81 FR 73428 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 81, Issue 206 (October 25, 2016)
Federal Register Volume 81, Issue 206 (October 25, 2016)
| 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
![73428 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
(open to the public) will convene at as soon as possible by contacting Ms. • Mail comments to: Cindy Bladey,
10:30 a.m. until approximately 11:30 Katherine Griffin at (202) 606–8322 or Office of Administration, Mail Stop:
a.m., followed by the discussion of kgriffin@neh.gov. Please also provide OWFN–12–H08, U.S. Nuclear
specific grant applications and programs advance notice of any special needs or Regulatory Commission, Washington,
before the Council (closed to the public) accommodations, including for a sign DC 20555–0001.
from 11:30 a.m. until 12:30 p.m. language interpreter. For additional direction on obtaining
Challenge Grants: Room 4089 Dated: October 19, 2016. information and submitting comments,
Education Programs: Conference Elizabeth Voyatzis,
see ‘‘Obtaining Information and
Room C. Submitting Comments’’ in the
Committee Management Officer.
Federal/State Partnership: 2002 SUPPLEMENTARY INFORMATION section of
[FR Doc. 2016–25723 Filed 10–24–16; 8:45 am]
Preservation and Access Programs: this document.
Room P002. BILLING CODE 7536–01–P
FOR FURTHER INFORMATION CONTACT:
Public Programs: Room P003 Lynn Ronewicz, Office of Nuclear
Research Programs/Digital Reactor Regulation, U.S. Nuclear
Humanities: Room 4002. NUCLEAR REGULATORY Regulatory Commission, Washington DC
The plenary session of the National COMMISSION 20555–0001; telephone: 301–415–1927,
Council on the Humanities will convene email: Lynn.Ronewicz@nrc.gov.
[NRC–2016–0214]
on November 18, 2016, at 9:00 a.m. in
the Conference Center at Constitution I. Obtaining Information and
Biweekly Notice; Applications and Submitting Comments
Center. The agenda for the morning Amendments to Facility Operating
session (open to the public) will be as Licenses and Combined Licenses A. Obtaining Information
follows: Involving No Significant Hazards Please refer to Docket ID NRC–2016–
A. Minutes of the Previous Meeting Considerations 0214, facility name, unit number(s),
B. Reports plant docket number, application date,
1. Chairman’s Remarks AGENCY: Nuclear Regulatory
Commission. and subject when contacting the NRC
2. Deputy Chairman’s Remarks about the availability of information for
3. Conversation with NEH Chairman ACTION: Biweekly notice.
this action. You may obtain publicly-
William D. Adams and Mellon available information related to this
Foundation President Earl Lewis SUMMARY: Pursuant to Section 189a.(2)
of the Atomic Energy Act of 1954, as action by any of the following methods:
4. Congressional Affairs Report • Federal Rulemaking Web site: Go to
5. Budget Report amended (the Act), the U.S. Nuclear
Regulatory Commission (NRC) is http://www.regulations.gov and search
6. Reports on Policy and General for Docket ID NRC–2016–0214.
Matters publishing this regular biweekly notice.
The Act requires the Commission to • NRC’s Agencywide Documents
a. Challenge Grants Access and Management System
b. Education Programs publish notice of any amendments
issued, or proposed to be issued, and (ADAMS): You may obtain publicly-
c. Federal/State Partnership available documents online in the
d. Preservation and Access Programs grants the Commission the authority to
issue and make immediately effective ADAMS Public Documents collection at
e. Public Programs http://www.nrc.gov/reading-rm/
f. Research Programs any amendment to an operating license
or combined license, as applicable, adams.html. To begin the search, select
g. Digital Humanities ‘‘ADAMS Public Documents’’ and then
The remainder of the plenary session upon a determination by the
Commission that such amendment select ‘‘Begin Web-based ADAMS
will be for consideration of specific Search.’’ For problems with ADAMS,
applications and therefore will be involves no significant hazards
consideration, notwithstanding the please contact the NRC’s Public
closed to the public. Document Room (PDR) reference staff at
As identified above, portions of the pendency before the Commission of a
request for a hearing from any person. 1–800–397–4209, 301–415–4737, or by
meeting of the National Council on the email to pdr.resource@nrc.gov. The
Humanities will be closed to the public This biweekly notice includes all
notices of amendments issued, or ADAMS accession number for each
pursuant to sections 552b(c)(4), document referenced (if it is available in
552b(c)(6) and 552b(c)(9)(b) of Title 5 proposed to be issued, from September
27, 2016, to October 7, 2016. The last ADAMS) is provided the first time that
U.S.C., as amended. The closed sessions it is mentioned in this document.
biweekly notice was published on
will include review of personal and/or
October 11, 2016. • NRC’s PDR: You may examine and
proprietary financial and commercial purchase copies of public documents at
information given in confidence to the DATES: Comments must be filed by the NRC’s PDR, Room O1–F21, One
agency by grant applicants, and November 25, 2016. A request for a White Flint North, 11555 Rockville
discussion of certain information, the hearing must be filed by December 27, Pike, Rockville, Maryland 20852.
premature disclosure of which could 2016.
significantly frustrate implementation of ADDRESSES: You may submit comments B. Submitting Comments
proposed agency action. I have made by any of the following methods. Please include Docket ID NRC–2016–
this determination pursuant to the • Federal Rulemaking Web site: Go to 0214, facility name, unit number(s),
authority granted me by the Chairman’s http://www.regulations.gov and search plant docket number, application date,
Delegation of Authority to Close for Docket ID NRC–2016–0214. Address and subject in your comment
mstockstill on DSK3G9T082PROD with NOTICES
Advisory Committee Meetings dated questions about NRC dockets to Carol submission.
April 15, 2016. Gallagher; telephone: 301–415–3463; The NRC cautions you not to include
Please note that individuals planning email: Carol.Gallagher@nrc.gov. For identifying or contact information that
to attend the public sessions of the technical questions, contact the you do not want to be publicly
meeting are subject to security screening individual listed in the FOR FURTHER disclosed in your comment submission.
procedures. If you wish to attend any of INFORMATION CONTACT section of this The NRC will post all comment
the public sessions, please inform NEH document. submissions at http://
VerDate Sep<11>2014 16:58 Oct 24, 2016 Jkt 241001 PO 00000 Frm 00051 Fmt 4703 Sfmt 4703 E:\FR\FM\25OCN1.SGM 25OCN1](https://thefederalregister.org/doc/81_FR_73634/page/1.svg)
![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73431
privacy information, such as social EAL scheme. KPS has permanently ceased Agency protective action guideline] exposure
security numbers, home addresses, or operation and is permanently defueled. The levels, as detailed in the EPA’s ‘‘Protective
home phone numbers in their filings, proposed amendment is conditioned on all Action Guide and Planning Guidance for
spent nuclear fuel being removed from wet Radiological Incidents,’’ Draft for Interim Use
unless an NRC regulation or other law
storage in the spent fuel pool and placed in and Public Comment dated March 2013 (PAG
requires submission of such dry storage within the ISFSI. Occurrence of Manual).
information. However, in some postulated accidents associated with spent The proposed amendment does not involve
instances, a petition will require fuel stored in a spent fuel pool is no longer a change in the plant’s design, configuration,
including information on local credible in a spent fuel pool devoid of such or operation. The proposed amendment does
residence in order to demonstrate a fuel. The proposed amendment has no effect not affect either the way in which the plant
proximity assertion of interest in the on plant systems, structures, and components structures, systems, and components perform
proceeding. With respect to copyrighted (SSCs) and no effect on the capability of any their safety function or their design margins.
plant SSC to perform its design function. The Because there is no change to the physical
works, except for limited excerpts that
proposed amendment would not increase the design of the plant, there is no change to any
serve the purpose of the adjudicatory likelihood of the malfunction of any plant of these margins.
filings and would constitute a Fair Use SSC. The proposed amendment would have Therefore, the proposed changes do not
application, participants are requested no effect on any of the previously evaluated involve a significant reduction in a margin of
not to include copyrighted materials in accidents in the KPS Updated Safety safety.
their submission. Analysis Report (USAR).
The Commission will issue a notice or Since KPS has permanently ceased The NRC staff has reviewed the
order granting or denying a hearing operation, the generation of fission products licensee’s analysis and, based on this
request or intervention petition, has ceased and the remaining source term review, it appears that the three
designating the issues for any hearing continues to decay. This continues to standards of 10 CFR 50.92(c) are
significantly reduce the consequences of satisfied. Therefore, the NRC staff
that will be held and designating the previously postulated accidents.
Presiding Officer. A notice granting a proposes to determine that the
Therefore, the proposed amendment does amendment request involves no
hearing will be published in the Federal not involve a significant increase in the
Register and served on the parties to the consequences of a previously evaluated
significant hazards consideration.
hearing. accident. Attorney for licensee: Lillian M.
For further details with respect to 2. Do the proposed changes create the Cuoco, Senior Counsel, Dominion
these license amendment applications, possibility of a new or different kind of Resource Services, Inc., 120 Tredegar
see the application for amendment, accident from any accident previously Street, RS–2, Richmond, VA 23219.
which is available for public inspection evaluated? NRC Branch Chief: Bruce A. Watson.
Response: No.
in ADAMS and at the NRC’s PDR. For The proposed amendment constitutes a Dominion Energy Kewanee, Inc. (DEK),
additional direction on accessing revision of the emergency planning function Docket No. 50–305, Kewanee Power
information related to this document, commensurate with the ongoing and Station (KPS), Carlton, Wisconsin
see the ‘‘Obtaining Information and anticipated reduction in radiological source Date of amendment request: July 28,
Submitting Comments’’ section of this term at KPS.
The proposed amendment does not involve
2016. A publicly available version is in
document.
a physical alteration of the plant. No new or ADAMS under Accession No.
Dominion Energy Kewanee, Inc. (DEK), different types of equipment will be installed ML16216A187.
Docket No. 50–305, Kewanee Power and there are no physical modifications to Description of amendment request:
Station (KPS), Carlton, Wisconsin existing equipment as a result of the The amendment would revise the KPS
Date of amendment request: proposed amendment. Updated Safety Analysis Report (USAR)
Similarly, the proposed amendment would Section 9.5.2.2.4, ‘‘Auxiliary Building
September 14, 2015. A publicly not physically change any SSCs involved in
available version is in ADAMS under Crane,’’ to: (1) Add a description of a
the mitigation of any postulated accidents. non-single failure proof intermediate
Accession No. ML15261A238. Thus, no new initiators or precursors of a
Description of amendment request: lifting device that DEK intends to use
new or different kind of accident are created.
The amendment would revise the KPS Furthermore, the proposed amendment does during a specific spent fuel cask
Permanently Defueled Emergency Plan not create the possibility of a new failure handling activity in the auxiliary
(PDEP) and the Permanently Defueled mode associated with any equipment or building, and (2) incorporate a new load
Emergency Action Levels (EAL) Bases personnel failures. The credible events for drop analysis applicable to the use of
Document. DEK requests revisions of the ISFSI remain unchanged. this intermediate lifting device. The
Therefore, the proposed amendment does amendment also includes (for
the PDEP and the EAL Bases Document not create the possibility of a new or different
that reflect DEK’s plan to transfer all information) a new Technical
kind of accident from any previously Requirements Manual section that
spent fuel to the independent spent fuel evaluated.
storage installation (ISFSI). governs the use of the non-single failure
3. Do the proposed changes involve a
Basis for proposed no significant significant reduction in the margin of safety? proof intermediate lifting device to
hazards consideration determination: Response: No. ensure compliance with the required
As required by 10 CFR 50.91(a), the Because the 10 CFR part 50 license for KPS parameters in the load drop analysis.
licensee has provided its analysis of the no longer authorizes operation of the reactor Basis for proposed no significant
issue of no significant hazards or emplacement or retention of fuel into the hazards consideration determination:
reactor vessel, as specified in 10 CFR As required by 10 CFR 50.91(a), the
consideration, which is presented
50.82(a)(2), the occurrence of postulated licensee has provided its analysis of the
below: accidents associated with reactor operation is
mstockstill on DSK3G9T082PROD with NOTICES
issue of no significant hazards
1. Do the proposed changes involve a no longer credible. With all nuclear spent
consideration, which is presented
significant increase in the probability or fuel pool transferred out of wet storage from
consequences of an accident previously the spent fuel pool and placed in dry storage below:
evaluated? within the ISFSI, a fuel handling accident is 1. Do the proposed changes involve a
Response: No. no longer credible. There are no longer significant increase in the probability or
The proposed amendment would modify credible events that would result in any consequences of an accident previously
the KPS renewed facility operating license by releases beyond the site boundary exceeding evaluated?
revising the emergency plan and revising the the EPA PAG [Environmental Protection Response: No.
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73433
Response: No. Explosive Gas Mixture TS requirements The proposed changes do not involve a
The proposed changes eliminate the and Liquid Holdup Tanks TS significant reduction in a margin of safety
operational requirements and certain design requirements to a licensee-controlled and are considered administrative in nature.
requirements associated with the storage of program in the Procedures and The proposed changes do not involve any
the spent fuel in the spent fuel pools, and actual change in the methodology used in the
relocate certain administrative controls to the
Programs TSs section, and (3) modify monitoring of explosive gas mixtures
Quality Assurance Program Description or the Gas Storage Tank Radioactivity contained in the Gaseous Waste Processing
other licensee controlled document. Monitoring Program TSs into an System. HNP does not currently utilize
After the removal of the spent fuel from the Explosive Gas and Storage Tank unprotected outdoor liquid storage tanks;
spent fuel pools and transfer to the ISFSI, Radioactivity Monitoring Program to therefore, there are no associated
there are no spent fuel assemblies that include controls for potentially methodology changes with this request.
remain in the spent fuel pools. Coupled with explosive gas mixtures and the quantity These changes provide for the relocation of
a prohibition against storage of fuel in the of radioactivity contained in procedural details outside of the technical
spent fuel pools, the potential for fuel related unprotected outdoor liquid storage specifications with the addition of
accidents is removed. The proposed changes appropriate administrative controls to
do not introduce any new failure modes.
tanks. provide continued assurance of compliance
Therefore, the proposed amendment does Basis for proposed no significant to applicable regulatory requirements.
not create the possibility of a new or different hazards consideration determination: Therefore, the proposed changes do not
kind of accident from any previously As required by 10 CFR 50.91(a), the involve a significant reduction in a margin of
evaluated. licensee has provided its analysis of the safety.
3. Does the proposed amendment involve issue of no significant hazards
a significant reduction in a margin of safety?
The NRC staff has reviewed the
consideration, which is presented licensee’s analysis and, based on this
Response: No. below:
The removal of all spent nuclear fuel from review, it appears that the three
the spent fuel pools into storage in casks 1. Does the proposed change involve a standards of 10 CFR 50.92(c) are
within an ISFSI, coupled with a prohibition significant increase in the probability or satisfied. Therefore, the NRC staff
against future storage of fuel within the spent consequences of an accident previously proposes to determine that the
fuel pools, removes the potential for fuel evaluated? amendment request involves no
related accidents. Response: No.
The proposed changes do not involve a
significant hazards consideration.
The design basis and accident assumptions Attorney for licensee: Lara Nichols,
within the CR–3 FSAR and the PDTS relating significant increase in the probability or
consequences of an accident previously Deputy General Counsel, Duke Energy
to safe management and safety of spent fuel
in the spent fuel pools are no longer evaluated. The proposed changes are Corporation, 550 South Tryon St., M/C
applicable. The proposed changes do not administrative in nature and alter only the DEC45A, Charlotte, NC 28202.
affect remaining plant operations, systems, or format and location of programmatic controls NRC Acting Branch Chief: Jeanne A.
components supporting decommissioning and procedural details relative to explosive Dion.
activities. gas monitoring and liquid holdup tanks.
The requirements for systems, structures, Existing TS containing procedural details are Energy Northwest, Docket No. 50–397,
and components (SSCs) that have been being relocated to licensee control. Columbia Generating Station, Benton
removed from the CR–3 PDTS are not Compliance with applicable regulatory County, Washington
credited in the existing accident analysis for requirements will continue to be maintained.
In addition, the proposed changes do not
Date of amendment request: July 12,
any applicable postulated accident; and as 2016. A publicly-available version is in
such, do not contribute to the margin of alter the conditions or assumptions in any of
safety associated with the accident analysis. the previous accident analyses. Because the ADAMS under Accession No.
Therefore, the proposed amendment does previous accident analyses remain bounding, ML16194A515.
not involve a significant reduction in a the radiological consequences previously Description of amendment request:
margin of safety. evaluated are not adversely affected by the The amendment would reduce the
proposed changes. minimum reactor dome pressure
The NRC staff has reviewed the Therefore, the proposed changes do not associated with the critical power
licensee’s analysis and, based on this involve a significant increase in the correlation from 785 pounds per square
review, it appears that the three probability or consequences of an accident
inch gauge (psig) to 685 psig in
standards of 10 CFR 50.92(c) are previously evaluated.
2. Does the proposed change create the Technical Specification (TS) 2.1.1,
satisfied. Therefore, the NRC staff
possibility of a new or different kind of ‘‘Reactor Core SLs [Safety Limits],’’ and
proposes to determine that the
accident from any previously evaluated? associated bases.
amendment request involves no
Response: No. Basis for proposed no significant
significant hazards consideration.
Attorney for licensee: Lara S. Nichols, The proposed changes do not create the hazards consideration determination:
possibility of a new or different kind of As required by 10 CFR 50.91(a), the
550 South Tryon Street, Charlotte, NC accident from any accident previously
28202. licensee has provided its analysis of the
evaluated. The proposed changes do not issue of no significant hazards
NRC Branch Chief: Bruce A. Watson. involve any change to the configuration or consideration, which is presented
Duke Energy Progress, LLC, Docket No. method of operation of any plant equipment.
Accordingly, no new failure modes have
below:
50–400, Shearon Harris Nuclear Power been defined for any plant system or 1. Does the proposed change involve a
Plant (HNP), Unit 1, Wake and Chatham component important to safety nor has any significant increase in the probability or
Counties, North Carolina new limiting single failure been identified as consequences of an accident previously
Date of amendment request: June 29, a result of the proposed changes. Also, there evaluated?
2016. A publicly-available version is in will be no change in types or increase in the Response: No.
mstockstill on DSK3G9T082PROD with NOTICES
ADAMS under Accession No. amounts of any effluents released offsite. The change does not involve a
Therefore, the proposed changes do not modification of any plant hardware; the
ML16182A387. create the possibility of a new or different probability and consequence of the Pressure
Description of amendment request: kind of accident from any accident Regulator Failure Open (PRFO) transient are
The amendment would revise HNP previously evaluated. essentially unchanged. The reduction in the
Technical Specifications (TSs) to (1) 3. Does the proposed change involve a reactor dome pressure safety limit (SL) from
delete the Gaseous Radwaste Treatment significant reduction in a margin of safety? 785 psig to 685 psig provides greater margin
System definition from TSs, (2) relocate Response: No. to accommodate the pressure reduction
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![73434 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
during the transient within the revised TS dated June 3, 2016, and September 19, methods used for handling and storage of
limit. 2016. Publicly available versions are such fuel in the fuel pool. The changes to the
The proposed change will continue to available in ADAMS under Accession administrative controls are administrative in
support the validity range for the correlations nature and do not affect any accidents
Nos. ML16015A456, ML16155A326,
and the calculation of Minimum Core Power applicable to the safe management of
Ratio (MCPR) as approved. The proposed TS and ML16263A237, respectively. irradiated fuel or the permanently shutdown
revision involves no significant changes to Description of amendment request: and defueled condition of the reactor.
the operation of any systems or components The licensee has provided a formal In a permanently defueled condition, the
in normal, accident or transient operating notification to the NRC of the intention only credible accident is the fuel handling
conditions. to permanently cease power operations accident.
Therefore, the proposed change does not of JAF at the end of the current The probability of occurrence of previously
involve a significant increase in the operating cycle. Once certifications for evaluated accidents is not increased, since
probability or consequences of an accident permanent cessation of operation and extended operation in a defueled condition
previously evaluated. will be the only operation allowed, and
2. Does the proposed change create the
permanent removal of fuel from the therefore bounded by the existing analyses.
possibility of a new or different kind of reactor are submitted to the NRC, Additionally, the occurrence of postulated
accident from any previously evaluated? certain staffing and training Technical accidents associated with reactor operation is
Response: No. Specifications (TSs) administrative no longer credible in a permanently defueled
The proposed reduction in the reactor controls will no longer be applicable or reactor. This significantly reduces the scope
dome pressure SL from 785 psig to 685 psig appropriate for the permanently of applicable accidents.
is a change based upon previously approved defueled condition. Therefore, ENO is Therefore, the proposed amendment does
documents and does not involve changes to requesting approval of changes to the not involve a significant increase in the
the plant hardware or its operating probability or consequences of an accident
characteristics. As a result, no new failure
staffing and training requirements in
previously evaluated.
modes are being introduced. Section 5.0, ‘‘Administrative Controls,’’ 2. Does the proposed amendment create
Therefore, the change does not introduce a of the JAF TSs. Specifically, the the possibility of a new or different kind of
new or different kind of accident from those amendment would revise and remove accident from any accident previously
previously evaluated. certain requirements in TS Sections 5.1, evaluated?
3. Does the proposed change involve a ‘‘Responsibility’’; 5.2, ‘‘Organization’’; Response: No.
significant reduction in a margin of safety? and 5.3, ‘‘Plant Staff Qualifications,’’ The proposed changes have no impact on
Response: No. and add additional definitions to TS facility SSCs affecting the safe storage of
The margin of safety is established through irradiated fuel, or on the methods of
the design of the plant structures, systems,
Section 1.1, ‘‘Definitions.’’ The
operation of such SSCs, or on the handling
and components, and through the parameters proposed amendment would not be
and storage of irradiated fuel itself. The
for safe operation and setpoints for the effective until the certification of administrative removal of or modifications of
actuation of equipment relied upon to permanent cessation of operation and the TS that are related only to administration
respond to transients and design basis certification of permanent removal of of facility cannot result in different or more
accidents. The proposed change in reactor fuel from the reactor vessel are adverse failure modes or accidents than
dome pressure enhances the safety margin, submitted to the NRC. previously evaluated because the reactor will
which protects the fuel cladding integrity The license amendment request was be permanently shutdown and defueled and
during a depressurization transient, but does originally noticed in the Federal JAF will no longer be authorized to operate
not change the requirements governing the reactor.
operation or availability of safety equipment
Register on March 1, 2016 (81 FR
10678). The notice is being reissued in The proposed deletion of requirements of
assumed to operate to preserve the margin of the JAF TS do not affect systems credited in
safety. The change does not alter the behavior its entirety to include the revised scope the accident analysis for the fuel handling
of plant equipment, which remains and description of the amendment accident at JAF. The proposed TS will
unchanged. The available pressure range is request. continue to require proper control and
expanded by the change, thus offering greater Basis for proposed no significant monitoring of safety significant parameters
margin for pressure reduction during the hazards consideration determination: and activities.
transient. As required by 10 CFR 50.91(a), the The proposed amendment does not result
Therefore, the proposed change does not licensee has provided its analysis of the in any new mechanisms that could initiate
involve a significant reduction in the margin damage to the remaining relevant safety
of safety.
issue of no significant hazards
consideration, with NRC staff revisions barriers for defueled plants (fuel cladding
The NRC staff has reviewed the and spent fuel cooling). Since extended
provided in [brackets], which is
licensee’s analysis and, based on this operation in a defueled condition will be the
presented below: only operation allowed, and therefore
review, it appears that the three
1. Does the proposed amendment involve bounded by the existing analyses, such a
standards of 10 CFR 50.92(c) are
a significant increase in the probability or condition does not create the possibility of a
satisfied. Therefore, the NRC staff consequences of an accident previously new or different kind of accident.
proposes to determine that the evaluated? Therefore, the proposed change does not
amendment request involves no Response: No. create the possibility of a new or different
significant hazards consideration. The proposed amendment would not take kind of accident from any previously
Attorney for licensee: William A. effect until JAF has permanently ceased evaluated.
Horin, Esq., Winston & Strawn, 1700 K operation and entered a permanently 3. Does the proposed amendment involve
Street NW., Washington, DC 20006– defueled condition. The proposed a significant reduction in a margin of safety?
3817. amendment would modify the JAF TS by Response: No.
NRC Branch Chief: Robert J. deleting the portions of the TS that are no Because the 10 CFR part 50 license for JAF
longer applicable to a permanently defueled will no longer authorize operation of the
mstockstill on DSK3G9T082PROD with NOTICES
Pascarelli.
facility, while modifying the other sections to reactor or emplacement or retention of fuel
Entergy Nuclear Operations, Inc. (ENO), correspond to the permanently defueled into the reactor vessel once the certifications
Docket No. 50–333, James A. FitzPatrick condition. required by 10 CFR 50.82(a)(1) are submitted,
Nuclear Power Plant (JAF), Oswego The deletion and modification of as specified in 10 CFR 50.82(a)(2), the
County, New York provisions of the administrative controls do occurrence of postulated accidents associated
not directly affect the design of structures, with reactor operation is no longer credible.
Date of amendment request: January systems, and components (SSCs) necessary The only remaining credible accident is a
15, 2016, as supplemented by letters for safe storage of irradiated fuel or the fuel handling accident (FHA). The proposed
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73435
amendment does not adversely affect the consequences of an accident previously kind of accident from any previously
inputs or assumptions of any of the design evaluated? evaluated.
basis analyses that impact the FHA. Response: No. 3. Does the proposed amendment involve
The proposed changes are limited to those The proposed change revises TS 4.0.5, a significant reduction in a margin of safety?
portions of the OL [operating license] and TS Surveillance Requirements for inservice Response: No.
that are not related to the safe storage of inspection and testing of ASME Code Class The proposed change eliminates some
irradiated fuel. The requirements that are 1, 2 & 3 components, by revising the provisions from the TS in lieu of provisions
proposed to be revised or deleted from the Inservice Testing Program and Inservice in the ASME Code, as modified by use of
JAF OL and TS are not credited in the Inspection Program specification. Code Case OMN–20 (IST) or ASME Boiler
existing accident analysis for the remaining Most requirements in the IST Program are and Pressure Vessel Code (ISI). Compliance
applicable as such, do not contribute to the removed, as they are duplicative of with the ASME Code is required by 10 CFR
margin of safety associated with the accident requirements in the ASME OM Code, as 50.55a. The proposed change also allows
analysis. Postulated DBAs [design-basis clarified by Code Case OMN–20, ‘‘Inservice inservice tests with frequencies greater than
accidents] involving the reactor are no longer Test Frequency.’’ The remaining two years to be extended by six months to
possible because the reactor will be requirements in the TS Section 4.0.5, IST facilitate test scheduling and consideration of
permanently shutdown and defueled and JAF Program are eliminated because the NRC has plant operating conditions that may not be
will no longer be authorized to operate the determined their inclusion in the TS is suitable for performance of the required
reactor. contrary to regulations. A new defined term, testing. The testing frequency extension will
Therefore, the proposed change does not ‘‘Inservice Testing Program,’’ is added to the not affect the ability of the components to
involve a significant reduction in a margin of TS, which references the requirements of 10 respond to an accident as the components are
CFR 50.55a(f). required to be operable during the testing
safety.
Similarly, the requirements in the ISI
period extension. The proposed change will
The NRC staff has reviewed the Program are revised, as they are [ ]
eliminate the existing TS SR 4.0.2 allowance
licensee’s analysis and, based on this duplicative of requirements in Section XI of
to perform a specified surveillance time
review, it appears that the three the ASME Boiler and Pressure Vessel Code
interval with a maximum allowable
standards of 10 CFR 50.92(c) are and applicable Addenda.
Performance of inservice testing or extension not to exceed 25% of the
satisfied. Therefore, the NRC staff inservice inspection is not an initiator to any surveillance interval, unless there is a
proposes to determine that the accident previously evaluated. As a result, specific SR referencing usage of the
amendment request involves no the probability of occurrence of an accident INSERVICE TESTING PROGRAM and TS SR
significant hazards consideration. is not significantly affected by the proposed 4.0.3 allowance to defer performance of
Attorney for licensee: Ms. Jeanne Cho, change. Inservice test frequencies under Code missed inservice tests up to the duration of
Case OMN–20 are equivalent to the current the specified testing frequency, and instead
Assistant General Counsel, Entergy
testing period allowed by the TS with the will require an assessment of the missed test
Nuclear Operations, Inc., 440 Hamilton on equipment operability. This assessment
Avenue, White Plains, NY 10601. exception that testing frequencies greater
than two years may be extended by up to six will consider the effect on a margin of safety
NRC Branch Chief: Travis L. Tate. months to facilitate test scheduling and (equipment operability). Should the
consideration of plant operating conditions component be inoperable, the Technical
Exelon Generation Company, LLC,
that may not be suitable for performance of Specifications provide actions to ensure that
Docket Nos. 50–352 and 50–353, the margin of safety is protected. The
the required testing. The testing frequency
Limerick Generating Station, Units 1 proposed change also eliminates a statement
extension will not affect the ability of the
and 2, Montgomery County, components to mitigate any accident that nothing in the ASME Code should be
Pennsylvania previously evaluated as the components are construed to supersede the requirements of
Date of amendment request: July 26, required to be operable during the testing any TS. However, elimination of the
period extension. Performance of inservice statement will have no effect on plant
2016. A publicly-available version is in operation or safety.
tests utilizing the allowances in OMN–20
ADAMS under Accession No. will not significantly affect the reliability of Therefore, the proposed change does not
ML16210A227. the tested components. As a result, the involve a significant reduction in a margin of
Description of amendment request: availability of the affected components, as safety.
The amendments would revise technical well as their ability to mitigate the The NRC staff has reviewed the
specification (TS) requirements relating consequences of accidents previously
evaluated, is not affected.
licensee’s analysis and, based on this
to: (1) The inservice inspection (ISI)
Therefore, the proposed change does not review, it appears that the three
program required by the American
involve a significant increase in the standards of 10 CFR 50.92(c) are
Society of Mechanical Engineers
probability or consequences of an accident satisfied. Therefore, the NRC staff
(ASME) Boiler and Pressure Code
previously evaluated. proposes to determine that the
(Code), and (2) the inservice testing 2. Does the proposed amendment create amendment request involves no
(IST) program required by the ASME the possibility of a new or different kind of significant hazards consideration.
Code for Operation and Maintenance of accident from any accident previously Attorney for licensee: Tamra Domeyer,
Nuclear Power Plants (OM Code). The evaluated?
Associate General Counsel, Exelon
proposed changes are based, in part, on Response: No.
The proposed change does not alter the Generation Company, LLC, 4300
Technical Specifications Task Force
design or configuration of the plant. The Winfield Road, Warrenville, IL 60555.
(TSTF) Traveler TSTF–545, Revision 3, NRC Branch Chief: Douglas A.
proposed change does not involve a physical
‘‘TS Inservice Testing Program Removal Broaddus.
alteration of the plant; no new or different
& Clarify SR Usage Rule Application to kind of equipment will be installed. The
Section 5.5 Testing.’’ proposed change does not alter the types of Exelon Generation Company, LLC,
Basis for proposed no significant inservice testing or inservice inspection Docket No. 50–289, Three Mile Island
hazards consideration determination: performed. In most cases, the frequency of Nuclear Station, Unit 1 (TMI–1),
mstockstill on DSK3G9T082PROD with NOTICES
As required by 10 CFR 50.91(a), the inservice testing and inservice inspection is Dauphin County, Pennsylvania
licensee has provided its analysis of the unchanged. However, the frequency of
Date of amendment request: July 15,
issue of no significant hazards testing or inspection would not result in a
new or different kind of accident from any 2016. A publicly-available version is in
consideration, which is presented below ADAMS under Package Accession No.
previously evaluated since the testing
with NRC staff edits in square brackets: methods are not altered. ML16201A306.
1. Does the proposed amendment involve Therefore, the proposed change does not Description of amendment request:
a significant increase in the probability or create the possibility of a new or different The amendment would revise the
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![73436 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
Radiological Emergency Plan Annex for containment structure) to limit the level of As required by 10 CFR 50.91(a), the
TMI–1. The proposed changes would radiation dose to the public. licensee has provided its analysis of the
decrease the radiation protection The proposed changes do not adversely issue of no significant hazards
affect existing plant safety margins or the
technician staffing from three to two reliability of the equipment assumed to
consideration, which is presented
technicians, remove two maintenance operate in the safety analyses. There are no below:
technicians currently assigned to the changes being made to safety analysis 1. Does the proposed change involve a
repair and corrective action function, assumptions, safety limits, or limiting safety significant increase in the probability or
and eliminate the on-shift Operations system settings that would adversely affect consequences of an accident previously
Support Center director position. plant safety as a result of the proposed evaluated?
Basis for proposed no significant changes. Margins of safety are unaffected by Response: No.
hazards consideration determination: the proposed changes to the ERO minimum The proposed change revises the TS for the
on-shift staffing. purpose of eliminating a non-conservative
As required by 10 CFR 50.91(a), the The proposed changes are associated with Required Action. The proposed TS change
licensee has provided its analysis of the the Emergency Plan staffing and do not does not introduce new equipment or new
issue of no significant hazards impact operation of the plant or its response equipment operating modes, nor does the
consideration, which is presented to transients or accidents. The proposed proposed change alter existing system
below: changes do not affect the Technical relationships. The proposed change does not
Specifications. The proposed changes do not affect normal plant operation. Further, the
1. Does the proposed amendment involve involve a change in the method of plant
a significant increase in the probability or proposed change does not increase the
operation, and no accident analyses will be likelihood of the malfunction of any SSC
consequences of an accident previously affected by the proposed changes. Safety
evaluated? [structure, system and component] or impact
analysis acceptance criteria are not affected any analyzed accident. Consequently, the
Response: No. by these proposed changes. The proposed
The proposed changes to the TMI probability of an accident previously
changes to the Emergency Plan will continue evaluated is not affected and there is no
Emergency Plan do not increase the to provide the necessary onsite ERO response
probability or consequences of an accident. significant increase in the consequences of
staff. any accident previously evaluated.
The proposed changes do not impact the Therefore, the proposed changes to the
function of plant Structures, Systems, or Therefore, the proposed change does not
Emergency Plan do not involve a significant involve a significant increase in the
Components (SSCs). The proposed changes reduction in a margin of safety.
do not affect accident initiators or accident probability or consequences of an accident
precursors, nor do the changes alter design The NRC staff has reviewed the previously evaluated.
assumptions. The proposed changes do not licensee’s analysis and, based on this 2. Does the proposed change create the
alter or prevent the ability of the onsite ERO review, it appears that the three possibility of a new or different kind of
[emergency response organization] to accident from any accident previously
standards of 10 CFR 50.92(c) are
perform their intended functions to mitigate evaluated?
satisfied. Therefore, the NRC staff Response: No.
the consequences of an accident or event. proposes to determine that the
The proposed changes remove onsite ERO The proposed change revises the TS for the
positions no longer credited or considered amendment request involves no purpose of eliminating a non-conservative
necessary in support of Emergency Plan significant hazards consideration. Required Action. The change does not
implementation. Attorney for licensee: Tamra Domeyer, involve a physical alteration of the plant (i.e.,
Therefore, the proposed changes to the Associate General Counsel, Exelon no new or different type of equipment will
Emergency Plan do not involve a significant Generation Company, LLC, 4300 be installed) or a change in the methods
increase in the probability or consequences Winfield Road, Warrenville, IL 60555. governing normal plant operations. The
of an accident previously evaluated. NRC Branch Chief: Douglas A. proposed change does not alter assumptions
2. Does the proposed amendment create made in the safety analysis. Further, the
Broaddus. proposed change does not introduce new
the possibility of a new or different kind of
accident from any previously evaluated? Northern States Power Company— accident initiators.
Response: No. Minnesota, Docket Nos. 50–282 and 50– Therefore, the proposed change does not
The proposed changes have no impact on 306, Prairie Island Nuclear Generating create the possibility of a new or different
the design, function, or operation of any kind of accident from any accident
Plant, Units 1 and 2, Goodhue County, previously evaluated.
plant SSCs. The proposed changes do not Minnesota
affect plant equipment or accident analyses. 3. Does the proposed change involve a
The proposed changes do not involve a Date of amendment request: August significant reduction a margin of safety?
physical alteration of the plant (i.e., no new 31, 2016. A publicly-available version is Response: No.
or different type of equipment will be in ADAMS under Accession No. The proposed change revises the TS for the
installed), a change in the method of plant purpose of eliminating a non-conservative
ML16244A493. Required Action. The proposed change does
operation, or new operator actions. The Brief description of amendment
proposed changes do not introduce failure not alter the manner in which safety limits,
request: The amendments would revise limiting safety system settings, or limiting
modes that could result in a new accident, the Required Actions and associated
and the proposed changes do not alter conditions for operation are determined. The
assumptions made in the safety analysis. The
Completion Times to Technical safety analysis assumptions and acceptance
proposed changes remove onsite ERO Specification (TS) 3.8.7, ‘‘Inverters— criteria are not affected by this change.
positions no longer credited or considered Operating.’’ Specifically, Condition B Therefore, the proposed change does not
necessary in support of Emergency Plan would be deleted and current Condition involve a significant reduction in a margin of
implementation. C would be re-lettered to Condition B. safety.
Therefore, the proposed changes to the Additionally, the Required Actions and The NRC staff has reviewed the
Emergency Plan do not create the possibility associated Completion Times for licensee’s analysis and, based on this
of a new or different kind of accident from Condition A would be modified to review, it appears that the three
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any accident previously evaluated. require restoration of one inoperable standards of 10 CFR 50.92(c) are
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
inverter to operability within 24 hours. satisfied. Therefore, the NRC staff
Response: No. These changes conform to Improved proposes to determine that the
Margin of safety is associated with Standard Technical Specification TS amendment requests involve no
confidence in the ability of the fission 3.8.7. significant hazards consideration.
product barriers (i.e., fuel cladding, reactor Basis for proposed no significant Attorney for licensee: Peter M. Glass,
coolant system pressure boundary, and hazards consideration determination: Assistant General Counsel, Xcel Energy
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73437
Services, Inc., 414 Nicollet Mall, affected by the proposed change. Inservice that the margin of safety is protected. The
Minneapolis, MN 55401. test frequencies under Code Case OMN–20 proposed change also eliminates a statement
NRC Branch Chief: David J. Wrona. are equivalent to the current testing period that nothing in the ASME Code should be
allowed by the TS with the exception that construed to supersede the requirements of
PSEG Nuclear LLC, Docket No. 50–354, testing frequencies greater than 2 years may any TS. [. . .] However, elimination of the
Hope Creek Generating Station, Salem be extended by up to 6 months to facilitate statement will have no effect on plant
County, New Jersey test scheduling and consideration of plant operation or safety.
operating conditions that may not be suitable Therefore, the proposed change does not
Date of amendment request: July 20, for performance of the required testing. The involve a significant reduction in a margin of
2016. A publicly-available version is in testing frequency extension will not affect the safety.
ADAMS under Accession No. ability of the components to mitigate any The NRC staff has reviewed the
ML16203A006. accident previously evaluated as the
licensee’s analysis and, based on this
Description of amendment request: components are required to be operable
during the testing period extension. review, it appears that the three
The amendment would revise the Hope
Performance of inservice tests utilizing the standards of 10 CFR 50.92(c) are
Creek Generating Station (Hope Creek)
allowances in OMN–20 will not significantly satisfied. Therefore, the NRC staff
Technical Specifications (TS), Section
affect the reliability of the tested proposes to determine that the
6.8.4.i, ‘‘Inservice Testing Program,’’ to components. As a result, the availability of amendment request involves no
remove requirements duplicated in the the affected components, as well as their significant hazards consideration.
American Society of Mechanical ability to mitigate the consequences of Attorney for licensee: Jeffrie J. Keenan,
Engineers (ASME) Code for Operations accidents previously evaluated, is not PSEG Nuclear LLC—N21, P.O. Box 236,
and Maintenance of Nuclear Power affected. Hancocks Bridge, NJ 08038.
Plants Case OMN–20, ‘‘Inservice Test Therefore, the proposed change does not NRC Branch Chief: Douglas A.
Frequency.’’ A new defined term, involve a significant increase in the
probability or consequences of an accident
Broaddus.
‘‘Inservice Testing Program,’’ will be
previously evaluated. South Carolina Electric & Gas Company
added to the TS 1.0, ‘‘Definitions,’’
2. Does the proposed change create the and South Carolina Public Service
section. The licensee stated that the possibility of a new or different kind of
proposed change to the TS is consistent Authority, Docket Nos. 52–027 and 52–
accident from any accident previously
with Technical Specifications Task 028, Virgil C. Summer Nuclear Station,
evaluated?
Force (TSTF) Traveler TSTF–545, Response: No. Units 2 and 3, Fairfield, South Carolina
Revision 3, ‘‘TS Inservice Testing The proposed change does not alter the Date of amendment request:
Program Removal & Clarity SR Usage design or configuration of the plant. The September 15, 2016. A publicly-
Rule Application to Section 5.5 Testing’’ proposed change does not involve a physical available version is in ADAMS under
(ADAMS Accession No. ML15294A555), alteration of the plant; no new or different Accession No. ML16259A310.
kind of equipment will be installed. The Description of amendment request:
with no proposed variations or
proposed change does not alter the types of
deviations. However, the Hope Creek TS The amendments would revise
inservice testing performed. In most cases,
uses different numbering for the frequency of inservice testing is Combined License Nos. NPF–93 and
surveillance requirements than the unchanged. However, the frequency of NPF–94 for the Virgil C. Summer
Standard Technical Specifications on testing would not result in a new or different Nuclear Station, Units 2 and 3. The
which TSTF–545 was based, so the kind of accident from any previously amendments propose changes to the
licensee changed the TSTF–545 evaluated since the testing methods are not Updated Final Safety Analysis Report
numbering to be consistent with the altered. (UFSAR) in the form of departures from
Hope Creek TS numbering. Therefore, the proposed change does not the incorporated plant-specific Design
Basis for proposed no significant create the possibility of a new or different Control Document Tier 2* information.
kind of accident from any previously Specifically, the proposed changes
hazards consideration determination: evaluated.
As required by 10 CFR 50.91(a), the would revise the Combined Licenses to
3. Does the proposed change involve a
licensee has provided its analysis of the significant reduction in a margin of safety? clarify information in WCAP–17179,
issue of no significant hazards Response: No. ‘‘AP1000® Component Interface Module
consideration, which is presented below The proposed change eliminates some Technical Report,’’ which demonstrates
with NRC staff edits in square brackets: requirements from the TS in lieu of design compliance with licensing bases
requirements in the ASME Code, as modified requirements. WCAP–17179 is
1. Does the proposed change involve a
significant increase in the probability or
by use of Code Case OMN–20. Compliance incorporated by reference into the
with the ASME Code is required by 10 CFR UFSAR to provide additional details
consequences of an accident previously
evaluated? 50.55a. The proposed change also allows regarding the component interface
Response: No. inservice tests with frequencies greater than module (CIM) system design. The
The proposed change revises TS Chapter 6, 2 years to be extended by 6 months to
facilitate test scheduling and consideration of
requested amendments also propose a
‘‘Administrative Controls,’’ Section 6.8, change to the CIM internal power
‘‘Procedures and Programs,’’ by eliminating plant operating conditions that may not be
suitable for performance of the required supply that will enable proper
the ‘‘Inservice Testing Program’’
specification. Most requirements in the testing. The testing frequency extension will functioning of the field programmable
Inservice Testing Program are removed, as not affect the ability of the components to gate arrays.
they are duplicative of requirements in the respond to an accident as the components are Basis for proposed no significant
ASME OM Code, as clarified by Code Case required to be operable during the testing hazards consideration determination:
OMN–20, ‘‘Inservice Test Frequency.’’ The period extension. The proposed change will As required by 10 CFR 50.91(a), the
remaining requirements in the Section 6.8 eliminate the existing TS 4.0.3 allowance to licensee has provided its analysis of the
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IST Program are eliminated [. . .]. A new defer performance of missed inservice tests issue of no significant hazards
defined term, ‘‘Inservice Testing Program,’’ is up to the duration of the specified testing
frequency, and instead will require an
consideration, which is presented
added to the TS, which references the
requirements of 10 CFR 50.55a(f). assessment of the missed test on equipment below:
Performance of inservice testing is not an operability. This assessment will consider 1. Does the proposed amendment involve
initiator to any accident previously the effect on a margin of safety (equipment a significant increase in the probability or
evaluated. As a result, the probability of operability). Should the component be consequences of an accident previously
occurrence of an accident is not significantly inoperable, the TS provide actions to ensure evaluated?
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![73438 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
Response: No. Furthermore, the proposed changes do not provide electrical isolation between the
The proposed change to the CIM internal result in a new failure mode, malfunction or non-Class 1E IDS battery monitors and
power supply enables the field sequence of events that could affect safety or their respective Class 1E battery banks.
programmable gate array (FPGA) to function safety-related equipment.
Because, this proposed change requires
properly. The proposed change to the FPGA Therefore, the proposed amendment does
core power has no adverse effect on the not create the possibility of a new or different a departure from Tier 1 information in
operation of the output actuation relays. The kind of accident from any accident the Westinghouse Electric Company’s
function of the internal power supply has no previously evaluated. AP1000 Design Control Document
input to plant safety analysis. The change to 3. Does the proposed amendment involve (DCD), the licensee also requested an
the CIM internal power supply has a a significant reduction in a margin of safety? exemption from the requirements of the
negligible effect on the 24 Vdc [volts direct Response: No. Generic DCD Tier 1 in accordance with
current] supplies and ultimately the plant The proposed change to the CIM internal 10 CFR 52.63(b)(1).
electrical system load and has no adverse power supply enables the FPGA to function
effect on the CIM functionality. properly. The function of the internal power Basis for proposed no significant
The proposed changes to clarify how supply has no input to plant safety analysis. hazards consideration determination:
licensing basis design documentation reflects The change to the CIM internal power As required by 10 CFR 50.91(a), the
compliance with license basis requirements, supplies has a negligible effect on the 24 Vdc licensee has provided its analysis of the
and the proposed change to the ownership of supplies and ultimately the plant electrical issue of no significant hazards
safety remote node controller (SRNC) and system load and has no adverse effect on the consideration, which is presented below
CIM intellectual property, are not technical CIM functionality.
changes. The proposed changes do not affect with NRC staff edits in square brackets:
The proposed clarified descriptions and
any accident initiator in the UFSAR, or affect the proposed change to the ownership of 1. Does the proposed amendment involve
the radioactive material releases in the SRNC and CIM intellectual property are not a significant increase in the probability or
UFSAR accident analyses. The proposed technical changes. The proposed changes do consequences of an accident previously
change does not alter the ability of the not adversely affect the design, construction, evaluated?
facility to prevent and mitigate abnormal or operation of any plant SSCs, including any Response: No.
events, e.g., accidents, anticipated equipment whose failure could initiate an The proposed changes to revise plant-
operational occurrences, earthquakes, floods accident or a failure of a fission product specific Tier 1, COL Appendix C, and
and turbine missiles, or their safety or design barrier. No analysis is adversely affected by [Updated Final Safety Analysis Report
analyses. No safety-related structure, system, the proposed changes. Furthermore, no (UFSAR)] information concerning details of
or component (SSC) or function is adversely system function, design function, or the IDS, specifically the addition of seven
affected. The change does not involve or equipment qualification will be adversely Class 1E fuse isolation panels at the
interface with any SSC accident initiator or affected by the changes. No safety analysis or interconnection of the non-Class 1E IDS
initiating sequence of events, and thus, the design basis acceptance limit/criterion is battery monitors and Class 1E IDS circuits,
probabilities of the accidents evaluated in the challenged or exceeded by the proposed are necessary to conform to Regulatory Guide
UFSAR are not affected. This activity does changes, thus no margin of safety is reduced. 1.75 Rev. 2 (consistent with UFSAR
not involve a new fission product release Therefore, the proposed amendment does Appendix 1A exceptions) and IEEE 384–1981
path, nor a new fission product barrier failure not involve a significant reduction in a to prevent a fault on non-Class 1E circuits or
mode, nor create a new sequence of events margin of safety. equipment from degrading the operation of
that would result in significant fuel cladding Class 1E IDS circuits and equipment below
failures. Because the proposed changes do The NRC staff has reviewed the an acceptable level. The proposed changes do
not change any safety-related SSC or function licensee’s analysis and, based on this not adversely affect the design functions of
credited in the mitigation of an accident, the review, it appears that the three the IDS, including the Class 1E battery banks
consequences of the accidents evaluated in standards of 10 CFR 50.92(c) are and the battery monitors.
the UFSAR are not affected. satisfied. Therefore, the NRC staff These proposed changes to revise plant-
Therefore, the requested amendment does proposes to determine that the specific Tier 1, COL Appendix C, and UFSAR
not involve a significant increase in the amendment request involves no information concerning details of the IDS,
probability or consequences of an accident specifically the addition of seven Class 1E
previously evaluated.
significant hazards consideration. fuse isolation panels at the interconnection of
2. Does the proposed amendment create Attorney for licensee: Ms. Kathryn M. the non-Class 1E IDS battery monitors and
the possibility of a new or different kind of Sutton, Morgan, Lewis & Bockius, LLC, Class 1E IDS circuits as described in the
accident from any accident previously 1111 Pennsylvania Ave. NW., current licensing basis do not have an
evaluated? Washington, DC 20004–2514. adverse effect on any of the design functions
Response: No. NRC Branch Chief: Michael T. of any plant systems. The proposed changes
The proposed change to the CIM internal Markley. do not adversely affect any plant electrical
power supply enables the FPGA to function system and do not affect the support, design,
properly and does not involve accident South Carolina Electric & Gas Company, or operation of mechanical and fluid systems
initiators. The change to the CIM internal Inc., Docket Nos. 52–027 and 52–028, required to mitigate the consequences of an
power supply has a negligible effect on the Virgil C. Summer Nuclear Station Units accident. There is no change to plant systems
24 Vdc supplies and ultimately the plant 2 and 3, Fairfield, South Carolina or the response of systems to postulated
electrical system load and has no adverse accident conditions. There is no change to
effect on CIM functionality. Date of amendment request: the predicted radioactive releases due to
The proposed clarified descriptions and September 28, 2016. A publicly- postulated accident conditions. The plant
the proposed change to the ownership of available version is in ADAMS under response to previously evaluated accidents or
SRNC and CIM intellectual property are not Accession No. ML16272A373. external events is not adversely affected, nor
technical changes. The proposed changes do Description of amendment request: do the proposed changes create any new
not affect other plant equipment or adversely The amendment request proposes accident precursors.
affect the design of the CIM. Therefore, the changes to revise plant-specific Tier 1, Therefore, the requested amendment does
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proposed changes do not affect any safety- plant-specific Tier 2, and Combined not involve a significant increase in the
related equipment itself, nor do they affect probability or consequences of an accident
License (COL) Appendix C information
equipment whose failure could initiate an previously evaluated.
accident or a failure of a fission product concerning the details of the Class 1E 2. Does the proposed amendment create
barrier. No analysis is adversely affected by direct current and uninterruptible the possibility of a new or different kind of
the proposed changes. No system or design power supply system (IDS), specifically accident from any accident previously
function or equipment qualification would be adding seven Class 1E fuse panels to the evaluated?
adversely affected by the proposed changes. IDS design. These proposed changes Response: No.
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73439
The proposed changes to revise plant- increased flow that will be experienced pumps do not affect any design code,
specific Tier 1, COL Appendix C, and UFSAR during condensate polishing system function, design analysis, safety analysis
information concerning details of the IDS, rinsing operations, for each unit, input or result, or design/safety margin. No
specifically the addition of seven Class 1E safety analysis or design basis acceptance
respectively. The proposed changes
fuse isolation panels at the interconnection of limit/criterion is challenged or exceeded by
the non-Class 1E IDS battery monitors and
include information in the combined the proposed change.
Class 1E IDS circuits, are necessary to license, Appendix C. An exemption Therefore, the proposed amendment does
conform to Regulatory Guide 1.75 Rev. 2 request relating to the proposed changes not involve a significant reduction in a
(consistent with UFSAR Appendix 1A to the AP1000 DCD Tier 1 is included margin of safety.
exceptions) and IEEE 384–1981 to prevent a with the request.
The NRC staff has reviewed the
fault on non-Class 1E circuits or equipment Basis for proposed no significant
from degrading the operation of Class 1E IDS licensee’s analysis and, based on this
hazards consideration determination:
circuits and equipment below an acceptable review, it appears that the three
As required by 10 CFR 50.91(a), the
level. The proposed changes do not adversely standards of 10 CFR 50.92(c) are
licensee has provided its analysis of the
affect any plant electrical system and do not satisfied. Therefore, the NRC staff
issue of no significant hazards
adversely affect the design function, support, proposes to determine that the
design, or operation of mechanical and fluid consideration, which is presented below
amendment request involves no
systems. The proposed changes do not result with NRC staff edits in square brackets:
significant hazards consideration.
in a new failure mechanism or introduce any 1. Does the proposed amendment involve Attorney for licensee: M. Stanford
new accident precursors. No design function a significant increase in the probability or Blanton, Balch & Bingham LLP, 1710
described in the UFSAR is adversely affected consequences of an accident previously
by the proposed changes. Sixth Avenue North, Birmingham, AL
evaluated?
Therefore, the proposed amendment does Response: No. 35203–2015.
not create the possibility of a new or different The proposed changes to identify that there NRC Branch Chief: Jennifer Dixon-
kind of accident from any accident is more than one turbine building sump and Herrity.
previously evaluated. to add two turbine building sump pumps
3. Does the proposed amendment involve (WWS–MP–07A and WWS–MP–07B) to COL
Southern Nuclear Operating Company,
a significant reduction in a margin of safety? Appendix C Subsection 2.3.29 and Inc., Docket Nos. 52–025 and 52–026,
Response: No. corresponding Table 2.3.29–1 will provide Vogtle Electric Generating Plant, Units 3
There is no safety-related [structure, consistency within the current licensing and 4, Burke County, Georgia
system, and component (SSC)] or function basis. The main turbine building sumps and Date of amendment request: August
adversely affected by the proposed change to sump pumps are not safety-related
add IDS fuse isolation panels to non-Class 1E components and do not interface with any
23, 2016. A publicly-available version is
IDS battery monitors and Class 1E IDS systems, structures, or components (SSCs) in ADAMS under Accession No.
circuits. No safety analysis or design basis accident initiator or initiating sequence of ML16236A266.
acceptance limit/criterion is challenged or events; thus, the probability of accidents Description of amendment request:
exceeded by the proposed changes and no evaluated within the [Updated Final Safety The proposed changes would amend
margin or safety is reduced. Analysis Report (UFSAR)] are not affected. Combined License Nos. NPF–91 and
Therefore, the proposed amendment does The proposed changes do not involve a NPF–92 for the Vogtle Electric
not involve a significant reduction in a change to the predicted radiological releases Generating Plant, Units 3 and 4. The
margin of safety. due to accident conditions, thus the
consequences of accidents evaluated in the
amendments propose changes to the
The NRC staff has reviewed the UFSAR are not affected. Updated Final Safety Analysis Report
licensee’s analysis and, based on this Therefore, the proposed amendment does (UFSAR) in the form of departures from
review, it appears that the three not involve a significant increase in the the incorporated plant-specific Design
standards of 10 CFR 50.92(c) are probability or consequences of an accident Control Document Tier 2 information
satisfied. Therefore, the NRC staff previously evaluated. and involve related changes to the
proposes to determine that the 2. Does the proposed amendment create Combined Operating License Appendix
the possibility of a new or different kind of C (and corresponding plant-specific
amendment request involves no
accident from any accident previously
significant hazards consideration. design control document Tier 1)
evaluated?
Attorney for licensee: Ms. Kathryn M. Response: No. information. Specifically, the proposed
Sutton, Morgan, Lewis & Bockius, LLC, The proposed changes to identify that there departures consist of changes to the
1111 Pennsylvania Ave. NW., is more than one turbine building sump and design reliability assurance program (D-
Washington, DC 20004–2514. to add two turbine building sump pumps to RAP) to identify the covers for the in-
NRC Branch Chief: Jennifer Dixon- the non-safety waste water system (WWS) do containment refueling water storage
Herrity. not affect any safety-related equipment, nor tank vents and overflow weirs as the
do they add any new interface to safety- risk-significant components included in
Southern Nuclear Operating Company, related SSCs. No system or design function the D-RAP and to differentiate between
Docket Nos. 52–025 and 52–026, Vogtle or equipment qualification is affected by
the rod drive motor-generator (MG) sets
Electric Generating Plant, Units 3 and 4, these changes. The changes do not introduce
a new failure mode, malfunction, or sequence field control relays and the rod drive
Burke County, Georgia power supply control cabinets in which
of events that could affect safety or safety-
Date of amendment request: related equipment. the relays are located.
December 22, 2015, as supplemented by Therefore, the proposed amendment does Basis for proposed no significant
letter dated July 27, 2016. Publicly- not create the possibility of a new or different hazards consideration determination:
available versions are in ADAMS under kind of accident from any accident As required by 10 CFR 50.91(a), the
Accession Nos. ML15356A655 and previously evaluated. licensee has provided its analysis of the
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ML16209A477, respectively. 3. Does the proposed amendment involve issue of no significant hazards
a significant reduction in a margin of safety?
Description of amendment request: consideration, which is presented
Response: No.
The proposed changes would revise the The WWS is a non-safety-related system below:
Combined License (COL) Appendix C that does not interface with any safety-related 1. Does the proposed amendment involve
and corresponding plant-specific Tier 1 equipment. The proposed changes to identify a significant increase in the probability or
information to add two turbine building that there is more than one turbine building consequences of an accident previously
sump pumps to accommodate the sump and to add two turbine building sump evaluated?
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![73440 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
Response: No. is affected by these changes. The changes do Report (UFSAR) in the form of
The in-containment refueling water storage not introduce a new failure mode, departures from the incorporated plant-
tank (IRWST) provides flooding of the malfunction or sequence of events that could specific Design Control Document Tier
refueling cavity for normal refueling. The affect plant safety or safety-related equipment
2* information. Specifically, the
tank also serves as a heat sink during Passive as the simplistic design of the cover louvers
Residual Heat Removal (PRHR) Heat and hinged flappers are not considered proposed changes would revise the
Exchanger (HX) operation and in the event of unique designs. No new credible failure Combined Licenses for the Vogtle
a loss-of-coolant-accident (LOCA) provides modes are introduced by the addition of the Electric Generating Plant, Units 3 and 4,
injection in support of long-term RCS [reactor covers. to clarify information in WCAP–17179,
coolant system] cooling. This activity adds The proposed changes to the description ‘‘AP1000® Component Interface Module
normally closed covers to the IRWST vents and equipment tag associated with the risk- Technical Report,’’ which demonstrates
and overflow weirs to prevent debris from significant control relays for the rod drive design compliance with licensing bases
entering the tank, prevent over-pressurization MG sets do not adversely affect any safety- requirements. WCAP–17179 is
and accommodate volume and mass related equipment, and do not add any new
increases in the tank. The vent and overflow incorporated by reference into the
interfaces to safety-related SSCs. No system
weir covers open upon differential pressures or design function or equipment qualification UFSAR to provide additional details
between the IRWST and containment. is affected by these changes. The changes do regarding the component interface
The rod drive MG sets provide the power not introduce a new failure mode, module (CIM) system design. The
to the control rod drive mechanisms through malfunction or sequence of events that could requested amendments also propose a
the reactor trip switchgear. This activity affect plant safety or safety-related equipment change to the CIM internal power
revises the equipment description and because the design function of the control supply that will enable proper
equipment tag associated with the risk- relays, control cabinets, or rod drive MG sets functioning of the field programmable
significant control relays which open to de- is not changed.
energize the rod drive MG sets and permit gate arrays.
Therefore, the proposed amendment does Basis for proposed no significant
rods to drop. not create the possibility of a new or different
The proposed changes to add the IRWST hazards consideration determination:
kind of accident from any accident
vent and overflow weir covers and to change previously evaluated. As required by 10 CFR 50.91(a), the
the description of the equipment and 3. Does the proposed amendment involve licensee has provided its analysis of the
equipment tag related to the rod drive MG a significant reduction in a margin of safety? issue of no significant hazards
sets does not inhibit the SSCs from Response: No. consideration, which is presented
performing their safety-related function. The The proposed changes maintain below:
design bases of the IRWST vents and compliance with the applicable Codes and
overflow weirs are not modified as a result 1. Does the proposed amendment involve
Standards, thereby maintaining the margin of a significant increase in the probability or
of the addition of the covers to the vents and safety associated with these SSCs. The
overflow weirs and the change to the control consequences of an accident previously
proposed changes do not alter any applicable evaluated?
cabinet relay description and equipment tag. design codes, code compliance, design
This proposed amendment does not have an Response: No.
function, or safety analysis. Consequently, no The proposed change to the CIM internal
adverse impact on the response to safety analysis or design basis acceptance power supply enables the field
anticipated transients or postulated accident limit/criterion is challenged or exceeded by programmable gate array (FPGA) to function
conditions because the functions of the SSCs the proposed change, thus the margin of properly. The proposed change to the FPGA
are not changed. Required IRWST venting is safety is not reduced. Because no safety core power has no adverse effect on the
not affected for any accident conditions. analysis or design basis acceptance limit/ operation of the output actuation relays. The
Required DAS functions are not affected for criterion is challenged or exceeded by these function of the internal power supply has no
any accident conditions. Safety-related changes, no margin of safety is reduced. input to plant safety analysis. The change to
structure, system, component (SSC) or Therefore, the proposed change does not the CIM internal power supply has a
function is not adversely affected by this involve a significant reduction in a margin of negligible effect on the 24 Vdc [volts direct
change. The changes to include the IRWST safety. current] supplies and ultimately the plant
covers and to change the control cabinet electrical system load and has no adverse
relay description and tag number do not The NRC staff has reviewed the
effect on the CIM functionality.
involve an interface with any SSC accident licensee’s analysis and, based on this The proposed changes to clarify how
initiator or initiating sequence of events, and review, it appears that the three licensing basis design documentation reflects
thus, the probabilities of the accidents standards of 10 CFR 50.92(c) are compliance with license basis requirements,
evaluated in the UFSAR are not affected. The satisfied. Therefore, the NRC staff and the proposed change to the ownership of
proposed changes do not involve a change to proposes to determine that the safety remote node controller (SRNC) and
the predicted radiological releases due to amendment request involves no CIM intellectual property, are not technical
postulated accident conditions, thus, the changes. The proposed changes do not affect
significant hazards consideration.
consequences of the accidents evaluated in any accident initiator in the UFSAR, or affect
Attorney for licensee: M. Stanford
the UFSAR are not affected. Probabilistic the radioactive material releases in the
Risk Assessment (PRA) modeling and Blanton, Balch & Bingham LLP, 1710
UFSAR accident analyses. The proposed
analyses associated with the SSCs are not Sixth Avenue North, Birmingham, AL change does not alter the ability of the
impacted by this change. 35203–2015. facility to prevent and mitigate abnormal
Therefore, the proposed amendment does NRC Branch Chief: Michael T. events, e.g., accidents, anticipated
not involve a significant increase in the Markley. operational occurrences, earthquakes, floods
probability or consequences of an accident and turbine missiles, or their safety or design
previously evaluated. Southern Nuclear Operating Company, analyses. No safety-related structure, system,
2. Does the proposed amendment create Inc., Docket Nos. 52–025 and 52–026, or component (SSC) or function is adversely
the possibility of a new or different kind of Vogtle Electric Generating Plant, Units 3 affected. The change does not involve or
accident from any accident previously and 4, Burke County, Georgia interface with any SSC accident initiator or
mstockstill on DSK3G9T082PROD with NOTICES
evaluated? Date of amendment request: August initiating sequence of events, and thus, the
Response: No. probabilities of the accidents evaluated in the
The proposed changes to the design of the
31, 2016. A publicly-available version is
UFSAR are not affected. This activity does
IRWST vent and overflow weir covers do not in ADAMS under Accession No. not involve a new fission product release
adversely affect any safety-related ML16244A836. path, nor a new fission product barrier failure
equipment, and do not add any new Description of amendment request: mode, nor create a new sequence of events
interfaces to safety-related SSCs. No system The amendments propose changes to that would result in significant fuel cladding
or design function or equipment qualification the Updated Final Safety Analysis failures. Because the proposed changes do
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![73442 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
loss-of-coolant accident (LOCA) and fuel/ Therefore, the proposed change does not consistent with previously evaluated
equipment handling accident. Each accident involve a significant reduction in a margin of accidents. On this basis, the proposed delay
requires time to drawdown the secondary safety. in performance of the SRs in this amendment
containment to less than atmospheric request does not involve a significant
pressure. The brief, inadvertent, The NRC staff has reviewed the increase in the consequences of an accident.
simultaneous opening of both an inner and licensee’s analysis and, based on this Therefore, the proposed change does not
outer personnel access door during normal review, it appears that the three involve a significant increase in the
entry and exit conditions followed by prompt standards of 10 CFR 50.92(c) are probability or consequences of an accident
closure does not challenge the design basis satisfied. Therefore, the NRC staff previously evaluated.
drawdown time and does not result in an proposes to determine that the 2. Does the proposed amendment create
increase in any on-site or offsite dose for the amendment request involves no the possibility of a new or different kind of
LOCA dose analysis. All dose consequences accident from any accident previously
significant hazards consideration. evaluated?
are within the regulatory limits established
for the fuel handling accident and bound the
Attorney for licensee: Damon D. Obie, Response: No.
case in which airlock doors are briefly, Esquire, Associate General Counsel, The proposed amendment does not involve
inadvertently opened.] As a result, the Talen Energy Supply, LLC, 835 a physical alteration of any system, structure,
consequences of any accident previously Hamilton St., Suite 150, Allentown, PA or component (SSC) or a change in the way
evaluated is not significantly increased. 18101. any SSC is operated. The proposed
Therefore, the proposed change does not NRC Branch Chief: Douglas A. amendment does not involve operation of
involve a significant increase in the Broaddus. any SSCs in a manner or configuration
probability or consequences of an accident different from those previously recognized or
previously evaluated. Tennessee Valley Authority, Docket No. evaluated. No new failure mechanisms will
2. Do the proposed changes create the 50–391, Watts Bar Nuclear Plant (WBN), be introduced by the one-time SR extensions
possibility of a new or different kind of Unit 2, Rhea County, Tennessee being requested.
accident from any accident previously Therefore, the proposed change does not
evaluated? Date of amendment request: create the possibility of a new or different
Response: No. September 30, 2016. A publicly- kind of accident from any accident
The proposed change does not involve a available version is in ADAMS under previously evaluated.
physical alteration of any plant equipment. Accession No. ML16277A477. 3. Does the proposed amendment involve
a significant reduction in a margin of safety?
No new equipment is being introduced, and Description of amendment request:
installed equipment is not being operated in Response: No.
The amendment would revise the The proposed amendment is a one-time
a new or different manner. There are not Technical Specifications (TSs) to allow extension of the performance interval of TS
setpoints, at which protective or mitigative a one-time extension of the frequency SRs [3.6.11.2] and 3.6.11.3. Extending these
actions are initiated, affected by the proposed
for performing TS Surveillance surveillance requirements does not involve a
change. The proposed change does not alter
Requirements (SRs) related to verifying modification of any TS limiting conditions
the manner in which equipment operation is for operation. Extending these SRs does not
initiated, nor will the function of credited the operability of the containment ice
bed. involve a change to any limit on accident
equipment be changed. No alterations in the consequences specified in the license or
procedures that ensure the plant remains Basis for proposed no significant regulations. Extending these SRs does not
within analyzed limits are being proposed, hazards consideration determination: involve a change in how accidents are
and no changes are being made to the As required by 10 CFR 50.91(a), the mitigated or a significant increase in the
procedures relied upon to respond to an off- licensee has provided its analysis of the consequences of an accident. Extending these
normal event described in the FSAR [Final issue of no significant hazards SRs does not involve a change in a
Safety Analysis Report]. As such, no new methodology used to evaluate consequences
consideration, which is presented
failure modes are being introduced. The of an accident. Extending these SRs does not
change does not alter the assumptions made below, with NRC edits in brackets:
involve a change in any operating procedure
in the safety analysis and licensing basis. 1. Does the proposed amendment involve or process.
Therefore, the proposed change does not a significant increase in the probability or Based on the limited additional period of
create the possibility of a new or different consequences of an accident previously time that the systems and components will
kind of accident from any previously evaluated? be in service before the surveillances are next
evaluated. Response: No. performed, as well as the operating
3. Do the proposed changes involve a The requested action is a one-time experience that these surveillances are
significant reduction in a margin of safety? extension to the performance interval for TS typically successful when performed, it is
Response: No. SRs [3.6.11.2] and 3.6.11.3. The performance reasonable to conclude that the margins of
The margin of safety is established through of these surveillances, or the extension of safety associated with these SRs will not be
equipment design, operating parameters, and these surveillances, is not a precursor to an affected by the requested extension.
the setpoints at which automatic actions are accident. Performing these surveillances or Therefore, the proposed change does not
initiated. The proposed change addresses failing to perform these surveillances does involve a significant reduction in a margin of
temporary conditions during which the not affect the probability of an accident. safety.
Secondary Containment SR is not met. The Therefore, the proposed delay in
allowance for both an inner and outer performance of the SRs in this amendment The NRC staff has reviewed the
Secondary Containment access door to be request does not increase the probability of licensee’s analysis and, based on this
open simultaneously for entry and exit does an accident previously evaluated. review, it appears that the three
not affect the safety function of the reactor A delay in performing these surveillances standards of 10 CFR 50.92(c) are
enclosure and refuel area Secondary does not result in a system being unable to satisfied. Therefore, the NRC staff
Containments as the doors are promptly perform its required function. In the case of proposes to determine that the
closed after entry of exit, thereby restoring this one-time extension request, the short amendment request involves no
the Secondary Containment boundary. In period of additional time that the systems
mstockstill on DSK3G9T082PROD with NOTICES
significant hazards consideration.
addition, brief, inadvertent simultaneous and components will be in service before the Attorney for licensee: Ms. Sherry A.
opening and closing of redundant Secondary next performance of the surveillance will not
Containment personnel access doors during affect the ability of those systems to operate
Quirk, Executive Vice President and
normal entry and exit conditions does not as designed. Therefore, the systems required General Counsel, Tennessee Valley
affect the ability of the SGTS to establish the to mitigate accidents will remain capable of Authority, 400 West Summit Hill Drive,
required Secondary Containment vacuum. performing their required function. No new 6A West Tower, Knoxville, TN 37902.
Therefore, the safety function of the failure modes have been introduced because NRC Acting Branch Chief: Jeanne A.
Secondary Containment is not affected. of this action and the consequences remain Dion.
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73443
Virginia Electric and Power Company, significant increase in the probability or Resources Services, Inc., 120 Tredegar
Docket Nos. 50–280 and 50–281, Surry consequences of an accident previously St., RS–2, Richmond, VA 23219.
Power Station, Unit Nos. 1 and 2, Surry evaluated. NRC Branch Chief: Michael T.
County, Virginia 2. Does the proposed change create the Markley.
possibility of a new or different kind of
Date of amendment request: May 18, accident from any accident previously Virginia Electric and Power Company,
2016. A publicly-available version is in evaluated? Docket Nos. 50–280 and 50–281, Surry
ADAMS under Accession No. Response: No. Power Station, Unit Nos. 1 and 2, Surry
ML16146A540. The proposed change extends the AOT for County, Virginia
Description of amendment request: only one operable CPSW or MCR/ESGRAC
flow path from 24 hours to 72 hours. In Date of amendment request: July 14,
The amendments would revise the
addition, the proposed change deletes the 2016. A publicly-available version is in
Surry Power Station, Unit Nos. 1 and 2, now expired and no longer necessary ADAMS under Accession No.
Technical Specification (TS) 3.14, requirements for the temporary SW jumper to ML16202A068.
‘‘Circulating and Service Water the CCHXs. The proposed change does not Description of amendment request:
Systems,’’ to extend the allowed outage involve a physical alteration of the plant (i.e., The amendments would revise the
time (AOT) for only one operable no new or different type of equipment will
Surry Power Station, Unit Nos. 1 and 2,
service water (SW) flow path to the be installed) and does not impact plant
operation. Furthermore, the proposed change Technical Specification (TS) 3.14,
charging pump service water (CPSW)
does not impose any new or different ‘‘Circulating and Service Water
subsystem and to the main control
requirements that could initiate an accident. Systems,’’ to extend the allowed outage
room/emergency switchgear room
The proposed change does not alter time (AOT) for emergency service water
(MCR/ESGR) air conditioning (AC) assumptions made in the safety analysis and (ESW) pump inoperability. The
subsystem. TS 3.14.A.5 and TS 3.14.A.7 is consistent with the safety analysis proposed revision would extend the TS
require two SW flow paths to the CPSW assumptions. 3.14.B AOT for one inoperable ESW
subsystem and to the MCR/ESGR AC Therefore, the proposed change does not pump from 7 days to 14 days to provide
subsystem, respectively, to be operable. create the possibility of a new or different
operational flexibility for ESW pump
Currently, the TS 3.14.C AOT for only kind of accident from any accident
previously evaluated. maintenance and repairs.
one operable CPSW or MCR/ESGR AC Basis for proposed no significant
flow path is 24 hours. The proposed 3. Does the proposed change involve a
significant reduction in a margin of safety? hazards consideration determination:
revision will extend the AOT for only As required by 10 CFR 50.91(a), the
Response: No.
one operable CPSW or MCR/ESGR AC The proposed change extends the AOT for licensee has provided its analysis of the
flow path from 24 hours to 72 hours. only one operable CPSW or MCR/ESGR AC issue of no significant hazards
Basis for proposed no significant flow path from 24 hours to 72 hours. The consideration, which is presented
hazards consideration determination: proposed change does not adversely affect below:
As required by 10 CFR 50.91(a), the any current plant safety margins or the
licensee has provided its analysis of the reliability of the equipment assumed in the 1. Does the proposed change involve a
issue of no significant hazards safety analysis. There are no changes being significant increase in the probability or
consideration, which is presented made to any safety analysis assumptions, consequences of an accident previously
safety limits, or limiting safety system evaluated?
below: Response: No.
settings that would adversely affect plant
1. Does the proposed change involve a safety as a result of the proposed change. The design function of the ESW pumps is
significant increase in the probability or Furthermore, as noted above, a supporting to ensure that water can be provided to the
consequences of an accident previously PRA [probabilistic risk assessment] was intake canal (i.e., the ultimate heat sink)
evaluated? performed for the proposed AOT changes. when power is not available to the
Response: No. The PRA concluded that the increase in risk Circulating Water (CW) pumps. The
The proposed change extends the AOT for associated with the proposed changes is proposed extension of the AOT for one
only one operable CPSW or MCR/ESGR AC consistent with the RG [Regulatory Guide] inoperable ESW pump from 7 to 14 days does
flow path from 24 hours to 72 hours. The 1.174 and RG 1.177 acceptance guidelines for not impact the design function of the ESW
CPSW subsystem is a support system for the a permanent TS AOT change. This PRA pumps. In addition, the number of ESW
Charging/High Head Safety Injection (HHSI) evaluation demonstrates that defense-in- pumps required to be operable for the
pumps; the proposed CPSW AOT extension depth will not be significantly impacted by specified plant operating conditions is not
aligns the CPSW support system AOT with changing the AOTs for only one operable SW impacted by the proposed AOT extension. As
the AOT for the supported components (i.e., flow path to the CPSW subsystem and to the a result, the proposed change does not
the Charging/HHSI pumps). The proposed MCR/ESGR AC subsystem from 24 to 72 involve a significant increase in the
MCR/ESGR AC AOT extension revises the hours. In addition, the proposed change probability or consequences of an accident
AOT to be the same as the CPSW AOT since deletes the now expired and no longer previously evaluated.
both subsystems share common piping. The necessary requirements for the temporary SW 2. Does the proposed change create the
design function of the CPSW system, which jumper to the CCHXs. The deletion of these possibility of a new or different kind of
is to provide cooling to the charging pump temporary requirements is administrative in accident from any accident previously
intermediate seal coolers and the charging nature. Therefore, the proposed change does evaluated?
pump lubricating oil coolers, is not impacted not involve a significant reduction in a Response: No.
by the proposed revision, nor is the design margin of safety. The proposed change does not involve a
function of the Charging/HHSI pumps physical alteration of the plant (i.e., no new
impacted. Furthermore, the design functions The NRC staff has reviewed the or different type of equipment will be
of the MCR/ESGR AC subsystem and the licensee’s analysis and, based on this installed) and does not impact plant
MCR/ESGR chillers are not impacted by the review, it appears that the three operation. Furthermore, the proposed change
mstockstill on DSK3G9T082PROD with NOTICES
proposed revision. In addition, the proposed standards of 10 CFR 50.92(c) are does not impose any new or different
change deletes the now expired and no satisfied. Therefore, the NRC staff requirements that could initiate an accident.
longer necessary requirements for the proposes to determine that the The proposed change does not alter
temporary SW jumper to the CCHXs assumptions made in the safety analysis and
[component cooling heat exchangers]. The
amendment request involves no is consistent with the safety analysis
deletion of these temporary requirements is significant hazards consideration. assumptions.
administrative in nature. As a result, the Attorney for licensee: Lillian M. Therefore, the proposed change does not
proposed change does not involve a Cuoco, Senior Counsel, Dominion create the possibility of a new or different
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![73444 Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices
kind of accident from any accident amendments satisfy the criteria for The Commission’s related evaluation
previously evaluated. categorical exclusion in accordance of the amendment is contained in a
3. Does the proposed change involve a with 10 CFR 51.22. Therefore, pursuant Safety Evaluation dated September 30,
significant reduction in a margin of safety?
to 10 CFR 51.22(b), no environmental 2016.
Response: No.
The proposed change does not adversely impact statement or environmental No significant hazards consideration
affect any current plant safety margins or the assessment need be prepared for these comments received: No.
reliability of the equipment assumed in the amendments. If the Commission has
safety analysis. There are no changes being prepared an environmental assessment Entergy Operations, Inc., Docket No. 50–
made to any safety analysis assumptions, under the special circumstances 313, Arkansas Nuclear One, Unit 1
safety limits, or limiting safety system provision in 10 CFR 51.22(b) and has (ANO–1), Pope County, Arkansas
settings that would adversely affect plant made a determination based on that
safety as a result of the proposed change. Date of amendment request: January
Furthermore, as noted above, a supporting assessment, it is so indicated.
For further details with respect to the 29, 2014, as supplemented by letters
PRA [probabilistic risk assessment] was
performed for the proposed AOT change. The action see (1) the applications for dated May 19, June 16, July 21, August
PRA concluded that the increase in risk amendment, (2) the amendment, and (3) 12, September 22, November 4, and
associated with the proposed change is the Commission’s related letter, Safety November 17, 2015; and January 15,
consistent with the RG [Regulatory Guide] Evaluation, and/or Environmental March 25, April 7, May 19, and August
1.174 and RG 1.177 acceptance guidelines for Assessment as indicated. All of these 29, 2016.
a permanent TS AOT change. This PRA items can be accessed as described in
evaluation demonstrates that defense-in- Brief description of amendment: The
depth will not be significantly impacted by the ‘‘Obtaining Information and amendment authorized the transition of
changing the AOT for one inoperable ESW Submitting Comments’’ section of this the ANO–1 fire protection program to a
pump from 7 to 14 days. document. risk-informed, performance-based
Therefore, the proposed change does not program based on National Fire
involve a significant reduction in a margin of
Dominion Nuclear Connecticut, Inc.,
safety. Docket No. 50–336, Millstone Power Protection Association (NFPA) 805, in
Station, Unit No. 2 (MPS2), New London accordance with 10 CFR 50.48(c). NFPA
The NRC staff has reviewed the County, Connecticut 805 allows the use of performance-based
licensee’s analysis and, based on this methods such as fire modeling and risk-
review, it appears that the three Date of amendment request:
September 1, 2015, as supplemented by informed methods such as fire
standards of 10 CFR 50.92(c) are probabilistic risk assessment to
satisfied. Therefore, the NRC staff letter dated March 24, 2016.
Brief description of amendment: The demonstrate compliance with the
proposes to determine that the
amendment revised the MPS2 Technical nuclear safety performance criteria.
amendment request involves no
significant hazards consideration. Specifications (TSs) to add the Date of issuance: October 7, 2016.
Attorney for licensee: Lillian M. evaluation model EMF–2328(P)(A), Effective date: As of the date of
Cuoco, Senior Counsel, Dominion Supplement 1, ‘‘PWR [pressurized water
issuance and shall be implemented as
Resources Services, Inc., 120 Tredegar reactor] Small Break LOCA [loss-of
described in the transition license
St., RS–2, Richmond, VA 23219. coolant accident] Evaluation Model S–
conditions.
NRC Branch Chief: Michael T. RELAP5 Based,’’ and EMF–92–
Markley. 116(P)(A), Supplement 1, ‘‘Generic Amendment No.: 256. A publicly-
Mechanical Design Criteria for PWR available version is in ADAMS under
III. Notice of Issuance of Amendments Fuel Designs,’’ to the TS Section Accession No. ML16223A481;
to Facility Operating Licenses and 6.9.1.8.b list of analytical methods used documents related to this amendment
Combined Licenses to determine core operating limits as a are listed in the Safety Evaluation
During the period since publication of result of reanalyzing the small break enclosed with the amendment.
the last biweekly notice, the LOCA. Renewed Facility Operating License
Commission has issued the following Date of issuance: September 30, 2016. No. DPR–51: The amendment revised
amendments. The Commission has Effective date: As of the date of the Facility Operating License and
determined for each of these issuance and shall be implemented
Technical Specifications.
amendments that the application within 60 days of issuance.
complies with the standards and Amendment No.: 329. A publicly- Date of initial notice in Federal
requirements of the Atomic Energy Act available version is in ADAMS under Register: July 8, 2014 (79 FR 38589).
of 1954, as amended (the Act), and the Accession No. ML16249A001; The supplemental letters dated May 19,
Commission’s rules and regulations. documents related to this amendment June 16, July 21, August 12, September
The Commission has made appropriate are listed in the Safety Evaluation 22, November 4, and November 17,
findings as required by the Act and the enclosed with the amendment. 2015; and January 15, March 25, April
Commission’s rules and regulations in Renewed Facility Operating License 7, May 19, and August 29, 2016,
10 CFR Chapter I, which are set forth in No. DPR–65: Amendment revised the provided additional information that
the license amendment. Renewed Operating License and TSs. clarified the application, did not expand
A notice of consideration of issuance Date of initial notice in Federal the scope of the application as originally
of amendment to facility operating Register: December 8, 2015 (80 FR noticed, and did not change the staff’s
license or combined license, as 76318). The supplemental letter dated original proposed no significant hazards
applicable, proposed no significant March 24, 2016, provided additional
mstockstill on DSK3G9T082PROD with NOTICES
consideration determination as
hazards consideration determination, information that clarified the published in the Federal Register.
and opportunity for a hearing in application, did not expand the scope of
the application as originally noticed, The Commission’s related evaluation
connection with these actions, was
and did not change the staff’s original of the amendment is contained in a
published in the Federal Register as
indicated. proposed no significant hazards Safety Evaluation dated October 7, 2016.
Unless otherwise indicated, the consideration determination as No significant hazards consideration
Commission has determined that these published in the Federal Register. comments received: No.
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![Federal Register / Vol. 81, No. 206 / Tuesday, October 25, 2016 / Notices 73447
No significant hazards consideration Detailed procedures for the conduct of information, analyze relevant issues and
comments received: No. and participation in ACRS meetings facts, and formulate proposed positions
Dated at Rockville, Maryland, this 13th day were published in the Federal Register and actions, as appropriate, for
of October 2016. on October 21, 2015, (80 FR 63846). deliberation by the Full Committee.
For the Nuclear Regulatory Commission. Detailed meeting agendas and meeting Members of the public desiring to
George A. Wilson, Jr.,
transcripts are available on the NRC provide oral statements and/or written
Web site at http://www.nrc.gov/reading- comments should notify the Designated
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
rm/doc-collections/acrs. Information Federal Official (DFO), Quynh Nguyen
Regulation. regarding topics to be discussed, (Telephone 301–415–5844 or Email:
[FR Doc. 2016–25641 Filed 10–24–16; 8:45 am]
changes to the agenda, whether the Quynh.Nguyen@nrc.gov) five days prior
meeting has been canceled or to the meeting, if possible, so that
BILLING CODE 7590–01–P
rescheduled, and the time allotted to arrangements can be made. Thirty-five
present oral statements can be obtained hard copies of each presentation or
NUCLEAR REGULATORY from the Web site cited above or by handout should be provided to the DFO
COMMISSION contacting the identified DFO. thirty minutes before the meeting. In
Moreover, in view of the possibility that addition, one electronic copy of each
Advisory Committee on Reactor the schedule for ACRS meetings may be presentation should be emailed to the
Safeguards (ACRS); Meeting of the adjusted by the Chairman as necessary DFO one day before the meeting. If an
ACRS Subcommittee on Digital I&C to facilitate the conduct of the meeting, electronic copy cannot be provided
Systems; Notice of Meeting persons planning to attend should check within this timeframe, presenters
with these references if such should provide the DFO with a CD
The ACRS Subcommittee on Digital rescheduling would result in a major
I&C Systems will hold a meeting on containing each presentation at least
inconvenience. thirty minutes before the meeting.
November 2, 2016, Room T–2B1, 11545 If attending this meeting, please enter
Rockville Pike, Rockville, Maryland. Electronic recordings will be permitted
through the One White Flint North only during those portions of the
The meeting will be open to public building, 11555 Rockville Pike,
attendance with the exception of meeting that are open to the public.
Rockville, MD. After registering with Detailed procedures for the conduct of
portions that may be closed to protect security, please contact Mr. Theron
information that is proprietary pursuant and participation in ACRS meetings
Brown (Telephone 240–888–9835) to be were published in the Federal Register
to 5 U.S.C. 552b(c)(4). The agenda for escorted to the meeting room.
the subject meeting shall be as follows: on October 21, 2015 (80 FR 63846).
Dated: October 18, 2016. Information regarding changes to the
Wednesday, November 2, 2016—8:30 Mark L. Banks, agenda, whether the meeting has been
a.m. Until 5:00 p.m. Chief, Technical Support Branch, Advisory canceled or rescheduled, and the time
The Subcommittee will review the Committee on Reactor Safeguards. allotted to present oral statements can
Proposed Rulemaking on Cyber Security [FR Doc. 2016–25767 Filed 10–24–16; 8:45 am] be obtained by contacting the identified
for Fuel Cycle Facilities. The BILLING CODE 7590–01–P DFO. Moreover, in view of the
Subcommittee will hear presentations possibility that the schedule for ACRS
by and hold discussions with the NRC meetings may be adjusted by the
staff and other interested persons NUCLEAR REGULATORY Chairman as necessary to facilitate the
regarding this matter. The COMMISSION conduct of the meeting, persons
Subcommittee will gather information, planning to attend should check with
analyze relevant issues and facts, and Advisory Committee on Reactor the DFO if such rescheduling would
formulate proposed positions and Safeguards (ACRS); Meeting of the result in a major inconvenience.
actions, as appropriate, for deliberation ACRS Subcommittee on Planning and
Procedures; Notice of Meeting If attending this meeting, please enter
by the Full Committee. through the One White Flint North
Members of the public desiring to The ACRS Subcommittee on Planning building, 11555 Rockville Pike,
provide oral statements and/or written and Procedures will hold a meeting on Rockville, MD. After registering with
comments should notify the Designated November 3, 2016, Room T–2B3, 11545 security, please contact Mr. Theron
Federal Official (DFO), Christina Rockville Pike, Rockville, Maryland. Brown (240–888–9835) to be escorted to
Antonescu (Telephone 301–415–6792 or The meeting will be open to public the meeting room.
Email: Christina.Antonescu@nrc.gov) attendance with the exception of a
five days prior to the meeting, if Dated: October 18, 2016.
portion that may be closed pursuant to
possible, so that appropriate 5 U.S.C. 552b(c)(2) and (6) to discuss Mark L. Banks,
arrangements can be made. Thirty-five organizational and personnel matters Chief, Technical Support Branch, Advisory
hard copies of each presentation or that relate solely to the internal Committee on Reactor Safeguards.
handout should be provided to the DFO personnel rules and practices of the [FR Doc. 2016–25796 Filed 10–24–16; 8:45 am]
thirty minutes before the meeting. In ACRS, and information the release of BILLING CODE 7590–01–P
addition, one electronic copy of each which would constitute a clearly
presentation should be emailed to the unwarranted invasion of personal
DFO one day before the meeting. If an privacy. NUCLEAR REGULATORY
electronic copy cannot be provided COMMISSION
mstockstill on DSK3G9T082PROD with NOTICES
The agenda for the subject meeting
within this timeframe, presenters shall be as follows:
should provide the DFO with a CD [NRC–2016–0001]
containing each presentation at least Thursday, November 3, 2016—12:00
thirty minutes before the meeting. p.m. Until 1:00 p.m. Sunshine Act Meeting Notice
Electronic recordings will be permitted The Subcommittee will discuss
only during those portions of the proposed ACRS activities and related DATES: October 24, 31, November 7, 14,
meeting that are open to the public. matters. The Subcommittee will gather 21, 28, 2016.
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Document Created: 2018-02-13 16:38:42
Document Modified: 2018-02-13 16:38:42
| 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 |
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