted and current Condition C would be re- lettered to Condition B. Additionally, the Required Actions and associated Completion Times for Condition A would be modified to require restoration of one inoperable inverter to operability within 24 hours. These changes conform to Improved Standard Technical Specification TS 3.8.7. 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 the TS for the purpose of eliminating a non-conservative Required Action. The proposed TS change does not introduce new equipment or new equipment operating modes, nor does the proposed change alter existing system relationships. The proposed change does not affect normal plant operation. Further, the proposed change does not increase the likelihood of the malfunction of any SSC [structure, system and component] or impact any analyzed accident. Consequently, the probability of an accident previously evaluated is not affected and there is no significant increase in the consequences of any 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 accident from any accident previously evaluated? Response: No. The proposed change revises the TS for the purpose of eliminating a non-conservative Required Action. 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 operations. The proposed change does not alter assumptions made in the safety analysis. Further, the proposed change does not introduce new accident initiators. 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 a margin of safety? Response: No. The proposed change revises the TS for the purpose of eliminating a non-conservative Required Action. 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 assumptions and acceptance criteria are not affected by this 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 requests involve no significant hazards consideration. Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy [[Page 73437]] Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401. NRC Branch Chief: David J. Wrona. PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: July 20, 2016. A publicly-available version is in ADAMS under Accession No. ML16203A006. Description of amendment request: The amendment would revise the Hope Creek Generating Station (Hope Creek) Technical Specifications (TS), Section 6.8.4.i, ``Inservice Testing Program,'' to remove requirements duplicated in the American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants Case OMN-20, ``Inservice Test Frequency.'' A new defined term, ``Inservice Testing Program,'' will be added to the TS 1.0, ``Definitions,'' section. The licensee stated that the proposed change to the TS is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarity SR Usage Rule Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555), with no proposed variations or deviations. However, the Hope Creek TS uses different numbering for surveillance requirements than the Standard Technical Specifications on which TSTF-545 was based, so the licensee changed the TSTF-545 numbering to be consistent with the Hope Creek TS numbering. 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 change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS Chapter 6, ``Administrative Controls,'' Section 6.8, ``Procedures and Programs,'' by eliminating the ``Inservice Testing Program'' specification. Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining requirements in the Section 6.8 IST Program are eliminated [. . .]. A new defined term, ``Inservice Testing Program,'' is added to the TS, which references the requirements of 10 CFR 50.55a(f). Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered. 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 change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS 4.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. [. . .] However, elimination of the statement will have no effect on plant operation or 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: Douglas A. Broaddus. South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South Carolina Date of amendment request: September 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16259A310. Description of amendment request: The amendments would revise Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer Nuclear Station, Units 2 and 3. The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* information. Specifically, the proposed changes would revise the Combined Licenses to clarify information in WCAP-17179, ``AP1000[supreg] Component Interface Module Technical Report,'' which demonstrates design compliance with licensing bases requirements. WCAP- 17179 is incorporated by reference into the UFSAR to provide additional details regarding the component interface module (CIM) system design. The requested amendments also propose a change to the CIM internal power supply that will enable proper functioning of the field programmable gate arrays. 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? [[Page 73438]] Response: No. The proposed change to the CIM internal power supply enables the field programmable gate array (FPGA) to function properly. The proposed change to the FPGA core power has no adverse effect on the operation of the output actuation relays. The function of the internal power supply has no input to plant safety analysis. The change to the CIM internal power supply has a negligible effect on the 24 Vdc [volts direct current] supplies and ultimately the plant electrical system load and has no adverse effect on the CIM functionality. The proposed changes to clarify how licensing basis design documentation reflects compliance with license basis requirements, and the proposed change to the ownership of safety remote node controller (SRNC) and CIM intellectual property, are not technical changes. The proposed changes do not affect any accident initiator in the UFSAR, or affect the radioactive material releases in the UFSAR accident analyses. The proposed change does not alter the ability of the facility to prevent and mitigate abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, or component (SSC) or function is adversely affected. The change does not involve or interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. This activity does not involve a new fission product release path, nor a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. Because the proposed changes do not change any safety-related SSC or function credited in the mitigation of an accident, the consequences of the accidents evaluated in the UFSAR are not affected. Therefore, the requested amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the CIM internal power supply enables the FPGA to function properly and does not involve accident initiators. The change to the CIM internal power supply has a negligible effect on the 24 Vdc supplies and ultimately the plant electrical system load and has no adverse effect on CIM functionality. The proposed clarified descriptions and the proposed change to the ownership of SRNC and CIM intellectual property are not technical changes. The proposed changes do not affect other plant equipment or adversely affect the design of the CIM. Therefore, the proposed changes do not affect any safety-related equipment itself, nor do they affect equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. No system or design function or equipment qualification would be adversely affected by the proposed changes. Furthermore, the proposed changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to the CIM internal power supply enables the FPGA to function properly. The function of the internal power supply has no input to plant safety analysis. The change to the CIM internal power supplies has a negligible effect on the 24 Vdc supplies and ultimately the plant electrical system load and has no adverse effect on the CIM functionality. The proposed clarified descriptions and the proposed change to the ownership of SRNC and CIM intellectual property are not technical changes. The proposed changes do not adversely affect the design, construction, or operation of any plant SSCs, including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes. No safety analysis or design basis acceptance limit/ criterion is challenged or exceeded by the proposed changes, thus no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514. NRC Branch Chief: Michael T. Markley. South Carolina Electric & Gas Company, Inc., Docket Nos. 52-027 and 52- 028, Virgil C. Summer Nuclear Station Units 2 and 3, Fairfield, South Carolina Date of amendment request: September 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16272A373. Description of amendment request: The amendment request proposes changes to revise plant-specific Tier 1, plant-specific Tier 2, and Combined License (COL) Appendix C information concerning the details of the Class 1E direct current and uninterruptible power supply system (IDS), specifically adding seven Class 1E fuse panels to the IDS design. These proposed changes provide electrical isolation between the non-Class 1E IDS battery monitors and their respective Class 1E battery banks. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below 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 changes to revise plant-specific Tier 1, COL Appendix C, and [Updated Final Safety Analysis Report (UFSAR)] information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect the design functions of the IDS, including the Class 1E battery banks and the battery monitors. These proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits as described in the current licensing basis do not have an adverse effect on any of the design functions of any plant systems. The proposed changes do not adversely affect any plant electrical system and do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors. Therefore, the requested amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. [[Page 73439]] The proposed changes to revise plant-specific Tier 1, COL Appendix C, and UFSAR information concerning details of the IDS, specifically the addition of seven Class 1E fuse isolation panels at the interconnection of the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a fault on non-Class 1E circuits or equipment from degrading the operation of Class 1E IDS circuits and equipment below an acceptable level. The proposed changes do not adversely affect any plant electrical system and do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. There is no safety-related [structure, system, and component (SSC)] or function adversely affected by the proposed change to add IDS fuse isolation panels to non-Class 1E IDS battery monitors and Class 1E IDS circuits. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes and no margin or safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. 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: December 22, 2015, as supplemented by letter dated July 27, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15356A655 and ML16209A477, respectively. Description of amendment request: The proposed changes would revise the Combined License (COL) Appendix C and corresponding plant-specific Tier 1 information to add two turbine building sump pumps to accommodate the increased flow that will be experienced during condensate polishing system rinsing operations, for each unit, respectively. The proposed changes include information in the combined license, Appendix C. An exemption request relating to the proposed changes to the AP1000 DCD Tier 1 is included with the request. 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 changes to identify that there is more than one turbine building sump and to add two turbine building sump pumps (WWS-MP-07A and WWS-MP-07B) to COL Appendix C Subsection 2.3.29 and corresponding Table 2.3.29-1 will provide consistency within the current licensing basis. The main turbine building sumps and sump pumps are not safety-related components and do not interface with any systems, structures, or components (SSCs) accident initiator or initiating sequence of events; thus, the probability of accidents evaluated within the [Updated Final Safety Analysis Report (UFSAR)] are not affected. The proposed changes do not involve a change to the predicted radiological releases due to accident conditions, thus the consequences of accidents evaluated in the UFSAR are not affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to identify that there is more than one turbine building sump and to add two turbine building sump pumps to the non-safety waste water system (WWS) do not affect any safety- related equipment, nor do they add any new interface to safety- related SSCs. No system or design function or equipment qualification is affected by these changes. The changes do not introduce a new failure mode, malfunction, or sequence of events that could affect safety or safety-related equipment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The WWS is a non-safety-related system that does not interface with any safety-related equipment. The proposed changes to identify that there is more than one turbine building sump and to add two turbine building sump pumps do not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/ criterion is challenged or exceeded by the proposed change. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Jennifer Dixon-Herrity. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: August 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16236A266. Description of amendment request: The proposed changes would amend Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating Plant, Units 3 and 4. The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information and involve related changes to the Combined Operating License Appendix C (and corresponding plant-specific design control document Tier 1) information. Specifically, the proposed departures consist of changes to the design reliability assurance program (D-RAP) to identify the covers for the in-containment refueling water storage tank vents and overflow weirs as the risk-significant components included in the D-RAP and to differentiate between the rod drive motor- generator (MG) sets field control relays and the rod drive power supply control cabinets in which the relays are located. 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? [[Page 73440]] Response: No. The in-containment refueling water storage tank (IRWST) provides flooding of the refueling cavity for normal refueling. The tank also serves as a heat sink during Passive Residual Heat Removal (PRHR) Heat Exchanger (HX) operation and in the event of a loss-of-coolant- accident (LOCA) provides injection in support of long-term RCS [reactor coolant system] cooling. This activity adds normally closed covers to the IRWST vents and overflow weirs to prevent debris from entering the tank, prevent over-pressurization and accommodate volume and mass increases in the tank. The vent and overflow weir covers open upon differential pressures between the IRWST and containment. The rod drive MG sets provide the power to the control rod drive mechanisms through the reactor trip switchgear. This activity revises the equipment description and equipment tag associated with the risk-significant control relays which open to de-energize the rod drive MG sets and permit rods to drop. The proposed changes to add the IRWST vent and overflow weir covers and to change the description of the equipment and equipment tag related to the rod drive MG sets does not inhibit the SSCs from performing their safety-related function. The design bases of the IRWST vents and overflow weirs are not modified as a result of the addition of the covers to the vents and overflow weirs and the change to the control cabinet relay description and equipment tag. This proposed amendment does not have an adverse impact on the response to anticipated transients or postulated accident conditions because the functions of the SSCs are not changed. Required IRWST venting is not affected for any accident conditions. Required DAS functions are not affected for any accident conditions. Safety- related structure, system, component (SSC) or function is not adversely affected by this change. The changes to include the IRWST covers and to change the control cabinet relay description and tag number do not involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. The proposed changes do not involve a change to the predicted radiological releases due to postulated accident conditions, thus, the consequences of the accidents evaluated in the UFSAR are not affected. Probabilistic Risk Assessment (PRA) modeling and analyses associated with the SSCs are not impacted by this change. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the design of the IRWST vent and overflow weir covers do not adversely affect any safety-related equipment, and do not add any new interfaces to safety-related SSCs. No system or design function or equipment qualification is affected by these changes. The changes do not introduce a new failure mode, malfunction or sequence of events that could affect plant safety or safety-related equipment as the simplistic design of the cover louvers and hinged flappers are not considered unique designs. No new credible failure modes are introduced by the addition of the covers. The proposed changes to the description and equipment tag associated with the risk-significant control relays for the rod drive MG sets do not adversely affect any safety-related equipment, and do not add any new interfaces to safety-related SSCs. No system or design function or equipment qualification is affected by these changes. The changes do not introduce a new failure mode, malfunction or sequence of events that could affect plant safety or safety-related equipment because the design function of the control relays, control cabinets, or rod drive MG sets is not changed. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes maintain compliance with the applicable Codes and Standards, thereby maintaining the margin of safety associated with these SSCs. The proposed changes do not alter any applicable design codes, code compliance, design function, or safety analysis. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change, thus the margin of safety is not reduced. Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by these changes, no margin of safety is reduced. 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52- 026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: August 31, 2016. A publicly-available version is in ADAMS under Accession No. ML16244A836. Description of amendment request: The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2* information. Specifically, the proposed changes would revise the Combined Licenses for the Vogtle Electric Generating Plant, Units 3 and 4, to clarify information in WCAP-17179, ``AP1000[supreg] Component Interface Module Technical Report,'' which demonstrates design compliance with licensing bases requirements. WCAP-17179 is incorporated by reference into the UFSAR to provide additional details regarding the component interface module (CIM) system design. The requested amendments also propose a change to the CIM internal power supply that will enable proper functioning of the field programmable gate arrays. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change to the CIM internal power supply enables the field programmable gate array (FPGA) to function properly. The proposed change to the FPGA core power has no adverse effect on the operation of the output actuation relays. The function of the internal power supply has no input to plant safety analysis. The change to the CIM internal power supply has a negligible effect on the 24 Vdc [volts direct current] supplies and ultimately the plant electrical system load and has no adverse effect on the CIM functionality. The proposed changes to clarify how licensing basis design documentation reflects compliance with license basis requirements, and the proposed change to the ownership of safety remote node controller (SRNC) and CIM intellectual property, are not technical changes. The proposed changes do not affect any accident initiator in the UFSAR, or affect the radioactive material releases in the UFSAR accident analyses. The proposed change does not alter the ability of the facility to prevent and mitigate abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, or component (SSC) or function is adversely affected. The change does not involve or interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. This activity does not involve a new fission product release path, nor a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. Because the proposed changes do [[Page 73441]] not change any safety-related SSC or function credited in the mitigation of an accident, the consequences of the accidents evaluated in the UFSAR are not affected. Therefore, the requested amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to the CIM internal power supply enables the FPGA to function properly and does not involve accident initiators. The change to the CIM internal power supply has a negligible effect on the 24 Vdc supplies and ultimately the plant electrical system load and has no adverse effect on CIM functionality. The proposed clarified descriptions and the proposed change to the ownership of SRNC and CIM intellectual property are not technical changes. The proposed changes do not affect other plant equipment or adversely affect the design of the CIM. Therefore, the proposed changes do not affect any safety-related equipment itself, nor do they affect equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. No system or design function or equipment qualification would be adversely affected by the proposed changes. Furthermore, the proposed changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to the CIM internal power supply enables the FPGA to function properly. The function of the internal power supply has no input to plant safety analysis. The change to the CIM internal power supplies has a negligible effect on the 24 Vdc supplies and ultimately the plant electrical system load and has no adverse effect on the CIM functionality. The proposed clarified descriptions and the proposed change to the ownership of SRNC and CIM intellectual property are not technical changes. The proposed changes do not adversely affect the design, construction, or operation of any plant SSCs, including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes. No safety analysis or design basis acceptance limit/ criterion is challenged or exceeded by the proposed changes, thus no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. NRC Branch Chief: Michael T. Markley. Southern Nuclear Operating Company, Docket No. 50-364, Joseph M. Farley Nuclear Plant, Unit 2, Houston County, Alabama Date of amendment request: September 8, 2016. A publicly-available version is in ADAMS under Accession No. ML16256A135. Description of amendment request: The amendment would correct an error in the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility Operating License No. NPF-8, for Condition 2.C.(23). Specifically, the Unit 2 referenced date prior to the period of extended operation was incorrectly entered as June 25, 2017. This date corresponds to the Unit 1 period of extended operation. The Unit 2 correct date for this license condition is March 31, 2021. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC). There are no accidents affected by this change, and therefore no 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 amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC). There are no accidents affected by this change, and therefore no possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC). There are no accidents affected by this change, and therefore no 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, Inc., 40 Inverness Center Parkway, Birmingham, AL 35242. NRC Branch Chief: Michael T. Markley. Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: July 27, 2016, as supplemented by letter dated September 13, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16210A001 and ML16257A598, respectively. Description of amendment request: The amendments would revise Technical Specification 3.6.4.1, ``Secondary Containment,'' Surveillance Requirement (SR) 3.6.4.1.3 to provide an allowance for brief, inadvertent, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions. 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, along with NRC edits in square brackets: 1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated? Response: No. The proposed change does not involve any physical change to structures, systems, or components (SSCs) and do not alter the method of operation of any SSCs. The proposed change addresses a temporary condition during which Secondary Containment SRs are not met. The Secondary Containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. [Two accidents credit the Secondary Containment from a dose consequence perspective. They are the [[Page 73442]] loss-of-coolant accident (LOCA) and fuel/equipment handling accident. Each accident requires time to drawdown the secondary containment to less than atmospheric pressure. The brief, inadvertent, simultaneous opening of both an inner and outer personnel access door during normal entry and exit conditions followed by prompt closure does not challenge the design basis drawdown time and does not result in an increase in any on-site or offsite dose for the LOCA dose analysis. All dose consequences are within the regulatory limits established for the fuel handling accident and bound the case in which airlock doors are briefly, inadvertently opened.] As a result, the consequences of any accident previously evaluated is not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration of any plant equipment. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are not setpoints, at which protective or mitigative actions are initiated, affected by the proposed change. The proposed change does not alter the manner in which equipment operation is initiated, nor will the function of credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event described in the FSAR [Final Safety Analysis Report]. As such, no new failure modes are being introduced. The change does not alter the assumptions made in the safety analysis and licensing basis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change addresses temporary conditions during which the Secondary Containment SR is not met. The allowance for both an inner and outer Secondary Containment access door to be open simultaneously for entry and exit does not affect the safety function of the reactor enclosure and refuel area Secondary Containments as the doors are promptly closed after entry of exit, thereby restoring the Secondary Containment boundary. In addition, brief, inadvertent simultaneous opening and closing of redundant Secondary Containment personnel access doors during normal entry and exit conditions does not affect the ability of the SGTS to establish the required Secondary Containment vacuum. Therefore, the safety function of the Secondary Containment is not affected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Damon D. Obie, Esquire, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101. NRC Branch Chief: Douglas A. Broaddus. Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant (WBN), Unit 2, Rhea County, Tennessee Date of amendment request: September 30, 2016. A publicly-available version is in ADAMS under Accession No. ML16277A477. Description of amendment request: The amendment would revise the Technical Specifications (TSs) to allow a one-time extension of the frequency for performing TS Surveillance Requirements (SRs) related to verifying the operability of the containment ice bed. 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 edits in brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The requested action is a one-time extension to the performance interval for TS SRs [3.6.11.2] and 3.6.11.3. The performance of these surveillances, or the extension of these surveillances, is not a precursor to an accident. Performing these surveillances or failing to perform these surveillances does not affect the probability of an accident. Therefore, the proposed delay in performance of the SRs in this amendment request does not increase the probability of an accident previously evaluated. A delay in performing these surveillances does not result in a system being unable to perform its required function. In the case of this one-time extension request, the short period of additional time that the systems and components will be in service before the next performance of the surveillance will not affect the ability of those systems to operate as designed. Therefore, the systems required to mitigate accidents will remain capable of performing their required function. No new failure modes have been introduced because of this action and the consequences remain consistent with previously evaluated accidents. On this basis, the proposed delay in performance of the SRs in this amendment request does not involve a significant increase in the consequences of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC) or a change in the way any SSC is operated. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the one-time SR extensions being requested. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment is a one-time extension of the performance interval of TS SRs [3.6.11.2] and 3.6.11.3. Extending these surveillance requirements does not involve a modification of any TS limiting conditions for operation. Extending these SRs does not involve a change to any limit on accident consequences specified in the license or regulations. Extending these SRs does not involve a change in how accidents are mitigated or a significant increase in the consequences of an accident. Extending these SRs does not involve a change in a methodology used to evaluate consequences of an accident. Extending these SRs does not involve a change in any operating procedure or process. Based on the limited additional period of time that the systems and components will be in service before the surveillances are next performed, as well as the operating experience that these surveillances are typically successful when performed, it is reasonable to conclude that the margins of safety associated with these SRs will not be affected by the requested extension. 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: Ms. Sherry A. Quirk, Executive Vice President and General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Acting Branch Chief: Jeanne A. Dion. [[Page 73443]] Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia Date of amendment request: May 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16146A540. Description of amendment request: The amendments would revise the Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS) 3.14, ``Circulating and Service Water Systems,'' to extend the allowed outage time (AOT) for only one operable service water (SW) flow path to the charging pump service water (CPSW) subsystem and to the main control room/emergency switchgear room (MCR/ESGR) air conditioning (AC) subsystem. TS 3.14.A.5 and TS 3.14.A.7 require two SW flow paths to the CPSW subsystem and to the MCR/ESGR AC subsystem, respectively, to be operable. Currently, the TS 3.14.C AOT for only one operable CPSW or MCR/ESGR AC flow path is 24 hours. The proposed revision will extend the AOT for only one operable CPSW or MCR/ESGR AC flow path from 24 hours to 72 hours. 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 extends the AOT for only one operable CPSW or MCR/ESGR AC flow path from 24 hours to 72 hours. The CPSW subsystem is a support system for the Charging/High Head Safety Injection (HHSI) pumps; the proposed CPSW AOT extension aligns the CPSW support system AOT with the AOT for the supported components (i.e., the Charging/HHSI pumps). The proposed MCR/ESGR AC AOT extension revises the AOT to be the same as the CPSW AOT since both subsystems share common piping. The design function of the CPSW system, which is to provide cooling to the charging pump intermediate seal coolers and the charging pump lubricating oil coolers, is not impacted by the proposed revision, nor is the design function of the Charging/HHSI pumps impacted. Furthermore, the design functions of the MCR/ESGR AC subsystem and the MCR/ESGR chillers are not impacted by the proposed revision. In addition, the proposed change deletes the now expired and no longer necessary requirements for the temporary SW jumper to the CCHXs [component cooling heat exchangers]. The deletion of these temporary requirements is administrative in nature. As a result, 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 extends the AOT for only one operable CPSW or MCR/ESGRAC flow path from 24 hours to 72 hours. In addition, the proposed change deletes the now expired and no longer necessary requirements for the temporary SW jumper to the CCHXs. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) and does not impact plant operation. Furthermore, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. 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 extends the AOT for only one operable CPSW or MCR/ESGR AC flow path from 24 hours to 72 hours. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. There are no changes being made to any safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Furthermore, as noted above, a supporting PRA [probabilistic risk assessment] was performed for the proposed AOT changes. The PRA concluded that the increase in risk associated with the proposed changes is consistent with the RG [Regulatory Guide] 1.174 and RG 1.177 acceptance guidelines for a permanent TS AOT change. This PRA evaluation demonstrates that defense-in-depth will not be significantly impacted by changing the AOTs for only one operable SW flow path to the CPSW subsystem and to the MCR/ESGR AC subsystem from 24 to 72 hours. In addition, the proposed change deletes the now expired and no longer necessary requirements for the temporary SW jumper to the CCHXs. The deletion of these temporary requirements is administrative in nature. 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219. NRC Branch Chief: Michael T. Markley. Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia Date of amendment request: July 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16202A068. Description of amendment request: The amendments would revise the Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS) 3.14, ``Circulating and Service Water Systems,'' to extend the allowed outage time (AOT) for emergency service water (ESW) pump inoperability. The proposed revision would extend the TS 3.14.B AOT for one inoperable ESW pump from 7 days to 14 days to provide operational flexibility for ESW pump maintenance and repairs. 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 design function of the ESW pumps is to ensure that water can be provided to the intake canal (i.e., the ultimate heat sink) when power is not available to the Circulating Water (CW) pumps. The proposed extension of the AOT for one inoperable ESW pump from 7 to 14 days does not impact the design function of the ESW pumps. In addition, the number of ESW pumps required to be operable for the specified plant operating conditions is not impacted by the proposed AOT extension. As a result, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) and does not impact plant operation. Furthermore, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different [[Page 73444]] kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. There are no changes being made to any safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Furthermore, as noted above, a supporting PRA [probabilistic risk assessment] was performed for the proposed AOT change. The PRA concluded that the increase in risk associated with the proposed change is consistent with the RG [Regulatory Guide] 1.174 and RG 1.177 acceptance guidelines for a permanent TS AOT change. This PRA evaluation demonstrates that defense-in-depth will not be significantly impacted by changing the AOT for one inoperable ESW pump from 7 to 14 days. 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219. NRC Branch Chief: Michael T. Markley. 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. Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2 (MPS2), New London County, Connecticut Date of amendment request: September 1, 2015, as supplemented by letter dated March 24, 2016. Brief description of amendment: The amendment revised the MPS2 Technical Specifications (TSs) to add the evaluation model EMF- 2328(P)(A), Supplement 1, ``PWR [pressurized water reactor] Small Break LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based,'' and EMF-92-116(P)(A), Supplement 1, ``Generic Mechanical Design Criteria for PWR Fuel Designs,'' to the TS Section 6.9.1.8.b list of analytical methods used to determine core operating limits as a result of reanalyzing the small break LOCA. Date of issuance: September 30, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 329. A publicly-available version is in ADAMS under Accession No. ML16249A001; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-65: Amendment revised the Renewed Operating License and TSs. Date of initial notice in Federal Register: December 8, 2015 (80 FR 76318). The supplemental letter dated March 24, 2016, 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 September 30, 2016. No significant hazards consideration comments received: No. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1 (ANO-1), Pope County, Arkansas Date of amendment request: January 29, 2014, as supplemented by letters dated May 19, June 16, July 21, August 12, September 22, November 4, and November 17, 2015; and January 15, March 25, April 7, May 19, and August 29, 2016. Brief description of amendment: The amendment authorized the transition of the ANO-1 fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods such as fire modeling and risk-informed methods such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria. Date of issuance: October 7, 2016. Effective date: As of the date of issuance and shall be implemented as described in the transition license conditions. Amendment No.: 256. A publicly-available version is in ADAMS under Accession No. ML16223A481; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-51: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 8, 2014 (79 FR 38589). The supplemental letters dated May 19, June 16, July 21, August 12, September 22, November 4, and November 17, 2015; and January 15, March 25, April 7, May 19, and August 29, 2016, 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 October 7, 2016. No significant hazards consideration comments received: No. [[Page 73445]] Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: October 15, 2015, as supplemented by a letter dated May 6, 2016. Description of amendment request: The amendments revised the St. Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) and licensing bases to reflect the use of the commercially available computer code ``Generation of Thermal-Hydraulic Information for Containments (GOTHIC Version 7.2b(QA))'' to model the containment response following the inadvertent actuation of the containment spray system during normal plant operation (referred to as the vacuum analysis). The amendments also updated the licensing bases to credit the design basis ability of the containment vessel to withstand a higher external pressure differential of 1.04 pounds per square inch (psi) for Unit No. 1 and 1.05 psi for Unit No. 2, and updated TS 3.6.1.4 for each unit to revise the allowable containment operating pressure range. Date of issuance: October 5, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 236 (Unit No. 1) and 186 (Unit No. 2). A publicly- available version is in ADAMS under Accession No. ML16166A424; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: February 16, 2016 (81 FR 7839). The supplemental letter dated May 6, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in an SE dated October 5, 2016. No significant hazards consideration comments received: No. Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: November 12, 2015, as supplemented by letters dated December 9, 2015, and March 14, March 29, April 7, April 20, August 16, September 16, September 21, and September 29, 2016. Brief description of amendments: By order dated May 6, 2016, as published in the Federal Register on May 23, 2016 (81 FR 32350), the NRC approved the transfer of Facility Operating License (FOL) Nos. NPF- 87 and NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, and the general license for the independent spent fuel storage installation facility from the current holder, Luminant Generation Company LLC, to Comanche Peak Power Company LLC, as owner, and TEX Operations Company LLC, as operator. The conforming amendments revised the FOLs to reflect the direct transfer of ownership and the indirect transfer of control of the licenses. Date of issuance: October 3, 2016. Effective date: As of the date of issuance and shall be implemented within 7 days from the date of issuance. Amendment Nos.: 167 (Unit No. 1) and 167 (Unit No. 2). A publicly- available version is in ADAMS under Accession No. ML16266A005; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-87 and NPF-89: Amendments revised the FOLs. Date of initial notice in Federal Register: February 8, 2016 (81 FR 6545). The supplemental letters dated March 14, March 29, April 7, April 20, August 16, September 16, September 21, and September 29, 2016, provided additional information that clarified the application and did not expand the scope of the application as originally noticed. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 6, 2016. PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date amendment request: October 12, 2015. Brief description of amendments: The amendments revised Salem Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications (TSs) by adding MODE 4 to the applicability of TS 3.6.2.3, ``Containment Cooling System.'' The amendments also revised TS 3.7.1.1, ``Safety Valves,'' to correct discrepancies between the applicable modes and the action statements. Date of issuance: September 29, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 315 (Unit No. 1) and 296 (Unit No. 2). A publicly- available version is in ADAMS under Accession No. ML16229A519; 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 TSs. Date of initial notice in Federal Register: January 5, 2016 (81 FR 264). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2016. No significant hazards consideration comments received: No. South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: March 14, 2016, as supplemented by letters dated May 12, 2016, and July 12, 2016. Description of amendment: The amendments incorporated changes that are consistent with those generically approved in WCAP-17524-P-A, Revision 1, ``AP1000 Core Reference Report,'' dated February 19, 2015. The amendments also approved changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 licensing basis information, involves changes to the UFSAR information that has been designated as Tier 2* information, and involves changes to the plant- specific Technical Specifications. Date of issuance: September 20, 2016. Effective date: As of the date of issuance and shall be implemented within 50 days of issuance. Amendment Nos.: 52 (Unit 2) and 52 (Unit 3). A publicly-available version is in ADAMS under Package Accession No. ML16144A591; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendments revised the Facility Combined Licenses and Technical Specifications. Date of initial notice in Federal Register: May 10, 2016 (81 FR 28900). The supplemental letters dated May 12, 2016, and July 12, 2016, provided additional information that clarified the [[Page 73446]] 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 September 20, 2016. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: October 10, 2014, as supplemented by letters dated May 4, 2015; October 15, 2015; and August 26, 2016. Brief description of amendments: The amendments revised Technical Specifications (TSs) by adopting 18 previously NRC-approved Technical Specifications Task Force (TSTF) Travelers, two TSTF T-Travelers, and one feature of the Improved Standard Technical Specifications not associated with a Traveler. Date of issuance: September 29, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 279 (Unit 1) and 223 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16231A041; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: March 31, 2015 (80 FR 17095). The supplemental letters dated May 4, 2015; October 15, 2015; and August 26, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the 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 September 29, 2016. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; City of Dalton, Georgia; Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request: October 15, 2015, as supplemented by letters dated March 16, May 9, and May 16, 2016. Brief description of amendments: The amendments revised the Technical Specifications Surveillance Requirement 3.6.4.1.3 to increase the allowable time from 2 minutes to 10 minutes for the standby gas treatment system to draw down the secondary containment to negative pressure. Date of issuance: September 30, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 280 (Unit No. 1) and 224 (Unit No. 2). A publicly- available version is in ADAMS under Accession No. ML16235A287; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: November 24, 2015 (80 FR 73240). The supplemental letters dated March 16, May 9, and May 16, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2016. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee Date of amendment request: May 16, 2016. Brief description of amendments: The amendments consisted of change to the Completion Date of Cyber Security Plan (CSP) Implementation Milestone 8--full implementation of the CSP from October 31, 2016 to December 31, 2017. Date of issuance: October 3, 2016. Effective date: As of its date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 337 (Unit 1) and 330 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16228A096; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Facility Operating License Nos. DPR-77 and DPR-79. The amendments revised the Technical Specifications. Date of initial notice in Federal Register: July 8, 2016 (81 FR 44665). The Commission's related evaluation of the amendments is contained in an SE dated October 3, 2016. No significant hazards consideration comments received: No. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date of amendment request: March 11, 2016, as supplemented by letters dated May 31, 2016, and July 22, 2016. Brief description of amendments: The amendments revised the Technical Specifications (TSs) by adding a new Condition A to TS 3.7.8, ``Essential Raw Cooling Water (ERCW) System,'' to extend the allowed completion time to restore ERCW System train to OPERABLE status from 72 hours to 7 days for planned maintenance when the opposite unit is defueled or in Mode 6 following defueled under certain restrictions. Date of issuance: September 29, 2016. Effective date: As of its date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 336 (Unit 1) and 329 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16225A276; documents related to these amendments are listed in the Safety Evaluation (SE) enclosed with the amendments. Facility Operating License Nos. DPR-77 and DPR-79. Amendments revised the TSs. Date of initial notice in Federal Register: April 12, 2016 (81 FR 21603). The supplemental letters dated May 31, 2016, and July 22, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in an SE dated September 29, 2016. [[Page 73447]] No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 13th day of October 2016. For the Nuclear Regulatory Commission. George A. Wilson, Jr., Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2016-25641 Filed 10-24-16; 8:45 am] BILLING CODE 7590-01-P
| 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 November 25, 2016. A request for a hearing must be filed by December 27, 2016. | |
| Contact | Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1927, email: [email protected] | |
| FR Citation | 81 FR 73428 |