82 FR 31089 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 82, Issue 127 (July 5, 2017)
Federal Register Volume 82, Issue 127 (July 5, 2017)
| Page Range | 31089-31106 | |
| FR Document | 2017-13804 |
[Federal Register Volume 82, Number 127 (Wednesday, July 5, 2017)] [Notices] [Pages 31089-31106] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2017-13804] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2017-0152] 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 June 3, 2017 to June 19, 2017. The last biweekly notice was published on June 19, 2017. DATES: Comments must be filed by August 4, 2017. A request for a hearing must be filed by September 5, 2017. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0152. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1506, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2017-0152, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods: Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0152. 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 [[Page 31090]] White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2017-0152, facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Sec. 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued. As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest. In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures. Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place [[Page 31091]] after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [email protected], or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC- issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system. A person filing electronically using the NRC's adjudicatory E- Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the ``Contact Us'' link located on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to [email protected], or by a toll-free call at 1-866-672- 7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays. Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd, unless excluded [[Page 31092]] pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of amendment request: April 3, 2017, as supplemented by letters dated April 3, 2017, and May 2, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17093A787, ML17093A796, and ML17122A223, respectively. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to extend the required frequency of certain 18-month Surveillance Requirements (SRs) to 24 months to accommodate a 24-month refueling cycle. In addition, the proposed amendment would revise certain programs in TS Section 5.5, ``Programs and Manuals,'' to change 18-month frequencies to 24 months. 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 changes the surveillance frequency from 18 months to 24 months for SRs in the TSs that are normally a function of the refueling interval. Duke Energy Progress, LLC's evaluations have shown that the reliability of protective instrumentation and equipment will be preserved for the maximum allowable surveillance interval. The proposed change does not involve any change to the design or functional requirements of the associated systems. That is, the proposed TS change neither degrades the performance of, nor increases the challenges to any safety systems assumed to function in the plant safety analysis. The proposed change will not give rise to any increase in operation power level, fuel operating limits or effluents. The proposed change does not affect any accident precursors since no accidents previously evaluated relate to the frequency of surveillance testing and the revision to the frequency does not introduce any accident initiators. The proposed change does not impact the usefulness of the SRs in evaluating the operability of required systems and components or the manner in which the surveillances are performed. In addition, evaluation of the proposed TS change demonstrates that the availability of equipment and systems required to prevent or mitigate the radiological consequences of an accident is not significantly affected because of the availability of redundant systems and equipment or the high reliability of the equipment. Since the impact on the systems is minimal, it is concluded that the overall impact on the plant safety analysis is negligible. Furthermore, an historical review of surveillance test results and associated maintenance records indicates there is no evidence of any failure that would invalidate the above conclusions. Therefore, the proposed TS change does not significantly increase 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 require a change to the plant design nor the mode of plant operation. No new or different equipment is being installed. No installed equipment is being operated in a different manner. As a result, no new failure modes are being introduced. In addition, the proposed change does not impact the usefulness of the SRs in evaluating the operability of required systems and components or the manner in which the surveillances are performed. Furthermore, an historical review of surveillance test results and associated maintenance records indicates there is no evidence of any failure that would invalidate the above conclusions. Therefore, the implementation of the proposed change will not create the possibility for an accident of a new or different type than previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment changes the surveillance frequency from 18 months to 24 months for SRs in the TSs that are normally a function of the refueling interval. SR 3.0.2 would allow a maximum surveillance interval of 30 months for these surveillances. Although the proposed change will result in an increase in the interval between surveillance tests, the impact on system availability is small based on other, more frequent testing that is performed, the existence of redundant systems and equipment or overall system reliability. There is no evidence of any time-dependent failures that would impact the availability of the systems. The proposed change does not significantly impact the condition or performance of structures, systems and components relied upon for accident mitigation. This change does not alter the existing TS allowable values or analytical limits. The existing operating margin between plant conditions and actual plant setpoints is not significantly reduced due to these changes. The assumptions and results in any safety analyses are not significantly impacted. Therefore, the proposed change does not involve a significant reduction in margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202. NRC Branch Chief: Undine S. Shoop. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Date of amendment request: April 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17114A398. Description of amendment request: The amendment would revise Technical Specification requirements regarding steam generator tube inspections and reporting as described in Technical Specification Task Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,'' using the Consolidated Line Item Improvement Process for Arkansas Nuclear One, Unit No. 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 [[Page 31093]] 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 Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a[n] SGTR is not increased. The consequences of a[n] SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a[n] SGTR to exceed those assumptions. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component. Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a[n] SG is maintained by ensuring the integrity of its tubes. SG tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas Date of amendment request: April 24, 2017. A publicly-available version is in ADAMS under Accession No. ML17114A399. Description of amendment request: The amendment would revise Technical Specification requirements regarding steam generator tube inspections and reporting as described in Technical Specifications Task Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,'' using the Consolidated Line Item Improvement Process for Arkansas Nuclear One, Unit No. 2. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, 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 Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a[n] SGTR is not increased. The consequences of a[n] SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a[n] SGTR to exceed those assumptions. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component. Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a[n] SG is maintained by ensuring the integrity of its tubes. SG tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana Date of amendment request: March 28, 2017. A publicly-available version is [[Page 31094]] in ADAMS under Accession No. ML17087A551. Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.8.1.3, ``Diesel Fuel Oil,'' by relocating the current stored diesel fuel oil numerical volume requirements from the TS to the TS Bases. In addition, the proposed amendment would revise TS 3.8.1.1, ``A.C. [Alternating Current] Sources--Operating,'' and TS 3.8.1.2, ``A.C. Sources--Shutdown,'' to relocate the specific numerical value for feed tank fuel oil volume to the TS Bases and replace it with the feed tank time requirement. The proposed changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-501, Revision 1, ``Relocate Fuel Oil and Lube Oil Volume Values to Licensee Control.'' 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 revise [TS] 3.8.1.3 (Diesel Fuel Oil) by removing the current stored diesel fuel oil numerical volume requirements from the TS and replacing them with diesel operating time requirements. The specific volume of fuel oil equivalent to a 7 and 6 day supply is calculated using the NRC approved methodology described in Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems for Standby Diesel Generators'' and [American Nuclear Standards Institute (ANSI)] N195-1976, ``Fuel Oil Systems for Standby Diesel- Generators'' using the time dependent load method as approved in Waterford 3 License Amendment 157. Because the requirement to maintain a 7 day supply of diesel fuel oil is not changed and is consistent with the assumptions in the accident analyses, and the actions taken when the volume of fuel oil is less than a 6 day supply have not changed, neither the probability nor the consequences of any accident previously evaluated will be affected. The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2 diesel feed tank fuel oil numerical volume requirements and replaces them with the diesel one hour diesel generator operation requirement. The specific volume and time is not changed and is consistent with the existing plant design basis to support a diesel generator under accident load conditions. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis but ensures that the diesel generator operates as assumed in the accident analysis. The proposed change 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 changes revise [TS] 3.8.1.3 (Diesel Fuel Oil) by removing the current stored diesel fuel oil numerical volume requirements from the TS and replacing them with diesel operating time requirements. As the bases for the existing limits on diesel fuel oil are not changed, no change is made to the accident analysis assumptions and no margin of safety is reduced as part of this change. The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2 diesel feed tank fuel oil numerical volume requirements and replaces them with the diesel one hour diesel generator operation requirement. As the basis for the existing limits on diesel fuel oil are not changed, no change is made to the accident analysis assumptions and no margin of safety is reduced as part of 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 request involves no significant hazards consideration. Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, Washington, DC 20001. NRC Branch Chief: Robert J. Pascarelli. Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and Lancaster Counties, Pennsylvania Date of amendment request: May 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17139D357. Description of amendment request: The amendment would revise the Technical Specifications (TSs) to decrease the number of safety relief valves and safety valves required to be operable when operating at a power level less than or equal to 3358 megawatts thermal (MWt). This change would be in effect for the current PBAPS, Unit 2, Cycle 22 that is scheduled to end in October 2018. 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 would revise TS Section 3.4.3 to decrease the required number of Safety Relief Valves (SRVs) and Safety Valves (SVs) from a total of 13 to 12, under reduced reactor thermal power operation of 3358 MWt (approximately 85% of Current Licensed Thermal Power (CLTP)). A compensatory reduction in maximum allowed reactor power to 3358 MWt has been determined to conservatively offset the impact/effects of operation with an additional (up to 2) SRVs/SVs Out-of-Service. The Reactor Pressure Vessel (RPV) overpressure protection capability of the 12 operable SRVs and SVs is adequate at the lower power level to ensure the ASME [American Society of Mechanical Engineers] code allowable peak pressure limits are not exceeded. With the maximum thermal power limitation condition, the proposed change has no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design basis accidents, Anticipated Operational Occurrence (AOO) events and Special Events remains bounded by existing analyses. The proposed change does not require any new or unusual operator actions. The proposed change does not introduce any new failure modes that could result in a new or different accident. The SRVs and SVs are not being modified or operated differently and will continue to operate to meet the design basis requirements for RPV overpressure protection. The proposed change does not alter the manner in which the RPV overpressure protection system is operated and functions and thus, there is no significant impact on reactor operation. There is no change being made to safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. For PBAPS, the limiting overpressure AOO event is the main steam isolation valve closure with scram on high flux (MSIVF). The PBAPS ATWS [anticipated transients without scram] Special Event evaluation considered the limiting cases for RPV overpressure and is analyzed under two cases: (1) Main Steam Isolation Valve Closure (MSIVC) and (2) Pressure Regulator Failure Open (PRFO). These events were analyzed under the proposed conditions and it was confirmed that the existing analyses remain bounding for the condition of adding a [[Page 31095]] second SRV/SV Out-of-Service with a limited maximum operating power level of 3358 MWt. 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 would revise TS Section 3.4.3 to decrease the required number of SRVs and SVs from a total of 13 to 12, under reduced reactor thermal power operation of 3358 MWt (approximately 85% of CLTP). A compensatory reduction in maximum allowed reactor power to 3358 MWt has been determined to conservatively offset the impact/effects of operation with an additional (up to 2) SRVs/SVs Out-of-Service. The RPV overpressure protection capability of the 12 operable SRVs and SVs is adequate at the lower power level to ensure the ASME code allowable peak pressure limits are not exceeded. The SRVs and SVs are not being modified or operated differently and will continue to operate to meet the design basis requirements for RPV overpressure protection. The proposed change does not introduce any new failure modes that could result in a new or different accident. The proposed reactor thermal power restriction of 3358 MWt is within the existing normal operating domain and no new or special operating actions are necessary to operate at the intermediate power level. The proposed change does not alter the manner in which the RPV overpressure protection system is operated and functions and thus, there is no new failure mechanisms for the overpressure protection system. The plant response to the design basis accidents, AOO events and Special Events remains bounded by existing analyses. [These] events were analyzed under the proposed conditions and it was confirmed that the existing analyses remain bounding for the condition of adding a second SRV/SV Out-of-Service with a limited maximum operating power level of 3358 MWt. 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 margin of safety is established though the design of the plant structures, systems and components, the parameters within which the plant is operated, and the establishment of setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not change the setpoints at which the protective actions are initiated. The proposed change would revise TS Section 3.4.3 to decrease the required number of SRVs and SVs under reduced reactor thermal power operation of 3358 MWt (approximately 85% of CLTP). A compensatory reduction in maximum allowed reactor power to 3358 MWt has been determined to conservatively offset the impact/effects of operation with an additional (up to 2) SRVs/SVs Out-of-Service. The RPV overpressure protection capability of the 12 operable SRVs and SVs is adequate at the lower power level to ensure the ASME code allowable peak pressure limits are not exceeded. The plant response to the design basis accidents, AOO events and Special Events remains bounded by existing analyses. These events were analyzed under the proposed conditions and it was confirmed that the existing analyses remain bounding for the condition of adding a second SRV/SV Out-of-Service with a limited maximum operating power level of 3358 MWt. 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 Rd., Warrenville, IL 60555. NRC Branch Chief: James G. Danna. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No.1, DeWitt County, Illinois Date of amendment request: May 4, 2017. A publicly-available version is in ADAMS under Accession No. ML17124A121. Description of amendment request: The proposed change would delete a surveillance requirement (SR) Note associated with technical specification (TS) 3.5.1, ``ECCS [emergency core cooling system]-- Operating,'' TS 3.5.2, ``ECCS--Shutdown,'' and TS 3.6.1.7, ``Residual Heat Removal (RHR) Containment Spray System,'' to more appropriately reflect the RHR system design, and ensure the RHR system operation is consistent with the TS limiting condition for operation (LCO) requirements. In addition, the proposed amendment would insert a Note in the LCO for TSs 3.5.1, 3.5.2, 3.6.1.7, 3.6.1.9, ``Feedwater Leakage Control System,'' and 3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool Cooling,'' to clarify that one of the required subsystems in each of the affected TS sections may be inoperable during alignment and operation of the RHR system for shutdown cooling (SDC) with the reactor steam dome pressure less than the RHR cut in permissive value. 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. No physical changes to the facility will occur as a result of this proposed amendment. The proposed changes will not alter the physical design. The current TS (CTS) Note in SR 3.5.1.4, SR 3.5.2.4, and 3.6.1.7 could make CPS susceptible to potential water hammer in the RHR system while operating in the SDC mode of RHR in MODE 3 when swapping from the SDC to LPCI [low-pressure coolant injection] and RHR containment spray modes of RHR. Deletion of the Note from SR 3.5.1.2, SR 3.5.2.4, and SR 3.6.1.7.1 will eliminate the risk for cavitation of the pump and voiding in the suction piping, thereby avoiding the potential to damage the RHR system, including water hammer. The addition of proposed TS note to LCO 3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3 will re- establish consistency of the CPS RHR system design with the original TS requirements. 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 physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function. Deletion of the Note from SR 3.5.1.2, SR 3.5.2.4 and SR 3.6.1.7.1 is appropriate because current TSs could put the plant at risk for potential cavitation of the pump and voiding in the suction piping, resulting in potential to damage the RHR system, including water hammer. The addition of proposed TS note to LCO 3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3 will re-establish consistency of the CPS RHR system design with the original TS requirements. 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 conforms to NRC regulatory guidance regarding the content of plant Technical Specifications. The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. 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 [[Page 31096]] review it appears the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No.1, DeWitt County, Illinois Date of amendment request: May 1, 2017. A publicly-available version is in ADAMS under Accession No. ML17121A517. Description of amendment request: The proposed change replaces existing technical specification (TS) requirements related to operations with a potential for draining the reactor vessel (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel. 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 replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated. The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event. The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystem to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that secondary containment and/or filtration would be available if needed. The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review it appears the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: David J. Wrona. Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station (LGS), Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: April 24, 2017. A publicly available version is in ADAMS under Accession No. ML17115A087. Description of amendment request: The amendments would revise the LGS, Units 1 and 2, Technical Specifications (TSs) to a set of Improved Technical Specifications (ITS) based on NUREG-1433, Revision 4, ``Standard Technical Specifications--General Electric Plants, BWR/4,'' published April 2012. Specifically, the amendments would relocate TS Section 3.3.7.12, ``Offgas Gas Monitoring Instrumentation''; TS 3.11.2.5, ``Explosive Gas Mixture''; and Surveillance Requirement (SR) 4.11.2.6.1, which requires continuously monitoring the main condenser gaseous effluent to the LGS Offsite Dose Calculation Manual or to the LGS Technical Requirements Manual. In [[Page 31097]] addition, associated with the relocation of the main condenser offgas noble gas activity monitor, (1) SR 4.11.2.6.2.b will be changed to account for the relocated instrument's requirements, and (2) associated with the relocation of the explosive gas mixture instrumentation and gaseous effluent TS sections, a new TS Program Section, 6.8.4.l, ``Explosive Gas Monitoring Program,'' will be added to TS Section 6.8, ``Procedures and Programs.'' 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 relocate certain operability and surveillance requirements for the Main Condenser Offgas Monitoring Instrumentation and Gaseous Effluents limits from the Limerick Generating Station (LGS) Technical Specifications (TS) to a licensee-controlled document under the control of 10 CFR 50.59 or under the control of regulatory requirements applicable to the licensee-controlled document. A new TS Administrative Program is proposed to be added to ensure the limit for Main Condenser Offgas hydrogen concentration is maintained. The proposed changes do not alter the physical design of any plant structure, system, or component; therefore, the proposed changes have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design basis accidents does not change. Operation or failure of the Main Condenser Offgas Radioactivity and Hydrogen Monitors capability are not assumed to be an initiator of any analyzed event in the Updated Final Safety Analysis Report (UFSAR) and cannot cause an accident. Whether the requirements for the Main Condenser Offgas Radioactivity and Hydrogen Monitor capability are located in TS or another licensee-controlled document has no effect on the probability or consequences of any accident previously evaluated. The proposed changes conform to NRC regulatory requirements regarding the content of plant TS as identified in 10 CFR 50.36, and also the guidance as approved by the NRC in NUREG-1433, ``Standard Technical Specifications--General Electric BWR/4 Plants.'' Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes relocate certain operability and surveillance requirements for the Main Condenser Offgas Monitoring Instrumentation and Gaseous Effluents limits from the LGS TS to a licensee-controlled document under the control of 10 CFR 50.59 or under the control of regulatory requirements applicable to the licensee-controlled document. A new TS Administrative Program is proposed to be added to ensure the limit for Main Condenser Offgas hydrogen concentration is maintained. The proposed changes do not alter the plant configuration (no new or different type of equipment is being installed) or require any new or unusual operator actions. The proposed changes do not alter the safety limits or safety analysis assumptions associated with the operation of the plant. The proposed changes do not introduce any new failure modes that could result in a new accident. The proposed changes do not reduce or adversely affect the capabilities of any plant structure, system, or component in the performance of their safety function. Also, the response of the plant and the operators following the design basis accidents is unaffected by the proposed changes. 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 relocate certain operability and surveillance requirements for the Main Condenser Offgas Monitoring Instrumentation and Gaseous Effluents limits from the LGS TS to a licensee-controlled document under the control of 10 CFR 50.59 or under the control of regulatory requirements applicable to the licensee-controlled document. A new TS Administrative Program is proposed to be added to ensure the limit for the Main Condenser Offgas hydrogen concentration is maintained. The relocated TS requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on items for which a TS must be established. The proposed changes have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design basis accidents does not change. 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 is no change being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: James G. Danna. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant (PNPP), Unit No. 1, Lake County, Ohio Date of amendment request: April 26, 2017. A publicly-available version is in ADAMS under Accession No. ML17116A575. Description of amendment request: The proposed amendment would revise the PNPP Environmental Protection Plan (nonradiological) to clarify and enhance wording, to remove duplicative or outdated program information, and to relieve the burden of submitting unnecessary or duplicative information to the NRC. 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 involves changes to the Environmental Protection Plan (EPP), which provides for protection of nonradiological environmental values during operation of the nuclear facility. The proposed amendment does not change the objectives of the EPP, does not change the way the plant is maintained or operated, and does not affect any accident mitigating feature or increase the likelihood of malfunction for plant structures, systems and components. The proposed amendment will not change any of the analyses associated with the PNPP Updated Safety Analysis Report Chapter 15 accidents because plant operation, plant structures, systems, components, accident initiators, and accident mitigation functions remain unchanged. 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 amendment involves changes to the EPP, which provides for protection of nonradiological environmental values during operation of the nuclear facility. The proposed amendment does not involve a physical alteration of the plant. No new or different type of equipment will be installed, and there are no physical modifications to existing installed equipment associated with the proposed changes. The [[Page 31098]] proposed amendment does not change the way the plant is operated or maintained and does not create a credible failure mechanism, malfunction or accident initiator not already considered in the design and licensing basis. 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. Safety margins are applied to design and licensing basis functions and to the controlling values of parameters to account for various uncertainties and to avoid exceeding regulatory or licensing limits. The proposed amendment involves changes to the EPP, which provides for protection of nonradiological environmental values during operation of the nuclear facility. The proposed amendment does not involve a physical change to the plant, does not change methods of plant operation within prescribed limits, or affect design and licensing basis functions or controlling values of parameters for plant systems, structures, and components. 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: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: David J. Wrona. 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: May 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17144A294. Description of amendment request: The amendments would revise the St. Lucie Plant Unit Nos. 1 and 2 Renewed Facility Operating Licenses, Nos. DPR-67 and NPF-16, respectively, fire protection license conditions. The revisions would incorporate new references into these license conditions that propose and approve a revision to plant modifications previously approved in the March 31, 2016, NRC issuance of amendments regarding transition to a risk-informed, performance- based fire protection program in accordance with 10 CFR 50.48(c), dated March 21, 2016 (ADAMS Accession No. ML15344A346) (known as the National Fire Protection Association Standard 805 (NFPA 805)). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes are clarifications to methods applied to ensure compliance with NFPA 30, section 2348. The revised methods comply with NFPA 30, section 2348. This LAR [license amendment request] is essentially an administrative change to revise the letter referenced by the Fire Protection Transition License Conditions. The actual design changes and any related procedural changes are being managed separately from this LAR per 10 CFR 50.59. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not increase the probability or consequence of an accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are clarifications to methods applied to ensure compliance with NFPA 30, section 2348. The revised methods of compliance align with NFPA 30, section 2348, and will not result in new or different kinds of accidents. This LAR is essentially an administrative change to revise the letter referenced by the Fire Protection Transition License Conditions. The actual design changes and any related procedural changes are being managed separately from this LAR per 10 CFR 50.59. The requirements in NFPA 30 address only fire protection. The impacts of fire effects on the plant have been evaluated. The proposed amendment does not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the Unit 1 and Unit 2 UFSARs [updated final safety analysis reports]. Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Operation of Plant St. Lucie (PSL) in accordance with the proposed amendment does not involve a reduction in the margin of safety. The proposed amendment does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the UFSAR. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition remain capable of performing their design function. Therefore, this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William S. Blair, Managing Attorney-- Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/ JB, Juno Beach, FL 33408-0420. NRC Branch Chief: Undine S. Shoop. Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant (CNP), Units Nos. 1 and 2, Berrien County, Michigan Date of amendment request: May 23, 2017. A publicly-available version is in ADAMS under Accession No. ML17146A073. Description of amendment request: The proposed changes update the emergency action levels (EALs) used at CNP, Unit Nos. 1 and 2, from the current scheme based on Nuclear Management and Resources Council (NUMARC) and National Environmental Studies Project (NESP) NUMARC/NESP- 007, ``Methodology for Development of Emergency Action Levels'' dated January 1992, to a scheme based on Nuclear Energy Institute 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors.'' 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. [[Page 31099]] The proposed changes to the CNP EALs do not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SSCs perform their design function. EALs are used as criteria for determining the need for notification and participation of local and State agencies, and for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to the CNP EALs do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new equipment. EALs are based on plant conditions, so the proposed changes will not alter the design configuration or the method of plant operation. The proposed changes will not introduce failure modes that could result in a new or different type of accident, and the change does not alter assumptions made in the safety analysis. The proposed changes to the CNP Emergency Plan are not initiators of any accidents. 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. Margin of safety is associated with 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 to the CNP EALs do not impact operation of the plant or its response to transient or accidents. The changes do not affect the Technical Specifications or the operating license. 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. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. The emergency plan will continue to activate an emergency response commensurate with the extent of degradation of plant safety. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed changes involve references to available plant indications to assess conditions for determination of entry into an emergency action level. There is no change to these established safety margins as a result of 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 request involves no significant hazards consideration. Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: David J. Wrona. South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: May 11, 2017. A publicly-available version is in ADAMS under Accession No. ML17135A225. Description of amendment request: The requested amendment proposes to depart from combined license (COL) Appendix C information (with corresponding changes to the associated plant-specific Tier 1 information) and involves associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR). Specifically, proposed changes clarify that there is more than one turbine building main sump and adds a second sump pump for each of the two turbine building main sumps into UFSAR Tier 2 and COL Appendix C (and associated plant-specific Tier 1) information. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 departures. 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 activity adds a second pump to each of the turbine building main sumps, and identifies that there is more than one turbine building sump. The reason for the additional pumps is to account for an increase in volume due to the changes to the [condensate polishing system (CPS)] rinse effluent flowpath from [component cooling water system (CCW)] CCW to [waste water system (WWS)] WWS via the Turbine Building sumps. The extra sump pumps will prevent potential overflowing and flooding of the sumps during CPS rinse operations. The CPS serves no safety-related function. By directing the effluent to the turbine building sumps it is subject to radiation monitoring. Under normal operating conditions, there are no significant amounts of radioactive contamination within the CPS. However, radioactive contamination of the CPS can occur as a result of a primary to secondary leakage in the steam generator should a steam generator tube leak develop while the CPS is in operation and radioactive condensate is processed by the CPS. Radiation monitors associated with the steam generator blowdown, steam generator, and turbine island vents, drains and relief systems provide the means to determine if the secondary side is radioactively contaminated. The main turbine building sumps and sump pumps are not safety-related components and do not interface with any systems, structures, or components (SSC) accident initiator or initiating sequence of events; thus, the probability of accidents evaluated within the plant-specific 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 the non-safety waste water system (WWS) do not affect any safety-related equipment, nor does it add any new interface to safety-related SSCs. No system or design function or equipment qualification is affected by this change. 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 nonsafety-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, [[Page 31100]] 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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: May 16, 2017. A publicly-available version is in ADAMS under Accession No. ML17137A107. Description of amendment request: The requested amendment consist of changes to inspections, tests, analyses, and acceptance criteria (ITAAC) in combined license (COL) Appendix C, with corresponding changes to the associated plant-specific Tier 1 information, to consolidate a number of ITAAC to improve efficiency of the ITAAC completion and closure process. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 departures. 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 non-technical change to COL Appendix C will consolidate, relocate and subsume redundant ITAAC in order to improve and create a more efficient process for the ITAAC Closure Notification submittals. No structure, system, or component (SSC) design or function is affected. No design or safety analysis is affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses 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 change to COL Appendix C does not affect the design or function of any SSC, but will consolidate, relocate and subsume redundant ITAAC in order to improve efficiency of the ITAAC completion and closure process. The proposed changes would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release. 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 COL Appendix C to consolidate, relocate and subsume redundant ITAAC in order to improve efficiency of the ITAAC completion and closure process is considered non-technical and would not affect any design parameter, function or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit/criterion is involved. 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 Avenue NW., Washington, DC 20004-2514. NRC Branch Chief: Jennifer Dixon-Herrity. Southern California Edison Company, et al., Docket Nos. 50-361 and 50- 362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California Date of amendment request: May 16, 2017. A publicly-available version is in ADAMS under Accession No. ML17142A315. Description of amendment request: The proposed amendment would revise the Facility Operating Licenses for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, to reflect deletion of the Cyber Security Plan from License Condition 2.E. This will allow Southern California Edison (SCE) to terminate the SONGS Cyber Security Plan and associated activities at the site. These changes will more fully reflect the permanently shutdown and defueled status of the facility, as well as the reduced scope of potential radiological accidents and security concerns that exist during the decommissioning process. 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 to remove the San Onofre Nuclear Generating Station (SONGS) Cyber Security Plan requirement does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components (SSCs) relied upon to mitigate the consequences of postulated accidents, and has no impact on the probability or consequences of an accident previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to remove the SONGS Cyber Security Plan requirement does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the SSCs relied upon to mitigate the consequences of postulated accidents, and does not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Plant safety margins are established through limiting conditions for operation, [[Page 31101]] limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the SONGS Cyber Security Plan does not change these established safety margins. 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: Walker A. Matthews, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770. NRC Branch Chief: Bruce Watson, CHP. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: May 5, 2017. A publicly-available version is in ADAMS under Accession No. ML17125A331. Description of amendment request: The amendment request proposes to depart from plant-specific Tier 1 emergency planning inspection, test, analysis, and acceptance criteria (ITAAC) information and associated combined license (COL) Appendix C information. The proposed changes do not involve changes to the approved emergency plan or the plant- specific Tier 2 Design Control Document (DCD). Specifically, the requested amendment proposes to revise plant-specific emergency planning inspections (ITAAC) in Appendix C of the VEGP Units 3 and 4 COLs. Also, proposed changes to COL Appendix C information also include changes to the list of acronyms and abbreviations. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design 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: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The VEGP 3 and 4 emergency planning inspections, tests, analyses, and acceptance criteria (ITAAC) provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations. The proposed changes do not affect the design of a system, structure, or component (SSC) use to meet the design bases of the nuclear plant. Nor do the changes affect the construction or operation of the nuclear plant itself, so there is no change to the probability or consequences of an accident previously evaluated. Changing the VEGP 3 and 4 emergency planning ITAAC and COL, Appendix C, list of acronyms and abbreviations do not affect prevention and mitigation of abnormal events (e.g., accidents, anticipated operational occurrences, earthquakes, floods, or turbine missiles) or their safety or design analyses. No safety- related structure, system, component (SSC) or function is adversely affected. The changes neither involve nor interface with any SSC accident initiator or initiating sequence of events, so the probabilities of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate 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 VEGP 3 and 4 emergency planning ITAAC provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commissioner's rules and regulations. The changes do not affect the design of an SSC used to meet the design bases of the nuclear plant. Nor do the changes affect the construction or operation of the nuclear plant. Consequently, there is no new or different kind of accident from any accident previously evaluated. The changes do not affect safety- related equipment, nor do they affect equipment that, if it failed, could initiate an accident or a failure of a fission product barrier. In addition, the changes do not result in a new failure mode, malfunction, or sequence of events that could affect safety or safety-related equipment. No analysis is adversely affected. No system or design function or equipment qualification is adversely affected by the changes. This activity will not allow for a new fission product release path, nor will it result in a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 2. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The VEGP 3 and 4 emergency planning ITAAC provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commissioner's rules and regulations. The changes do not affect the assessments or the plant itself. The changes do not adversely affect the safety- related equipment or fission product barriers. No safety analysis or design basis acceptance limit or 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: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. 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: May 19, 2017. A publicly-available version is in ADAMS under Accession No. ML17139D394. Description of amendment request: The requested amendment proposes to depart from combined license (COL) Appendix C information (with corresponding changes to the associated plant-specific Tier 1 information) and involves associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR). Specifically, proposed changes clarify that there is more than one turbine building main sump and adds a second sump pump for each of the two turbine building main sumps into the UFSAR Tier 2 and COL Appendix C (and associated plant-specific Tier 1) information. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 departures. 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 [[Page 31102]] consequences of an accident previously evaluated? Response: No. The activity adds a second pump to each of the turbine building main sumps, and identifies that there is more than one turbine building sump. The reason for the additional pumps is to account for an increase in volume due to the changes to the condensate polishing system (CPS) rinse effluent flowpath from CPS to waste water system (WWS) via the turbine building sumps. The extra sump pumps will prevent potential overflowing and flooding of the sumps during CPS rinse operations. The CPS serves no safety-related function. By directing the effluent to the turbine building sumps it is subject to radiation monitoring. Under normal operating conditions, there are is no significant amount of radioactive contamination within the CPS. However, radioactive contamination of the CPS can occur as a result of a primary-to-secondary leakage in the steam generator should a steam generator tube leak develop while the CPS is in operation and radioactive condensate is processed by the CPS. Radiation monitors associated with the steam generator blowdown, steam generator, and turbine island vents, drains and relief systems provide the means to determine if the secondary side is radioactively contaminated. The main turbine building sumps and sump pumps are not safety-related components and do not interface with any systems, structures, or components (SSC) accident initiator or initiating sequence of events; thus, the probability of accidents evaluated within the plant-specific UFSAR are not affected. The proposed changes do not involve a change to the predicted radioactive 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 the nonsafety-related 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 this change. 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 nonsafety-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. 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: March 13, 2017. A publicly available version is in ADAMS under Accession No. ML17073A018. Description of amendment request: The amendments would modify the Surveillance Requirement (SR) 3.8.1.17 of the Technical Specification (TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' to delete the note to allow the performance of the SR in Modes 1 through 4 when the associated load is out of service for maintenance or 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: 1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated? Response: No. The proposal does not alter the function of any structure, system or component functions, does not modify the manner in which the plant is operated, and does not alter equipment out-of-service time. This request does not degrade the ability of the emergency diesel generator or equipment downstream of the load sequencers to perform their intended function. 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 involve any physical changes to plant safety related structure, system or component or alter the modes of plant operation in a manner that is outside the bounds of the current emergency diesel generator system design analyses. The proposed change to revise the note modifying SR 3.8.1.17 to allow the performance of the SR in Modes 1 through 4 when the associated equipment is out of service for maintenance or testing does not create the possibility for an accident or malfunction of a different type than any evaluated previously in SQN's Updated Final Safety Analysis Report. The proposal does not alter the way any structure, system or component function and does not modify the manner in which the plant is operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to TS 3.8.1, ``AC Sources--Operating'' to revise the note modifying SR 3.8.1.17 to allow the performance of the SR in Modes 1 through 4 when the associated equipment is out of service for maintenance or testing does not reduce the margin of safety because the test methodologies are not being changed and LCO [limiting condition for operation] allowed outage times are not being changed. The results of accident analyses remain unchanged by this request. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Branch Chief: Undine S. Shoop. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee Date of amendment request: March 31, 2017. A publicly available version is in ADAMS under Accession No. ML17093A854. Description of amendment request: The amendment would revise Technical Specification (TS) 5.7.2.14, ``Ventilation Filter Testing Program (VFTP),'' to delete references to the reactor building (RB) purge filters. A previous amendment deleted the reactor building purge air cleanup system from the TSs based on partial implementation of the alternate source term methodology; however, references to the RB purge filters were not removed from TS 5.7.2.14 at that time due to an administrative oversight. The proposed change corrects the administrative [[Page 31103]] oversight by deleting references to the RB purge filters in TS 5.7.2.14. 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 revision to WBN TS 5.7.2.1.14 is administrative in nature. Nuclear Regulatory Commission (NRC) Amendment Number 92 (ML13141A564) deleted TS 3.9.8, ``Reactor Building Purge Air Cleanup Units,'' based on implementation of the alternate source term (AST) methodology because no credit is taken for the operation of reactor building air cleanup units for the dose analysis during a fuel handling accident (FHA). However, TVA neglected to remove the references to the RB purge filters in TS 5.7.2.14. The proposed change corrects this oversight by deleting the references to the RB purge filters in TS 5.7.2.14a. through d. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes would not require any new or different accidents to be postulated and subsequently evaluated because no changes are being made to the plant that would introduce any new accident causal mechanisms. This license amendment request does not impact any plant systems that are potential accident initiators, nor does it have any significantly adverse impact on any accident mitigating systems. No new or different accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of these changes. 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 does not alter the permanent plant design, including instrument setpoints, nor does it change the assumptions contained in the safety analyses. Margin of safety is related to the ability of the fission product barriers to perform their design functions during and following accident conditions. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these barriers will not be significantly degraded by the proposed changes. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Branch Chief: Undine S. Shoop. Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee Date of amendment request: March 28, 2017. A publicly-available version is in ADAMS under Accession No. ML17093A608. Description of amendment request: The amendment would revise the Facility Operating License (OL) to extend the completion date for Condition 2.C.(5) regarding the reporting of actions taken to resolve issues identified in Nuclear Regulatory Commission Bulletin 2012-01, ``Design Vulnerability in Electric Power System,'' dated July 27, 2012 (ADAMS Accession No. ML12074A115). 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 revise the completion date for OL Condition 2.C(5) for WBN Unit 2 regarding the reporting of actions taken to resolve issues identified in NRC Bulletin 2012-01 from December 31, 2017 to December 31, 2018 do not affect the structures, systems, or components (SSCs) of the plant, affect plant operations, or any design function or any analysis that verifies the capability of an SSC to perform a design function. No change is being made to any of the previously evaluated accidents in the WBN Updated Final Safety Analysis Report (UFSAR). The proposed changes do not (1) require physical changes to plant SSCs; (2) prevent the safety function of any safety-related system, structure, or component during a design basis event; (3) alter, degrade, or prevent action described or assumed in any accident described in the WBN UFSAR from being performed because the safety-related SSCs are not modified; (4) alter any assumptions previously made in evaluating radiological consequences; or (5) affect the integrity of any fission product barrier. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not introduce any new accident causal mechanisms, because no physical changes are being made to the plant, nor do they affect any plant systems that are potential accident initiators. 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 margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed changes will have no effect on the availability, operability, or performance of safety-related systems and components. The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis. The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely affect plant-operating margins or the reliability of equipment credited in the safety analyses. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Branch Chief: Undine S. Shoop. 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 [[Page 31104]] applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendment request: November 18, 2016. Brief description of amendments: The amendments adopted the approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-535, revising the Technical Specification definition of Shutdown Margin (SDM) to require calculation of the SDM at a reactor moderator temperature of 68 degrees Fahrenheit, or a higher temperature that represents the most reactive state throughout the operating cycle. Date of issuance: June 7, 2017. Effective date: As of date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 277 and 305. A publicly-available version is in ADAMS under Accession No. ML17088A396; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. DPR-71 and DPR-62: Amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: January 17, 2017 (82 FR 4929). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 7, 2017. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant, Van Buren County, Michigan Date of amendment request: November 9, 2016. Brief description of amendment: The amendment revises Technical Specification (TS) 5.5.10, ``Ventilation Filter Testing Program,'' to correct and modify the description of the control room ventilation and fuel handling area ventilation systems. In addition, the amendment corrects an editorial omission in TS Limiting Condition for Operation 3.0.9. Date of issuance: June 8, 2017. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment No.: 263. A publicly-available version is in ADAMS under Accession No. ML17121A510; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-20: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 14, 2017 (82 FR 10596). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8, 2017. No significant hazards consideration comments received: No. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of application for amendment: October 26, 2016. Brief description of amendment: The amendment changed the Technical Specifications (TS) to revise requirements for unavailable barriers by adding new Limiting Condition for Operation (LCO) 3.0.9. This LCO establishes conditions under which systems would remain operable when required physical barriers are not capable of providing their related support function. This amendment is consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-427, Revision 2, ``Allowance for Non Technical Specification Barrier Degradation on Supported System OPERABILITLY.'' The Notice of Availability of this TS improvement and the model application was published in the Federal Register on October 3, 2006 (71 FR 58444), as part of the consolidated line item improvement process. Date of issuance: June 7, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No: 212. A publicly-available version is in ADAMS under Accession No. ML17116A032; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-29: The amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: December 20, 2016 (81 FR 92866). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 7, 2017. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: November 1, 2016. Brief description of amendment: The amendment revised Technical Specification (TS) 2.1.1, ``Reactor Core Safety Limits,'' to reduce the reactor steam dome pressure value specified in TS 2.1.1.1 and TS 2.1.1.2 from 785 pounds per square inch gauge (psig) to 686 psig. Date of issuance: June 19, 2017. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 176. A publicly-available version is in ADAMS under Accession No. ML17139C372; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-58: Amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: December 20, 2016 (81 FR 92868). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 19, 2017. No significant hazards consideration comments received: No. [[Page 31105]] Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2, Berrien County, Michigan Date of amendment request: October 18, 2016, as supplemented by letter dated February 27, 2017. Brief description of amendments: The amendments revised the CNP, Unit Nos. 1 and 2, Technical Specification 5.5.14, ``Containment Leakage Rate Testing Program,'' to clarify the containment leakage rate testing pressure criteria. Date of issuance: June 7, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 336 for Unit No. 1 and 318 for Unit No. 2. A publicly-available version is in ADAMS under Accession No. ML17131A277; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: December 6, 2016 (81 FR 87972). The supplemental letter dated February 27, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the 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 June 7, 2017. No significant hazards consideration comments received: No. Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: July 28, 2016. Brief description of amendment: The amendment adopts TSTF-545, Revision 3, ``TS [technical specification] Inservice Testing Program Removal & Clarify SR [surveillance requirements] Usage Rule Application to Section 5.5 Testing.'' Date of issuance: June 16, 2017. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment No.: 194. A publicly-available version is in ADAMS under Accession No. ML17123A321; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-22: Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: October 11, 2016 (81 FR 70181). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 16, 2017. 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: October 9, 2015, as supplemented on December 1, 2015, August 11, 2016, and December 21, 2016. Description of amendment: This amendment revises License Condition (LC) 2.D(12)(c)1. related to initial Emergency Action Levels (EALs). The LC will require the licensee to submit a fully-developed set of EALs before initial fuel load in accordance with the criteria defined in this license amendment. Date of issuance: April 10, 2017. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment Nos.: 68 (Unit 2) and 68 (Unit 3). A publicly-available version is in ADAMS under Accession Package No. ML16214A135; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses. Date of initial notice in Federal Register: January 19, 2016 (81 FR 2919). The supplemental letters dated December 1, 2015, August 11, 2016, and December 21, 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 the Safety Evaluation dated April 10, 2017. No significant hazards consideration comments received: No. South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield, South Carolina Date of amendment request: January 20, 2017, and supplemented by letter dated March 8, 2017. Description of amendment: The amendment consists of changes to the VCSNS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant specific Design Control Document Tier 2 information. Specifically, the amendment consists of changes to the UFSAR to provide clarification of the interface criteria for nonsafety-related instrumentation that monitors safety-related fluid systems. Date of issuance: May 31, 2017. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 74. A publicly-available version is in ADAMS under Accession No. ML17130A903; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses. Date of initial notice in Federal Register: February 28, 2017 (82 FR 12130). The supplemental letter dated March 8, 2017, provided additional information that clarified the application, did not expand the scope of the application request 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 the Safety Evaluation dated May 31, 2017. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: February 15, 2016, as supplemented by letters dated August 19, 2016, August 26, 2016, September 13, 2016, December 16, 2016, and March 17, 2017. Description of amendment: The amendment authorizes changes to the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information and involves related changes to the associated plant-specific Tier 2* information. Specifically, the departures [[Page 31106]] consist of changes to UFSAR text and tables, and information incorporated by reference into the UFSAR related to updates to WCAP- 16096, ``Software Program Manual for Common Q\TM\ Systems,'' and WCAP- 16097, ``Common Qualified Platform Topical Report.'' Date of issuance: June 8, 2017. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment Nos.: 79 (Unit 3) and 78 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML17104A109; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined License. Date of initial notice in Federal Register: April 12, 2016 (81 FR 21602). The supplemental letters dated August 19, 2016, August 26, 2016, September 13, 2016, December 16, 2016, and March 17, 2017, provided additional information that clarified the application, did not expand the scope of the application request as noticed on February 15, 2016, and did not change the staff's proposed no significant hazards consideration determination as published in the Federal Register on April 12, 2016. The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated June 8, 2017. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 23rd day of June 2017. For the Nuclear Regulatory Commission. Kathryn M. Brock, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2017-13804 Filed 7-3-17; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices 31089
Secretary, 11555 Rockville Pike, hearing or ADR, the provisions specified this document describes a different
Rockville, Maryland, 20852, Attention: in Section IV above shall be final 30 method for submitting comments on a
Rulemaking and Adjudications Staff. days from the date of this Order without specific subject):
Participants filing adjudicatory further order or proceedings. If an • Federal Rulemaking Web site: Go to
document in this manner are extension of time for requesting a http://www.regulations.gov and search
responsible for serving the document on hearing or ADR has been approved, the for Docket ID NRC–2017–0152. Address
all other participants. Filing is provisions specified in Section IV shall questions about NRC dockets to Carol
considered complete by first-class mail be final when the extension expires if a Gallagher; telephone: 301–415–3463;
as of the time of deposit in the mail, or hearing or ADR request has not been email: Carol.Gallagher@nrc.gov. For
by courier, express mail, or expedited received. If ADR is requested, the technical questions, contact the
delivery service upon depositing the provisions specified in Section IV shall individual listed in the FOR FURTHER
document with the provider of the be final upon termination of an ADR INFORMATION CONTACT section of this
service. A presiding officer, having process that did not result in issuance document.
granted an exemption request from of an Order. If payment has not been • Mail comments to: Cindy Bladey,
using E-Filing, may require a participant made by the time specified above, the Office of Administration, Mail Stop:
or party to use E-Filing if the presiding matter may be referred to the Attorney TWFN–8–D36M, U.S. Nuclear
officer subsequently determines that the General for collection. Regulatory Commission, Washington,
reason for granting the exemption from Dated at Rockville, Maryland, this 27th day DC 20555–0001.
use of E-Filing no longer exists. of June 2017. For additional direction on obtaining
Documents submitted in adjudicatory For the Nuclear Regulatory Commission. information and submitting comments,
proceedings will appear in the NRC’s see ‘‘Obtaining Information and
electronic hearing docket, which is Patricia K. Holahan,
Director, Office of Enforcement. Submitting Comments’’ in the
available to the public at https:// SUPPLEMENTARY INFORMATION section of
adams.nrc.gov/ehd/, unless excluded [FR Doc. 2017–14069 Filed 7–3–17; 8:45 am]
this document.
pursuant to an Order of the Commission BILLING CODE 7590–01–P
or the presiding officer. If you do not FOR FURTHER INFORMATION CONTACT: Kay
have an NRC-issued digital ID certificate Goldstein, Office of Nuclear Reactor
as described above, click ‘‘Cancel’’ NUCLEAR REGULATORY Regulation, U.S. Nuclear Regulatory
when the link requests certificates and COMMISSION Commission, Washington, DC 20555–
you will be automatically directed to the 0001; telephone: 301–415–1506, email:
[NRC–2017–0152]
NRC’s electronic hearing dockets where Kay.Goldstein@nrc.gov.
you will be able to access any publicly Biweekly Notice; Applications and SUPPLEMENTARY INFORMATION:
available documents in a particular Amendments to Facility Operating
I. Obtaining Information and
hearing docket. Participants are Licenses and Combined Licenses
requested not to include personal Submitting Comments
Involving No Significant Hazards
privacy information, such as social Considerations A. Obtaining Information
security numbers, home addresses, or
AGENCY: Nuclear Regulatory Please refer to Docket ID NRC–2017–
home phone numbers in their filings,
Commission. 0152, facility name, unit number(s),
unless an NRC regulation or other law
ACTION: Biweekly notice. plant docket number, application date,
requires submission of such
information. For example, in some and subject when contacting the NRC
SUMMARY: Pursuant to Section 189a.(2) about the availability of information for
instances, individuals provide home of the Atomic Energy Act of 1954, as
addresses in order to demonstrate this action. You may obtain publicly-
amended (the Act), the U.S. Nuclear available information related to this
proximity to a facility or site. With
Regulatory Commission (NRC) is action by any of the following methods:
respect to copyrighted works, except for
limited excerpts that serve the purpose
publishing this regular biweekly notice. • Federal rulemaking Web site: Go to
The Act requires the Commission to http://www.regulations.gov and search
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publish notice of any amendments for Docket ID NRC–2017–0152.
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participants are requested not to include
issued, or proposed to be issued, and • NRC’s Agencywide Documents
grants the Commission the authority to Access and Management System
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If a person other than Somascan any amendment to an operating license available documents online in the
requests a hearing, that person shall set or combined license, as applicable, ADAMS Public Documents collection at
forth with particularity the manner in upon a determination by the http://www.nrc.gov/reading-rm/
which his interest is adversely affected Commission that such amendment adams.html. To begin the search, select
by this Order and shall address the involves no significant hazards ‘‘ADAMS Public Documents’’ and then
criteria set forth in 10 CFR 2.309(d) and consideration, notwithstanding the select ‘‘Begin Web-based ADAMS
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Somascan or a person whose interest is request for a hearing from any person. please contact the NRC’s Public
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proposed to be issued, from June 3, 2017 email to pdr.resource@nrc.gov. The
sradovich on DSK3GMQ082PROD with NOTICES
place of any hearings. If a hearing is
held, the issue to be considered at such to June 19, 2017. The last biweekly ADAMS accession number for each
hearing shall be whether this Order notice was published on June 19, 2017. document referenced (if it is available in
should be sustained. DATES: Comments must be filed by ADAMS) is provided the first time that
In the absence of any request for a August 4, 2017. A request for a hearing it is mentioned in this document.
hearing or alternative dispute resolution must be filed by September 5, 2017. • NRC’s PDR: You may examine and
(ADR), or written approval of an ADDRESSES: You may submit comments purchase copies of public documents at
extension of time in which to request a by any of the following methods (unless the NRC’s PDR, Room O1–F21, One
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![Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices 31093
consideration, which is presented Therefore, it is concluded that this change probability or consequences of an accident
below: does not involve a significant reduction in a previously evaluated.
margin of safety. 2. Does the proposed change create the
1. Does the proposed change involve a possibility of a new or different kind of
significant increase in the probability or The NRC staff has reviewed the accident from any accident previously
consequences of an accident previously licensee’s analysis and, based on this evaluated?
evaluated? review, it appears that the three Response: No.
Response: No. standards of 10 CFR 50.92(c) are The proposed changes to the SG Program
The proposed change revises the Steam satisfied. Therefore, the NRC staff will not introduce any adverse changes to the
Generator (SG) Program to modify the proposes to determine that the plant design basis or postulated accidents
frequency of verification of SG tube integrity resulting from potential tube degradation.
and SG tube sample selection. A steam
amendment request involves no
The proposed change does not affect the
generator tube rupture (SGTR) event is one of significant hazards consideration. design of the SGs or their method of
the design basis accidents that are analyzed Attorney for licensee: Ms. Anna operation. In addition, the proposed change
as part of a plant’s licensing basis. The Vinson Jones, Senior Counsel, Entergy does not impact any other plant system or
proposed SG tube inspection frequency and Services, Inc., 101 Constitution Avenue component.
sample selection criteria will continue to NW., Suite 200 East, Washington, DC Therefore, it is concluded that this change
ensure that the SG tubes are inspected such 20001. does not create the possibility of a new or
that the probability of a[n] SGTR is not NRC Branch Chief: Robert J. different kind of accident from an accident
increased. The consequences of a[n] SGTR previously evaluated.
Pascarelli. 3. Does the proposed change involve a
are bounded by the conservative assumptions
in the design basis accident analysis. The Entergy Operations, Inc., Docket No. 50– significant reduction in a margin of safety?
proposed change will not cause the 368, Arkansas Nuclear One, Unit No. 2, Response: No.
consequences of a[n] SGTR to exceed those The SG tubes in pressurized water reactors
Pope County, Arkansas
assumptions. are an integral part of the reactor coolant
Therefore, it is concluded that this change Date of amendment request: April 24, pressure boundary and, as such, are relied
does not involve a significant increase in the 2017. A publicly-available version is in upon to maintain the primary system’s
probability or consequences of an accident ADAMS under Accession No. pressure and inventory. As part of the reactor
previously evaluated. ML17114A399. coolant pressure boundary, the SG tubes are
2. Does the proposed change create the unique in that they are also relied upon as
Description of amendment request: a heat transfer surface between the primary
possibility of a new or different kind of
accident from any accident previously
The amendment would revise Technical and secondary systems such that residual
evaluated? Specification requirements regarding heat can be removed from the primary
Response: No. steam generator tube inspections and system. In addition, the SG tubes also isolate
The proposed changes to the SG Program reporting as described in Technical the radioactive fission products in the
will not introduce any adverse changes to the Specifications Task Force (TSTF) primary coolant from the secondary system.
plant design basis or postulated accidents Traveler TSTF–510, Revision 2, In summary, the safety function of a[n] SG is
resulting from potential tube degradation. maintained by ensuring the integrity of its
‘‘Revision to Steam Generator Program
The proposed change does not affect the tubes.
Inspection Frequencies and Tube SG tube integrity is a function of the
design of the SGs or their method of Sample Selection,’’ using the
operation. In addition, the proposed change design, environment, and the physical
does not impact any other plant system or
Consolidated Line Item Improvement condition of the tube. The proposed change
component. Process for Arkansas Nuclear One, Unit does not affect tube design or operating
Therefore, it is concluded that this change No. 2. environment. The proposed change will
does not create the possibility of a new or Basis for proposed no significant continue to require monitoring of the
different kind of accident from an accident hazards consideration determination: physical condition of the SG tubes such that
previously evaluated. there will not be a reduction in the margin
As required by 10 CFR 50.91(a), the
3. Does the proposed change involve a of safety compared to the current
licensee has provided its analysis of the requirements.
significant reduction in a margin of safety? issue of no significant hazards
Response: No. Therefore, it is concluded that this change
consideration, which is presented does not involve a significant reduction in a
The SG tubes in pressurized water reactors
are an integral part of the reactor coolant
below: margin of safety.
pressure boundary and, as such, are relied 1. Does the proposed change involve a The NRC staff has reviewed the
upon to maintain the primary system’s significant increase in the probability or licensee’s analysis and, based on this
pressure and inventory. As part of the reactor consequences of an accident previously review, it appears that the three
coolant pressure boundary, the SG tubes are evaluated? standards of 10 CFR 50.92(c) are
unique in that they are also relied upon as Response: No.
a heat transfer surface between the primary
satisfied. Therefore, the NRC staff
The proposed change revises the Steam
and secondary systems such that residual Generator (SG) Program to modify the proposes to determine that the
heat can be removed from the primary frequency of verification of SG tube integrity amendment request involves no
system. In addition, the SG tubes also isolate and SG tube sample selection. A steam significant hazards consideration.
the radioactive fission products in the generator tube rupture (SGTR) event is one of Attorney for licensee: Ms. Anna
primary coolant from the secondary system. the design basis accidents that are analyzed Vinson Jones, Senior Counsel, Entergy
In summary, the safety function of a[n] SG is as part of a plant’s licensing basis. The Services, Inc., 101 Constitution Avenue
maintained by ensuring the integrity of its proposed SG tube inspection frequency and NW., Suite 200 East, Washington, DC
tubes. sample selection criteria will continue to 20001.
SG tube integrity is a function of the ensure that the SG tubes are inspected such NRC Branch Chief: Robert J.
design, environment, and the physical that the probability of a[n] SGTR is not
sradovich on DSK3GMQ082PROD with NOTICES
Pascarelli.
condition of the tube. The proposed change increased. The consequences of a[n] SGTR
does not affect tube design or operating are bounded by the conservative assumptions Entergy Operations, Inc., Docket No. 50–
environment. The proposed change will in the design basis accident analysis. The 382, Waterford Steam Electric Station,
continue to require monitoring of the proposed change will not cause the Unit 3 (Waterford 3), St. Charles Parish,
physical condition of the SG tubes such that consequences of a[n] SGTR to exceed those Louisiana
there will not be a reduction in the margin assumptions.
of safety compared to the current Therefore, it is concluded that this change Date of amendment request: March
requirements. does not involve a significant increase in the 28, 2017. A publicly-available version is
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![31094 Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices
in ADAMS under Accession No. accident from any accident previously level less than or equal to 3358
ML17087A551. evaluated? megawatts thermal (MWt). This change
Description of amendment request: Response: No. would be in effect for the current
The proposed amendment would revise The change does not involve a physical
PBAPS, Unit 2, Cycle 22 that is
alteration of the plant (i.e., no new or
Technical Specification (TS) 3.8.1.3, scheduled to end in October 2018.
different type of equipment will be installed)
‘‘Diesel Fuel Oil,’’ by relocating the or a change in the methods governing normal Basis for proposed no significant
current stored diesel fuel oil numerical plant operation. The change does not alter hazards consideration determination:
volume requirements from the TS to the assumptions made in the safety analysis but As required by 10 CFR 50.91(a), the
TS Bases. In addition, the proposed ensures that the diesel generator operates as licensee has provided its analysis of the
amendment would revise TS 3.8.1.1, assumed in the accident analysis. The issue of no significant hazards
‘‘A.C. [Alternating Current] Sources— proposed change is consistent with the safety consideration, which is presented
Operating,’’ and TS 3.8.1.2, ‘‘A.C. analysis assumptions. Therefore, the below, with NRC staff edits in square
Sources—Shutdown,’’ to relocate the proposed change does not create the brackets:
possibility of a new or different kind of
specific numerical value for feed tank 1. Does the proposed change involve a
accident from any accident previously
fuel oil volume to the TS Bases and evaluated. significant increase in the probability or
replace it with the feed tank time 3. Does the proposed change involve a consequences of an accident previously
requirement. The proposed changes are significant reduction in a margin of safety? evaluated?
consistent with Technical Specifications Response: No. Response: No.
Task Force (TSTF) Traveler TSTF–501, The proposed changes revise [TS] 3.8.1.3 The proposed change would revise TS
Revision 1, ‘‘Relocate Fuel Oil and Lube (Diesel Fuel Oil) by removing the current Section 3.4.3 to decrease the required
stored diesel fuel oil numerical volume number of Safety Relief Valves (SRVs) and
Oil Volume Values to Licensee
requirements from the TS and replacing them Safety Valves (SVs) from a total of 13 to 12,
Control.’’ under reduced reactor thermal power
Basis for proposed no significant with diesel operating time requirements. As
the bases for the existing limits on diesel fuel operation of 3358 MWt (approximately 85%
hazards consideration determination: of Current Licensed Thermal Power (CLTP)).
oil are not changed, no change is made to the
As required by 10 CFR 50.91(a), the accident analysis assumptions and no margin A compensatory reduction in maximum
licensee has provided its analysis of the of safety is reduced as part of this change. allowed reactor power to 3358 MWt has been
issue of no significant hazards The proposed change also removes the TS determined to conservatively offset the
consideration, which is presented 3.8.1.1 and TS 3.8.1.2 diesel feed tank fuel impact/effects of operation with an
below: oil numerical volume requirements and additional (up to 2) SRVs/SVs Out-of-Service.
replaces them with the diesel one hour diesel The Reactor Pressure Vessel (RPV)
1. Does the proposed change involve a overpressure protection capability of the 12
generator operation requirement. As the basis
significant increase in the probability or operable SRVs and SVs is adequate at the
consequences of an accident previously for the existing limits on diesel fuel oil are
not changed, no change is made to the lower power level to ensure the ASME
evaluated? [American Society of Mechanical Engineers]
Response: No. accident analysis assumptions and no margin
of safety is reduced as part of this change. code allowable peak pressure limits are not
The proposed changes revise [TS] 3.8.1.3 exceeded. With the maximum thermal power
(Diesel Fuel Oil) by removing the current Therefore, the proposed change does not
involve a significant reduction in a margin of limitation condition, the proposed change
stored diesel fuel oil numerical volume
safety. has no adverse effect on plant operation, or
requirements from the TS and replacing them
the availability or operation of any accident
with diesel operating time requirements. The The NRC staff has reviewed the mitigation equipment. The plant response to
specific volume of fuel oil equivalent to a 7 licensee’s analysis and, based on this the design basis accidents, Anticipated
and 6 day supply is calculated using the NRC
approved methodology described in review, it appears that the three Operational Occurrence (AOO) events and
Regulatory Guide 1.137, Revision 1, ‘‘Fuel- standards of 10 CFR 50.92(c) are Special Events remains bounded by existing
Oil Systems for Standby Diesel Generators’’ satisfied. Therefore, the NRC staff analyses. The proposed change does not
and [American Nuclear Standards Institute proposes to determine that the require any new or unusual operator actions.
(ANSI)] N195–1976, ‘‘Fuel Oil Systems for amendment request involves no The proposed change does not introduce any
Standby Diesel-Generators’’ using the time new failure modes that could result in a new
significant hazards consideration. or different accident. The SRVs and SVs are
dependent load method as approved in Attorney for licensee: Ms. Anna
Waterford 3 License Amendment 157. not being modified or operated differently
Because the requirement to maintain a 7 day
Vinson Jones, Senior Counsel, Entergy and will continue to operate to meet the
supply of diesel fuel oil is not changed and Services, Inc., 101 Constitution Avenue design basis requirements for RPV
is consistent with the assumptions in the NW., Suite 200 East, Washington, DC overpressure protection. The proposed
accident analyses, and the actions taken 20001. change does not alter the manner in which
when the volume of fuel oil is less than a 6 NRC Branch Chief: Robert J. the RPV overpressure protection system is
day supply have not changed, neither the Pascarelli. operated and functions and thus, there is no
probability nor the consequences of any significant impact on reactor operation.
accident previously evaluated will be Exelon Generation Company, LLC and There is no change being made to safety
affected. PSEG Nuclear LLC, Docket No. 50–277, limits or limiting safety system settings that
The proposed change also removes the TS Peach Bottom Atomic Power Station would adversely affect plant safety as a result
3.8.1.1 and TS 3.8.1.2 diesel feed tank fuel (PBAPS), Unit 2, York and Lancaster of the proposed change.
oil numerical volume requirements and Counties, Pennsylvania For PBAPS, the limiting overpressure AOO
replaces them with the diesel one hour diesel event is the main steam isolation valve
generator operation requirement. The specific Date of amendment request: May 19, closure with scram on high flux (MSIVF).
volume and time is not changed and is 2017. A publicly-available version is in The PBAPS ATWS [anticipated transients
consistent with the existing plant design ADAMS under Accession No. without scram] Special Event evaluation
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basis to support a diesel generator under ML17139D357. considered the limiting cases for RPV
accident load conditions. Description of amendment request: overpressure and is analyzed under two
Therefore, the proposed changes do not cases: (1) Main Steam Isolation Valve Closure
involve a significant increase in the
The amendment would revise the (MSIVC) and (2) Pressure Regulator Failure
probability or consequences of an accident Technical Specifications (TSs) to Open (PRFO). These events were analyzed
previously evaluated. decrease the number of safety relief under the proposed conditions and it was
2. Does the proposed change create the valves and safety valves required to be confirmed that the existing analyses remain
possibility of a new or different kind of operable when operating at a power bounding for the condition of adding a
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second SRV/SV Out-of-Service with a limited additional (up to 2) SRVs/SVs Out-of-Service. issue of no significant hazards
maximum operating power level of 3358 The RPV overpressure protection capability consideration which is presented below:
MWt. of the 12 operable SRVs and SVs is adequate
Therefore, the proposed change does not at the lower power level to ensure the ASME 1. Does the proposed amendment involve
involve a significant increase in the code allowable peak pressure limits are not a significant increase in the probability or
probability or consequences of an accident exceeded. The plant response to the design consequences of an accident previously
previously evaluated. basis accidents, AOO events and Special evaluated?
2. Does the proposed change create the Events remains bounded by existing Response: No.
possibility of a new or different kind of No physical changes to the facility will
analyses. These events were analyzed under
accident from any accident previously occur as a result of this proposed
the proposed conditions and it was
evaluated? amendment. The proposed changes will not
confirmed that the existing analyses remain
Response: No. alter the physical design. The current TS
bounding for the condition of adding a
The proposed change would revise TS (CTS) Note in SR 3.5.1.4, SR 3.5.2.4, and
second SRV/SV Out-of-Service with a limited 3.6.1.7 could make CPS susceptible to
Section 3.4.3 to decrease the required maximum operating power level of 3358
number of SRVs and SVs from a total of 13 potential water hammer in the RHR system
MWt. while operating in the SDC mode of RHR in
to 12, under reduced reactor thermal power Therefore, the proposed change does not MODE 3 when swapping from the SDC to
operation of 3358 MWt (approximately 85% involve a significant reduction in a margin of LPCI [low-pressure coolant injection] and
of CLTP). A compensatory reduction in safety. RHR containment spray modes of RHR.
maximum allowed reactor power to 3358
The NRC staff has reviewed the Deletion of the Note from SR 3.5.1.2, SR
MWt has been determined to conservatively
licensee’s analysis and, based on this 3.5.2.4, and SR 3.6.1.7.1 will eliminate the
offset the impact/effects of operation with an
risk for cavitation of the pump and voiding
additional (up to 2) SRVs/SVs Out-of-Service. review, it appears that the three
in the suction piping, thereby avoiding the
The RPV overpressure protection capability standards of 10 CFR 50.92(c) are potential to damage the RHR system,
of the 12 operable SRVs and SVs is adequate satisfied. Therefore, the NRC staff including water hammer. The addition of
at the lower power level to ensure the ASME proposes to determine that the proposed TS note to LCO 3.5.1, LCO 3.5.2,
code allowable peak pressure limits are not amendment request involves no LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3
exceeded. The SRVs and SVs are not being will re-establish consistency of the CPS RHR
modified or operated differently and will
significant hazards consideration.
Attorney for licensee: Tamra Domeyer, system design with the original TS
continue to operate to meet the design basis requirements.
requirements for RPV overpressure Associate General Counsel, Exelon
Therefore, the proposed change does not
protection. The proposed change does not Generation Company, LLC, 4300 involve a significant increase in the
introduce any new failure modes that could Winfield Rd., Warrenville, IL 60555. probability or consequences of an accident
result in a new or different accident. The NRC Branch Chief: James G. Danna. previously evaluated.
proposed reactor thermal power restriction of 2. Does the proposed change create the
3358 MWt is within the existing normal Exelon Generation Company, LLC,
possibility of a new or different kind of
operating domain and no new or special Docket No. 50–461, Clinton Power accident from any accident previously
operating actions are necessary to operate at Station (CPS), Unit No.1, DeWitt evaluated?
the intermediate power level. The proposed County, Illinois Response: No.
change does not alter the manner in which The proposed change does not alter the
Date of amendment request: May 4,
the RPV overpressure protection system is physical design, safety limits, or safety
operated and functions and thus, there is no 2017. A publicly-available version is in analysis assumptions associated with the
new failure mechanisms for the overpressure ADAMS under Accession No. operation of the plant. Accordingly, the
protection system. The plant response to the ML17124A121. change does not introduce any new accident
design basis accidents, AOO events and Description of amendment request: initiators, nor does it reduce or adversely
Special Events remains bounded by existing The proposed change would delete a affect the capabilities of any plant structure,
analyses. [These] events were analyzed under surveillance requirement (SR) Note system, or component to perform their safety
the proposed conditions and it was associated with technical specification function. Deletion of the Note from SR
confirmed that the existing analyses remain (TS) 3.5.1, ‘‘ECCS [emergency core 3.5.1.2, SR 3.5.2.4 and SR 3.6.1.7.1 is
bounding for the condition of adding a appropriate because current TSs could put
cooling system]—Operating,’’ TS 3.5.2, the plant at risk for potential cavitation of the
second SRV/SV Out-of-Service with a limited
maximum operating power level of 3358
‘‘ECCS—Shutdown,’’ and TS 3.6.1.7, pump and voiding in the suction piping,
MWt. ‘‘Residual Heat Removal (RHR) resulting in potential to damage the RHR
Therefore, the proposed change does not Containment Spray System,’’ to more system, including water hammer. The
create the possibility of a new or different appropriately reflect the RHR system addition of proposed TS note to LCO 3.5.1,
kind of accident from any accident design, and ensure the RHR system LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and
previously evaluated. operation is consistent with the TS LCO 3.6.2.3 will re-establish consistency of
3. Does the proposed change involve a the CPS RHR system design with the original
limiting condition for operation (LCO) TS requirements.
significant reduction in a margin of safety? requirements. In addition, the proposed
Response: No. Therefore, the proposed change does not
amendment would insert a Note in the create the possibility of a new or different
The margin of safety is established though
the design of the plant structures, systems
LCO for TSs 3.5.1, 3.5.2, 3.6.1.7, 3.6.1.9, kind of accident from any previously
and components, the parameters within ‘‘Feedwater Leakage Control System,’’ evaluated.
and 3.6.2.3, ‘‘Residual Heat Removal 3. Does the proposed change involve a
which the plant is operated, and the
(RHR) Suppression Pool Cooling,’’ to significant reduction in a margin of safety?
establishment of setpoints for the actuation of
clarify that one of the required Response: No.
equipment relied upon to respond to an
The proposed change conforms to NRC
event. The proposed change does not change subsystems in each of the affected TS regulatory guidance regarding the content of
the setpoints at which the protective actions sections may be inoperable during plant Technical Specifications. The proposed
are initiated. The proposed change would
sradovich on DSK3GMQ082PROD with NOTICES
alignment and operation of the RHR change does not alter the physical design,
revise TS Section 3.4.3 to decrease the system for shutdown cooling (SDC) with safety limits, or safety analysis assumptions
required number of SRVs and SVs under
reduced reactor thermal power operation of
the reactor steam dome pressure less associated with the operation of the plant.
than the RHR cut in permissive value. Therefore, the proposed change does not
3358 MWt (approximately 85% of CLTP). A involve a significant reduction in a margin of
compensatory reduction in maximum Basis for proposed no significant
hazards consideration determination: safety.
allowed reactor power to 3358 MWt has been
determined to conservatively offset the As required by 10 CFR 50.91(a), the The NRC staff has reviewed the
impact/effects of operation with an licensee has provided its analysis of the licensee’s analysis and, based on this
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![31098 Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices
proposed amendment does not change the Basis for proposed no significant does not involve a reduction in the margin
way the plant is operated or maintained and hazards consideration determination: of safety. The proposed amendment does not
does not create a credible failure mechanism, As required by 10 CFR 50.91(a), the alter the manner in which safety limits,
malfunction or accident initiator not already limiting safety system settings or limiting
considered in the design and licensing basis.
licensee has provided its analysis of the
conditions for operation are determined. The
Therefore, the proposed amendment does issue of no significant hazards safety analysis acceptance criteria are not
not create the possibility of a new or different consideration, which is presented affected by this change. The proposed
kind of accident from any accident below: amendment does not adversely affect existing
previously evaluated. 1. Does the proposed change involve a plant safety margins or the reliability of
3. Does the proposed amendment involve significant increase in the probability or equipment assumed to mitigate accidents in
a significant reduction in a margin of safety? consequences of an accident previously the UFSAR. The proposed amendment does
Response: No. evaluated? not adversely affect the ability of SSCs to
Safety margins are applied to design and Response: No. perform their design function. SSCs required
licensing basis functions and to the The proposed changes are clarifications to to safely shut down the reactor and to
controlling values of parameters to account maintain it in a safe shutdown condition
methods applied to ensure compliance with
for various uncertainties and to avoid remain capable of performing their design
NFPA 30, section 2348. The revised methods
exceeding regulatory or licensing limits. The function.
comply with NFPA 30, section 2348. This
proposed amendment involves changes to the Therefore, this change does not involve a
LAR [license amendment request] is
EPP, which provides for protection of significant reduction in a margin of safety.
essentially an administrative change to revise
nonradiological environmental values during
the letter referenced by the Fire Protection
operation of the nuclear facility. The The NRC staff has reviewed the
Transition License Conditions. The actual
proposed amendment does not involve a licensee’s analysis and, based on this
design changes and any related procedural
physical change to the plant, does not change review, it appears that the three
methods of plant operation within prescribed changes are being managed separately from
this LAR per 10 CFR 50.59. standards of 10 CFR 50.92(c) are
limits, or affect design and licensing basis
functions or controlling values of parameters The proposed change does not adversely satisfied. Therefore, the NRC staff
for plant systems, structures, and affect accident initiators or precursors nor proposes to determine that the
components. alter the design assumptions, conditions, and amendment request involves no
Therefore, the proposed amendment does configuration of the facility or the manner in significant hazards consideration.
not involve a significant reduction in a which the plant is operated and maintained. Attorney for licensee: William S.
margin of safety. The proposed changes do not adversely affect
the ability of structures, systems and
Blair, Managing Attorney—Nuclear,
The NRC staff has reviewed the components (SSCs) to perform their intended Florida Power & Light Company, 700
licensee’s analysis and, based on this safety function to mitigate the consequences Universe Boulevard, MS LAW/JB, Juno
review, it appears that the three of an initiating event within the assumed Beach, FL 33408–0420.
standards of 10 CFR 50.92(c) are acceptance limits. The proposed change does NRC Branch Chief: Undine S. Shoop.
satisfied. Therefore, the NRC staff not increase the probability or consequence
of an accident. Indiana Michigan Power Company
proposes to determine that the (I&M), Docket Nos. 50–315 and 50–316,
Therefore, this change does not involve a
amendment request involves no significant increase in the probability or Donald C. Cook Nuclear Plant (CNP),
significant hazards consideration. consequences of an accident previously Units Nos. 1 and 2, Berrien County,
Attorney for licensee: David W. evaluated. Michigan
Jenkins, Attorney, FirstEnergy 2. Does the proposed change create the
Corporation, Mail Stop A–GO–15, 76 possibility of a new or different kind of Date of amendment request: May 23,
South Main Street, Akron, OH 44308. accident from any accident previously 2017. A publicly-available version is in
NRC Branch Chief: David J. Wrona. evaluated? ADAMS under Accession No.
Response: No. ML17146A073.
Florida Power & Light Company, et al., The proposed changes are clarifications to Description of amendment request:
Docket Nos. 50–335 and 50–389, St. methods applied to ensure compliance with The proposed changes update the
Lucie Plant Unit Nos. 1 and 2, St. Lucie NFPA 30, section 2348. The revised methods
emergency action levels (EALs) used at
County, Florida of compliance align with NFPA 30, section
2348, and will not result in new or different CNP, Unit Nos. 1 and 2, from the
Date of amendment request: May 2, kinds of accidents. This LAR is essentially an current scheme based on Nuclear
2017. A publicly-available version is in administrative change to revise the letter Management and Resources Council
ADAMS under Accession No. referenced by the Fire Protection Transition (NUMARC) and National Environmental
ML17144A294. License Conditions. The actual design Studies Project (NESP) NUMARC/
Description of amendment request: changes and any related procedural changes NESP–007, ‘‘Methodology for
The amendments would revise the St. are being managed separately from this LAR Development of Emergency Action
Lucie Plant Unit Nos. 1 and 2 Renewed per 10 CFR 50.59. Levels’’ dated January 1992, to a scheme
Facility Operating Licenses, Nos. DPR– The requirements in NFPA 30 address only
based on Nuclear Energy Institute 99–
fire protection. The impacts of fire effects on
67 and NPF–16, respectively, fire 01, Revision 6, ‘‘Development of
the plant have been evaluated. The proposed
protection license conditions. The amendment does not involve new failure Emergency Action Levels for Non-
revisions would incorporate new mechanisms or malfunctions that could Passive Reactors.’’
references into these license conditions initiate a new or different kind of accident Basis for proposed no significant
that propose and approve a revision to beyond those already analyzed in the Unit 1 hazards consideration determination:
plant modifications previously and Unit 2 UFSARs [updated final safety As required by 10 CFR 50.91(a), the
approved in the March 31, 2016, NRC analysis reports]. licensee has provided its analysis of the
issuance of amendments regarding Therefore, this change does not create the
sradovich on DSK3GMQ082PROD with NOTICES
issue of no significant hazards
transition to a risk-informed, possibility of a new or different kind of
consideration, which is presented
accident from an accident previously
performance-based fire protection below:
evaluated.
program in accordance with 10 CFR 3. Does the proposed change involve a 1. Does the proposed amendment involve
50.48(c), dated March 21, 2016 (ADAMS significant reduction in a margin of safety? a significant increase in the probability or
Accession No. ML15344A346) (known Response: No. consequences of an accident previously
as the National Fire Protection Operation of Plant St. Lucie (PSL) in evaluated?
Association Standard 805 (NFPA 805)). accordance with the proposed amendment Response: No.
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![Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices 31099
The proposed changes to the CNP EALs do plan will continue to activate an emergency consequences of an accident previously
not impact the physical function of plant response commensurate with the extent of evaluated?
structures, systems, or components (SSC) or degradation of plant safety. Response: No.
the manner in which SSCs perform their Plant safety margins are established The activity adds a second pump to each
design function. EALs are used as criteria for through limiting conditions for operation, of the turbine building main sumps, and
determining the need for notification and limiting safety system settings, and safety identifies that there is more than one turbine
participation of local and State agencies, and limits specified in the technical building sump. The reason for the additional
for determining when and what type of specifications. The proposed changes involve pumps is to account for an increase in
protective measures should be considered references to available plant indications to volume due to the changes to the [condensate
within and outside the site boundary to assess conditions for determination of entry polishing system (CPS)] rinse effluent
protect health and safety. The proposed into an emergency action level. There is no flowpath from [component cooling water
changes neither adversely affect accident change to these established safety margins as system (CCW)] CCW to [waste water system
initiators or precursors, nor alter design a result of this change. (WWS)] WWS via the Turbine Building
assumptions. The proposed changes do not Therefore, the proposed change does not sumps. The extra sump pumps will prevent
alter or prevent the ability of SSCs to perform involve a significant reduction in a margin of potential overflowing and flooding of the
their intended function to mitigate the safety. sumps during CPS rinse operations. The CPS
consequences of an initiating event within serves no safety-related function. By
assumed acceptance limits. No operating The NRC staff has reviewed the directing the effluent to the turbine building
procedures or administrative controls that licensee’s analysis and, based on this sumps it is subject to radiation monitoring.
function to prevent or mitigate accidents are review, it appears that the three Under normal operating conditions, there are
affected by the proposed changes. standards of 10 CFR 50.92(c) are no significant amounts of radioactive
Therefore, the proposed changes do not satisfied. Therefore, the NRC staff contamination within the CPS. However,
involve a significant increase in the proposes to determine that the radioactive contamination of the CPS can
probability or consequences of an accident occur as a result of a primary to secondary
previously evaluated.
amendment request involves no leakage in the steam generator should a steam
2. Does the proposed amendment create significant hazards consideration. generator tube leak develop while the CPS is
the possibility of a new or different kind of Attorney for licensee: Robert B. in operation and radioactive condensate is
accident from any accident previously Haemer, Senior Nuclear Counsel, One processed by the CPS. Radiation monitors
evaluated? Cook Place, Bridgman, MI 49106. associated with the steam generator
Response: No. NRC Branch Chief: David J. Wrona. blowdown, steam generator, and turbine
The proposed changes to the CNP EALs do island vents, drains and relief systems
not involve any physical changes to plant South Carolina Electric & Gas Company, provide the means to determine if the
systems or equipment. The proposed changes Docket Nos. 52–027 and 52–028, Virgil secondary side is radioactively contaminated.
do not involve the addition of any new C. Summer Nuclear Station, Units 2 and The main turbine building sumps and sump
equipment. EALs are based on plant 3, Fairfield County, South Carolina pumps are not safety-related components and
conditions, so the proposed changes will not do not interface with any systems, structures,
alter the design configuration or the method Date of amendment request: May 11, or components (SSC) accident initiator or
of plant operation. The proposed changes 2017. A publicly-available version is in initiating sequence of events; thus, the
will not introduce failure modes that could ADAMS under Accession No. probability of accidents evaluated within the
result in a new or different type of accident, ML17135A225. plant-specific UFSAR are not affected. The
and the change does not alter assumptions Description of amendment request: proposed changes do not involve a change to
made in the safety analysis. The proposed The requested amendment proposes to the predicted radiological releases due to
changes to the CNP Emergency Plan are not accident conditions, thus the consequences
depart from combined license (COL)
initiators of any accidents. of accidents evaluated in the UFSAR are not
Therefore, the proposed change does not Appendix C information (with affected.
create the possibility of a new or different corresponding changes to the associated Therefore, the proposed amendment does
kind of accident from any previously plant-specific Tier 1 information) and not involve a significant increase in the
evaluated. involves associated Tier 2 information probability or consequences of an accident
3. Does the proposed amendment involve in the Updated Final Safety Analysis previously evaluated.
a significant reduction in a margin of safety? Report (UFSAR). Specifically, proposed 2. Does the proposed amendment create
Response: No. changes clarify that there is more than the possibility of a new or different kind of
Margin of safety is associated with the one turbine building main sump and accident from any accident previously
ability of the fission product barriers (i.e., evaluated?
adds a second sump pump for each of
fuel cladding, reactor coolant system Response: No.
pressure boundary, and containment the two turbine building main sumps The proposed changes to the non-safety
structure) to limit the level of radiation dose into UFSAR Tier 2 and COL Appendix waste water system (WWS) do not affect any
to the public. The proposed changes to the C (and associated plant-specific Tier 1) safety-related equipment, nor does it add any
CNP EALs do not impact operation of the information. new interface to safety-related SSCs. No
plant or its response to transient or accidents. Pursuant to the provisions of 10 CFR system or design function or equipment
The changes do not affect the Technical 52.63(b)(1), an exemption from elements qualification is affected by this change. The
Specifications or the operating license. The of the design as certified in the 10 CFR changes do not introduce a new failure mode,
proposed changes do not involve a change in part 52, Appendix D, design malfunction, or sequence of events that could
the method of plant operation, and no affect safety or safety-related equipment.
accident analyses will be affected by the
certification rule is also requested for Therefore, the proposed amendment does
proposed changes. the plant-specific Design Control not create the possibility of a new or different
Additionally, the proposed changes will Document Tier 1 departures. kind of accident from any accident
not relax any criteria used to establish safety Basis for proposed no significant previously evaluated.
limits and will not relax any safety system hazards consideration determination: 3. Does the proposed amendment involve
sradovich on DSK3GMQ082PROD with NOTICES
settings. The safety analysis acceptance As required by 10 CFR 50.91(a), the a significant reduction in a margin of safety?
criteria are not affected by these changes. The licensee has provided its analysis of the Response: No.
proposed changes will not result in plant issue of no significant hazards The WWS is a nonsafety-related system
operation in configuration outside the design that does not interface with any safety-related
consideration, which is presented
basis. The proposed changes do not adversely equipment. The proposed changes to identify
affect systems that respond to safely shut below: that there is more than one turbine building
down the plant and to maintain the plant in 1. Does the proposed amendment involve sump and to add two turbine building sump
a safe shutdown condition. The emergency a significant increase in the probability or pumps do not affect any design code,
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![31102 Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices
consequences of an accident previously input or result, or design/safety margin. No structure, system or component or alter the
evaluated? safety analysis or design basis acceptance modes of plant operation in a manner that is
Response: No. limit/criterion is challenged or exceeded by outside the bounds of the current emergency
The activity adds a second pump to each the proposed change. diesel generator system design analyses. The
of the turbine building main sumps, and Therefore, the proposed amendment does proposed change to revise the note modifying
identifies that there is more than one turbine not involve a significant reduction in a SR 3.8.1.17 to allow the performance of the
building sump. The reason for the additional margin of safety. SR in Modes 1 through 4 when the associated
pumps is to account for an increase in The NRC staff has reviewed the equipment is out of service for maintenance
volume due to the changes to the condensate or testing does not create the possibility for
polishing system (CPS) rinse effluent
licensee’s analysis and, based on this an accident or malfunction of a different type
flowpath from CPS to waste water system review, it appears that the three than any evaluated previously in SQN’s
(WWS) via the turbine building sumps. The standards of 10 CFR 50.92(c) are Updated Final Safety Analysis Report. The
extra sump pumps will prevent potential satisfied. Therefore, the NRC staff proposal does not alter the way any structure,
overflowing and flooding of the sumps proposes to determine that the system or component function and does not
during CPS rinse operations. The CPS serves amendment request involves no modify the manner in which the plant is
no safety-related function. By directing the significant hazards consideration. operated. Therefore, the proposed change
effluent to the turbine building sumps it is Attorney for licensee: M. Stanford does not create the possibility of a new or
subject to radiation monitoring. Under Blanton, Balch & Bingham LLP, 1710 different kind of accident from any accident
normal operating conditions, there are is no previously evaluated.
Sixth Avenue North, Birmingham, AL
significant amount of radioactive 3. Does the proposed amendment involve
contamination within the CPS. However, 35203–2015.
a significant reduction in a margin of safety?
radioactive contamination of the CPS can NRC Branch Chief: Jennifer Dixon-
Response: No.
occur as a result of a primary-to-secondary Herrity. The proposed change to TS 3.8.1, ‘‘AC
leakage in the steam generator should a steam Tennessee Valley Authority, Docket Sources—Operating’’ to revise the note
generator tube leak develop while the CPS is modifying SR 3.8.1.17 to allow the
Nos. 50–327 and 50–328, Sequoyah
in operation and radioactive condensate is performance of the SR in Modes 1 through
processed by the CPS. Radiation monitors
Nuclear Plant (SQN), Units 1 and 2,
4 when the associated equipment is out of
associated with the steam generator Hamilton County, Tennessee service for maintenance or testing does not
blowdown, steam generator, and turbine Date of amendment request: March reduce the margin of safety because the test
island vents, drains and relief systems 13, 2017. A publicly available version is methodologies are not being changed and
provide the means to determine if the in ADAMS under Accession No. LCO [limiting condition for operation]
secondary side is radioactively contaminated. ML17073A018. allowed outage times are not being changed.
The main turbine building sumps and sump Description of amendment request: The results of accident analyses remain
pumps are not safety-related components and unchanged by this request. Therefore, the
do not interface with any systems, structures,
The amendments would modify the
proposed change does not involve a
or components (SSC) accident initiator or Surveillance Requirement (SR) 3.8.1.17 significant reduction in a margin of safety.
initiating sequence of events; thus, the of the Technical Specification (TS)
probability of accidents evaluated within the 3.8.1, ‘‘AC [Alternating Current] The NRC staff has reviewed the
plant-specific UFSAR are not affected. The Sources—Operating,’’ to delete the note licensee’s analysis and, based on this
proposed changes do not involve a change to to allow the performance of the SR in review, it appears that the three
the predicted radioactive releases due to Modes 1 through 4 when the associated standards of 10 CFR 50.92(c) are
accident conditions, thus the consequences load is out of service for maintenance or satisfied. Therefore, the NRC staff
of accidents evaluated in the UFSAR are not testing. proposes to determine that the
affected. amendment request involves no
Basis for proposed no significant
Therefore, the proposed amendment does significant hazards consideration.
not involve a significant increase in the hazards consideration determination:
As required by 10 CFR 50.91(a), the Attorney for licensee: General
probability or consequences of an accident
licensee has provided its analysis of the Counsel, Tennessee Valley Authority,
previously evaluated.
2. Does the proposed amendment create issue of no significant hazards 400 West Summit Hill Drive, 6A West
the possibility of a new or different kind of consideration, which is presented Tower, Knoxville, TN 37902.
accident from any accident previously below: NRC Branch Chief: Undine S. Shoop.
evaluated? Tennessee Valley Authority, Docket No.
1. Does the proposed amendment involve
Response: No.
The proposed changes to the nonsafety-
a significant increase in the probability or 50–390, Watts Bar Nuclear Plant (WBN),
consequence of an accident previously Unit 1, Rhea County, Tennessee
related WWS do not affect any safety-related
evaluated?
equipment, nor do they add any new Date of amendment request: March
Response: No.
interface to safety-related SSCs. No system or The proposal does not alter the function of 31, 2017. A publicly available version is
design function or equipment qualification is any structure, system or component in ADAMS under Accession No.
affected by this change. The changes do not functions, does not modify the manner in ML17093A854.
introduce a new failure mode, malfunction, which the plant is operated, and does not Description of amendment request:
or sequence of events that could affect safety alter equipment out-of-service time. This
or safety-related equipment. Therefore, the The amendment would revise Technical
request does not degrade the ability of the Specification (TS) 5.7.2.14, ‘‘Ventilation
proposed amendment does not create the emergency diesel generator or equipment
possibility of a new or different kind of downstream of the load sequencers to
Filter Testing Program (VFTP),’’ to
accident from any accident previously perform their intended function. delete references to the reactor building
evaluated. Therefore, the proposed change does not (RB) purge filters. A previous
3. Does the proposed amendment involve involve a significant increase in the amendment deleted the reactor building
a significant reduction in a margin of safety? purge air cleanup system from the TSs
sradovich on DSK3GMQ082PROD with NOTICES
probability or consequences of an accident
Response: No. previously evaluated. based on partial implementation of the
The WWS is a nonsafety-related system 2. Does the proposed amendment create alternate source term methodology;
that does not interface with any safety-related the possibility of a new or different kind of
equipment. The proposed changes to identify however, references to the RB purge
accident from any accident previously
that there is more than one turbine building evaluated? filters were not removed from TS
sump and to add two turbine building sump Response: No. 5.7.2.14 at that time due to an
pumps do not affect any design code, The proposed change does not involve any administrative oversight. The proposed
function, design analysis, safety analysis physical changes to plant safety related change corrects the administrative
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![Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices 31105
Indiana Michigan Power Company, Renewed Facility Operating License South Carolina Electric & Gas Company,
Docket Nos. 50–315 and 50–316, Donald No. DPR–22: Amendment revised the Docket Nos. 52–027 and 52–028, Virgil
C. Cook Nuclear Plant (CNP), Unit Nos. Renewed Facility Operating License and C. Summer Nuclear Station (VCSNS),
1 and 2, Berrien County, Michigan Technical Specifications. Units 2 and 3, Fairfield, South Carolina
Date of amendment request: October Date of initial notice in Federal Date of amendment request: January
18, 2016, as supplemented by letter Register: October 11, 2016 (81 FR 20, 2017, and supplemented by letter
dated February 27, 2017. 70181). dated March 8, 2017.
Brief description of amendments: The The Commission’s related evaluation Description of amendment: The
amendments revised the CNP, Unit Nos. amendment consists of changes to the
of the amendment is contained in a
1 and 2, Technical Specification 5.5.14, VCSNS Units 2 and 3 Updated Final
Safety Evaluation dated June 16, 2017.
‘‘Containment Leakage Rate Testing Safety Analysis Report (UFSAR) in the
Program,’’ to clarify the containment No significant hazards consideration form of departures from the
leakage rate testing pressure criteria. comments received: No. incorporated plant specific Design
Date of issuance: June 7, 2017. South Carolina Electric & Gas Company Control Document Tier 2 information.
Effective date: As of the date of and South Carolina Public Service Specifically, the amendment consists of
issuance and shall be implemented changes to the UFSAR to provide
Authority, Docket Nos. 52–027 and 52–
within 90 days of issuance. clarification of the interface criteria for
028, Virgil C. Summer Nuclear Station,
Amendment Nos.: 336 for Unit No. 1 nonsafety-related instrumentation that
Units 2 and 3, Fairfield County, South
and 318 for Unit No. 2. A publicly- monitors safety-related fluid systems.
available version is in ADAMS under Carolina Date of issuance: May 31, 2017.
Accession No. ML17131A277; Date of amendment request: October Effective date: As of the date of
documents related to these amendments 9, 2015, as supplemented on December issuance and shall be implemented
are listed in the Safety Evaluation 1, 2015, August 11, 2016, and December within 30 days of issuance.
enclosed with the amendments. 21, 2016. Amendment No.: 74. A publicly-
Renewed Facility Operating License available version is in ADAMS under
Nos. DPR–58 and DPR–74: Amendments Description of amendment: This Accession No. ML17130A903;
revised the Renewed Facility Operating amendment revises License Condition documents related to this amendment
Licenses and Technical Specifications. (LC) 2.D(12)(c)1. related to initial are listed in the Safety Evaluation
Date of initial notice in Federal Emergency Action Levels (EALs). The enclosed with the amendment.
Register: December 6, 2016 (81 FR LC will require the licensee to submit a Facility Combined Licenses Nos. NPF–
87972). The supplemental letter dated fully-developed set of EALs before 93 and NPF–94: Amendment revised the
February 27, 2017, provided additional initial fuel load in accordance with the Facility Combined Licenses.
information that clarified the criteria defined in this license Date of initial notice in Federal
application, did not expand the scope of amendment. Register: February 28, 2017 (82 FR
the application as originally noticed, Date of issuance: April 10, 2017. 12130). The supplemental letter dated
and did not change the staff’s original March 8, 2017, provided additional
proposed no significant hazards Effective date: As of the date of information that clarified the
consideration determination as issuance and shall be implemented application, did not expand the scope of
published in the Federal Register. within 180 days of issuance. the application request as originally
The Commission’s related evaluation Amendment Nos.: 68 (Unit 2) and 68 noticed, and did not change the staff’s
of the amendments is contained in a (Unit 3). A publicly-available version is original proposed no significant hazards
Safety Evaluation dated June 7, 2017. in ADAMS under Accession Package consideration determination as
No significant hazards consideration No. ML16214A135; documents related published in the Federal Register.
comments received: No. to this amendment are listed in the The Commission’s related evaluation
Safety Evaluation enclosed with the of the amendment is contained in the
Northern States Power Company—
amendment. Safety Evaluation dated May 31, 2017.
Minnesota (NSPM), Docket No. 50–263,
No significant hazards consideration
Monticello Nuclear Generating Plant, Facility Combined Licenses Nos. NPF–
comments received: No.
Wright County, Minnesota 93 and NPF–94: Amendment revised the
Date of amendment request: July 28, Facility Combined Licenses. Southern Nuclear Operating Company,
2016. Date of initial notice in Federal Docket Nos. 52–025 and 52–026, Vogtle
Brief description of amendment: The Register: January 19, 2016 (81 FR Electric Generating Plant (VEGP), Units
amendment adopts TSTF–545, Revision 3 and 4, Burke County, Georgia
2919). The supplemental letters dated
3, ‘‘TS [technical specification] December 1, 2015, August 11, 2016, and Date of amendment request: February
Inservice Testing Program Removal & December 21, 2016, provided additional 15, 2016, as supplemented by letters
Clarify SR [surveillance requirements] information that clarified the dated August 19, 2016, August 26, 2016,
Usage Rule Application to Section 5.5 application, did not expand the scope of September 13, 2016, December 16, 2016,
Testing.’’ the application as originally noticed, and March 17, 2017.
Date of issuance: June 16, 2017. and did not change the staff’s original Description of amendment: The
Effective date: As of the date of proposed no significant hazards amendment authorizes changes to the
issuance and shall be implemented VEGP Units 3 and 4 Updated Final
sradovich on DSK3GMQ082PROD with NOTICES
consideration determination as
within 90 days. published in the Federal Register. Safety Analysis Report (UFSAR) in the
Amendment No.: 194. A publicly- form of departures from the
available version is in ADAMS under The Commission’s related evaluation incorporated plant-specific Design
Accession No. ML17123A321; of the amendment is contained in the Control Document Tier 2 information
documents related to this amendment Safety Evaluation dated April 10, 2017. and involves related changes to the
are listed in the Safety Evaluation No significant hazards consideration associated plant-specific Tier 2*
enclosed with the amendment. comments received: No. information. Specifically, the departures
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![31106 Federal Register / Vol. 82, No. 127 / Wednesday, July 5, 2017 / Notices
consist of changes to UFSAR text and WEDNESDAY, JULY 12, 2017, THURSDAY, JULY 13, 2017,
tables, and information incorporated by CONFERENCE ROOM T–2B1, 11545 CONFERENCE ROOM T–2B1, 11545
reference into the UFSAR related to ROCKVILLE PIKE, ROCKVILLE, ROCKVILLE PIKE, ROCKVILLE,
updates to WCAP–16096, ‘‘Software MARYLAND 20852 MARYLAND 20852
Program Manual for Common QTM 8:30 a.m.–8:35 a.m.: Opening
Systems,’’ and WCAP–16097, ‘‘Common 8:30 a.m.–8:35 a.m.: Opening
Remarks by the ACRS Chairman Remarks by the ACRS Chairman
Qualified Platform Topical Report.’’ (Open)—The ACRS Chairman will make
(Open)—The ACRS Chairman will make
Date of issuance: June 8, 2017. opening remarks regarding the conduct
opening remarks regarding the conduct
of the meeting.
Effective date: As of the date of of the meeting.
8:35 a.m.–10:00 p.m.: Future ACRS
issuance and shall be implemented 8:35 a.m.–10:30 a.m.: License Renewal Activities/Report of the Planning and
within 30 days of issuance. Application for the South Texas Project Procedures Subcommittee and
Amendment Nos.: 79 (Unit 3) and 78 (STP) (Open)—The Committee will hear Reconciliation of ACRS Comments and
(Unit 4). A publicly-available version is briefings by and hold discussions with Recommendations (Open/Closed)—The
in ADAMS under Accession No. representatives of the NRC staff and the Committee will discuss the
ML17104A109; documents related to STP Nuclear Operating Co. regarding recommendations of the Planning and
this amendment are listed in the Safety the associated safety evaluation for Procedures Subcommittee regarding
Evaluation enclosed with the license renewal. items proposed for consideration by the
amendment. 10:45 a.m.–12:15 p.m.: NuScale Full Committee during future ACRS
Facility Combined Licenses Nos. NPF– Topical Report TR–0815–16497, ‘‘Safety Meetings, and matters related to the
91 and NPF–92: Amendment revised the Classification of Passive Nuclear Power conduct of ACRS business, including
Facility Combined License. Plant Electrical Systems’’ (Open/ anticipated workload and member
Closed)—The Committee will hear assignments. The Committee will
Date of initial notice in Federal discuss the responses from the NRC
Register: April 12, 2016 (81 FR 21602). briefings by and hold discussions with
representatives of the NRC staff and Executive Director for Operations to
The supplemental letters dated August comments and recommendations
19, 2016, August 26, 2016, September NuScale regarding the safety evaluation
included in recent ACRS reports and
13, 2016, December 16, 2016, and March associated with the subject topical
letters. [NOTE: A portion of this meeting
17, 2017, provided additional report. [NOTE: A portion of this session
may be closed pursuant to 5 U.S.C. 552b
information that clarified the may be closed in order to discuss and
(c) (2) and (6) to discuss organizational
application, did not expand the scope of protect information designated as
and personnel matters that relate solely
proprietary, pursuant to 5 U.S.C.
the application request as noticed on to internal personnel rules and practices
552b(c)(4)].
February 15, 2016, and did not change of the ACRS, and information the
the staff’s proposed no significant 1:15 p.m.–3:45 p.m.: Advanced Power release of which would constitute a
hazards consideration determination as Reactor 1400 (APR1400) (Open/ clearly unwarranted invasion of
published in the Federal Register on Closed)—The Committee will hear personal privacy.]
April 12, 2016. briefings by and hold discussions with 10:15 a.m.–12:00 p.m.: Preparation of
representatives of the NRC staff and ACRS Reports (Open/Closed)—The
The Commission’s related evaluation
Korea Hydro & Nuclear Power regarding Committee will continue its discussion
of the amendment is contained in the
selected chapters of the safety of proposed ACRS reports. [NOTE: A
Safety Evaluation dated June 8, 2017.
evaluation associated with the APR1400 portion of this session may be closed in
No significant hazards consideration Design Certification. [NOTE: A portion order to discuss and protect information
comments received: No. of this session may be closed in order designated as proprietary, pursuant to 5
Dated at Rockville, Maryland, this 23rd day to discuss and protect information U.S.C. 552b(c)(4)].
of June 2017. designated as proprietary, pursuant to 5 1:00 p.m.–6:00 p.m.: Preparation of
U.S.C. 552b(c)(4)]. ACRS Reports (Open/Closed)—The
For the Nuclear Regulatory Commission.
4:00 p.m.–5:30 p.m.: WCAP–17642P Committee will continue its discussion
Kathryn M. Brock, of proposed ACRS reports. [NOTE: A
Deputy Director, Division of Operating
Westinghouse Performance Analysis
and Design Model (PAD5) (Closed)— portion of this session may be closed in
Reactor Licensing, Office of Nuclear Reactor order to discuss and protect information
Regulation. The Committee will hear briefings by
and hold discussions with designated as proprietary, pursuant to 5
[FR Doc. 2017–13804 Filed 7–3–17; 8:45 am] U.S.C. 552b(c)(4)].
representatives of the NRC staff and
BILLING CODE 7590–01–P
Westinghouse regarding the safety FRIDAY, JULY 14, 2017, CONFERENCE
evaluation associated with the subject ROOM T–2B1, 11545 ROCKVILLE
NUCLEAR REGULATORY topical report. [NOTE: This session will PIKE, ROCKVILLE, MARYLAND 20852
COMMISSION be closed in order to discuss and protect 8:30 a.m.–12:00 p.m.: Preparation of
information designated as proprietary, ACRS Reports (Open/Closed)—The
Advisory Committee on Reactor pursuant to 5 U.S.C 552b(c)(4)]. Committee will continue its discussion
Safeguards; Notice of Meeting 5:30 p.m.–6:00 p.m.: Preparation of of proposed ACRS reports. [NOTE: A
ACRS Reports (Open/Closed)—The portion of this session may be closed in
sradovich on DSK3GMQ082PROD with NOTICES
In accordance with the purposes of Committee will discuss proposed ACRS order to discuss and protect information
Sections 29 and 182b of the Atomic reports on matters discussed during this designated as proprietary, pursuant to 5
Energy Act (42 U.S.C. 2039, 2232b), the meeting. [NOTE: A portion of this U.S.C. 552b(c)(4)].
Advisory Committee on Reactor session may be closed in order to 1:00 p.m.–5:30 p.m.: Preparation of
Safeguards (ACRS) will hold a meeting discuss and protect information ACRS Reports (Open/Closed)—The
July 12–14, 2017, 11545 Rockville Pike, designated as proprietary, pursuant to 5 Committee will continue its discussion
Rockville, Maryland 20852. U.S.C 552b(c)(4)]. of proposed ACRS reports. [NOTE: A
VerDate Sep<11>2014 17:57 Jul 03, 2017 Jkt 241001 PO 00000 Frm 00068 Fmt 4703 Sfmt 4703 E:\FR\FM\05JYN1.SGM 05JYN1](https://thefederalregister.org/doc/82_FR_31216/page/18.svg)
Document Created: 2017-07-04 02:00:55
Document Modified: 2017-07-04 02:00:55
| 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 August 4, 2017. A request for a hearing must be filed by September 5, 2017. | |
| Contact | Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1506, email: [email protected] | |
| FR Citation | 82 FR 31089 |
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