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 Nuclear Regulatory Commission [NRC-2016-0026]
 AGENCY: Nuclear Regulatory Commission.
 ACTION: Biweekly notice.
 SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (AEA), 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 January 16, 2016, to February 1, 2016. The last biweekly notice was published on February 2, 2016.
 DATES: Comments must be filed by March 17, 2016. A request for a hearing must be filed by April 18, 2016.
 ADDRESSES: You may submit comments by any of the following methods (unless 
                        ( print page 7836)
                        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-2016-0026. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: 
                        *Carol.Gallagher@nrc.gov.*
                    Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the 
                        **SUPPLEMENTARY INFORMATION**
                         section of this document.
 FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: 
                        *Paula.Blechman@nrc.gov.*


SUPPLEMENTARY INFORMATION:  I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0026 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods: Federal Rulemaking Web site: Go to*http://www.regulations.gov*
                     and search for Docket ID NRC-2016-0026.
                NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at*http://www.nrc.gov/​reading-rm/​adams.html.*
                     To begin the search, select “
                    *ADAMS Public Documents”*
                     and then select “
                    *Begin Web-based ADAMS Search.”*
                     For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to 
                    *pdr.resource@nrc.gov.*
                     The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the 
                    **SUPPLEMENTARY INFORMATION**
                     section of this document.
                NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2016-0026, facility name, unit number(s), 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 posts all comment submissions at 
                    *http://www.regulations.gov,*
                     as well as entering 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 submissions 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 § 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, (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 should 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. Should the Commission take 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. Should the Commission make 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 person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at 
                    *http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​.*
                     If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's 
                    ( print page 7837)
                    right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii). If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by April 18, 2016. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c). If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by April 18, 2016. B. Electronic Submissions (E-Filing) All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at 
                    *hearing.docket@nrc.gov,*
                     or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on the NRC's public Web site at 
                    *http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html.*
                     System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at 
                    *http://www.nrc.gov/​site-help/​e-submittals.html.*
                     Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software. If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the 
                    ( print page 7838)
                    participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at 
                    *http://www.nrc.gov/​site-help/​e-submittals.html.* Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at 
                    *http://www.nrc.gov/​site-help/​e-submittals.html.*
                     A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System 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 
                    *MSHD.Resource@nrc.gov,*
                     or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 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 requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at 
                    *http://ehd1.nrc.gov/​ehd/​,*
                     unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii). For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document. *Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2 (ANO-2), Pope County, Arkansas* *Date of amendment request:*
                     December 22, 2015. A publicly-available version is in ADAMS under Accession No. ML15356A657. *Description of amendment request:*
                     The amendment would revise the Technical Specifications (TSs) to add a short Allowed Outage Time (AOT) to restore an inoperable system for conditions under which the existing specifications require a plant shutdown. The proposed amendment is consistent with the NRC-approved Technical Specifications Task Force (TSTF) change traveler TSTF-426, Revision 5, “Revise or Add Actions to Preclude Entry into LCO [Limiting Condition for Operation] 3.0.3—RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c.” The availability of TSTF-426, Revision 5, was published in the 
                    **Federal Register**
                     on May 30, 2013 (78 FR 32476). The AOT would be added to specifications governing the pressurizer heaters, containment spray trains, and control room emergency air conditioning and ventilation systems. In addition to the scope of the TSTF-426 TSs revisions, the amendment would add a TS Action to address a single pressurizer proportional heater group having a capacity of less than 150 kilowatts. *Basis for proposed no significant hazards consideration determination:*
                     As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, with NRC staff revisions provided in [brackets], which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change provides a short AOT to restore an inoperable system for conditions under which the existing TSs require a plant shutdown to begin within one hour in accordance with LCO 3.0.3. In addition, a new TS Action associated with Pressurizer proportional heater capacity for a single proportional heater group is proposed. Entering into TS Actions is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not significantly increased. The consequences of any accident previously evaluated that may occur during the proposed AOTs are no different from the 
                        ( print page 7839)
                        consequences of the same accident during the existing one-hour allowance. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new or different accidents result from utilizing the proposed change. The proposed change does not involve a physical alteration of the plant (
                        *i.e.,*
                         no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the proposed change does not impose any new or different requirements. The proposed change does not alter assumptions made in the safety analysis. 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 increases the time the plant may operate without the ability to perform an assumed safety function. The analyses in WCAP-16125-NP-A, “Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent plant Shutdown,” Revision 2, August 2010, demonstrated that there is an acceptably small increase in risk due to a limited period of continued operation in these conditions and that this risk is balanced by avoiding the risks associated with a plant shutdown. As a result, the change to the margin of safety provided by requiring a plant shutdown within one hour is not significant. The new Pressurizer proportional heater capacity Action permits 72 hours to restore the affect heater group to an operable status, consistent with the STS [Standard TSs] and consistent with TS requirements associated with single train inoperabilities. The proportional heaters are not credited in the ANO-2 accident analyses, but aid in Pressurizer pressure control during a loss of offsite power event that results in the need to perform a natural circulation cool down of the plant. The associated STS bases for the standard 72-hour AOT assumes [that] the likelihood of a loss of offsite power event during this time period that would require a demand on the proportional heaters is minimal and acknowledges the use of non-vital powered backup heater groups absent a loss of offsite power event. Note also that under emergency conditions, an Emergency Diesel Generator or the Alternate AC [alternating current] Diesel Generator (
                        *i.e.,*
                         Station Blackout diesel) can be aligned to power any of the non-vital Pressurizer backup heater groups. As a result, the change to the margin of safety provided by the new 72-hour AOT for a single proportional heater train is not significant. 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:*
                     Joseph A. Aluise, Associate General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113. *NRC Branch Chief:*
                     Meena K. Khanna. *Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida* *Date of amendment request:*
                     October 15, 2015. A publicly-available version is in ADAMS under Accession No. ML15301A765. *Description of amendment request:*
                     The amendments would revise the St. Lucie Plant, Unit Nos. 1 and 2, Renewed Facility Operating Licenses' licensing bases to allow the use of the commercially available code “Generation of Thermal-Hydraulic Information for Containments (GOTHIC Version 7.2b(QA)),” to model the containment response following the inadvertent actuation of the containment spray system during normal plant operation (referred to as the vacuum analysis). The amendments would also update the licensing bases to credit the design-basis ability of the containment vessel to withstand a higher external pressure differential of 1.04 pounds per square inch (psi) (1.05 psi for Unit No. 2), and will update Technical Specification 3.6.1.4 for both units to revise the allowable containment operating pressure range. *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. This proposed amendment is related to the analysis of the maximum external pressure that the reactor containment building will experience. A proposed change to the Technical Specifications will limit the allowable external pressure during operation to a value consistent with that considered in the analysis. The analysis is being revised to consider containment spray pump flow higher than previously considered. Containment spray pumps cool and depressurize the containment building; therefore, higher flow impacts the analysis of external pressure on the containment building. The proposed amendment is for the use of a different analysis methodology using the GOTHIC computer code instead of the A-TEMPT and WATEMPT codes that were originally used for the Unit 1 and Unit 2 analyses respectively. The original codes are not currently available. The GOTHIC code is an accepted code for similar analysis. The analysis performed demonstrates that in the postulated event of an inadvertent start of two containment spray pumps, the loading the reactor containment building will experience is within the design of the structure. With this load, the stresses experienced by the reactor containment building remain below the code allowable stresses. The probability of occurrence of an event that would expose the containment building to external pressure is not increased by the change in the analysis methodology used. The probability of the initiating event, inadvertent start of both containment spray pumps, is unchanged. The consequences of an event where the containment building is exposed to external pressure will not be increased as the resulting external pressure on the containment vessel remains within the design, which provides a large margin to the buckling pressure. 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. This proposed amendment changes the methodology for analyzing an event that results in exposing the reactor containment vessel to external pressure. A proposed change to the Technical Specifications will limit the external pressure during operation to a value consistent with the initial condition considered in the analysis. The potential for a new or different kind of accident is not created by the use of a different analysis methodology for a previously defined event. 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. This proposed amendment changes the methodology for analyzing an event that results in exposing the reactor containment building to external pressure. A proposed change to the Technical Specifications will limit the allowable external pressure during operation to a value consistent with the starting point considered in the analysis. The technical evaluation demonstrates that the use of the GOTHIC computer code to determine maximum containment external pressure will result in realistic results similar 
                        ( print page 7840)
                        to the original analysis with the A-TEMPT and WATEMPT codes. The margin of safety in this analysis is maintained by assuring the resulting external pressure acting on the reactor containment vessel maintains significant margin to the buckling pressure in accordance with Section III of the ASME [American Society of Mechanical Engineers] code. For Unit 2, the original code of record limited the maximum external pressure to 
                        1/3
                         of the expected buckling pressure. The analysis of the increased external pressure for Unit 2 has been performed in accordance with the original code of record. The original code of record for Unit 1 was under development at the time and made reference to ASME Section VIII for the analysis of external pressure. The rules of ASME Section VIII at that time limited the maximum external pressure to 
                        1/4
                         of the expected buckling pressure. In order to increase the allowable external pressure, the analysis of external pressure was performed using a later version of the ASME code which allows a maximum external pressure of 
                        1/3
                         of the buckling pressure. The later version of the code used for Unit 1 uses a methodology for determining the maximum external pressure consistent with the code used for Unit 2. Although the margin between the allowable external pressure and the expected buckling pressure for Unit 1 will be changed from a factor of 4 to a factor of 3, substantial margin is maintained in accordance with more current versions of ASME III. The proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:*
                     William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/JB, Juno Beach, FL 33408-0420. *NRC Branch Chief:*
                     Benjamin G. Beasley. *South Carolina Electric and Gas Company Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina* *Date of amendment request:*
                     December 17, 2015. A publicly-available version is in ADAMS under Accession No. ML15351A165. *Description of amendment request:*
                     The proposed change, if approved, would amend Combined License (COL) Nos. NPF-93 and NPF-94 for VCSNS. The requested amendment proposes to rename, relocate, and add radiation detectors to provide monitoring of the radiologically controlled area ventilation system (VAS) exhaust from the radiologically controlled areas of the auxiliary building and annex building. The changes in the proposed amendment are located primarily in the VCSNS Updated Final Safety Analysis Report (UFSAR) Tier 2 information, and involve require conforming changes to COL Appendix C, “Inspections, Tests, Analyses, and Acceptance Criteria,” and departing from certified AP1000 Design Control Document (DCD) Tier 1 information. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 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 design functions of the VAS include prevention of the unmonitored release of airborne radioactivity to the atmosphere or adjacent plant areas by providing monitoring of the VAS exhaust from radiologically controlled areas of the auxiliary building and annex building, and to automatically isolate the selected building areas and start the containment air filtration system (VFS) upon detection of high radioactivity. The proposed changes to the VAS to relocate and add radiation detectors are acceptable as they maintain these design functions. These proposed changes to the VAS design as described in the current licensing basis do not have an adverse effect on any of the design functions of the systems. The proposed changes do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes described create any new accident precursors. 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 revise the VAS design as described in the current licensing basis to enable the system to perform required design functions, and are consistent with other UFSAR information. The proposed changes do not change the design requirements for the system. The relocated and new VAS radiation detectors are designed to the same equipment specifications, including required sensitivity and range, as the existing radiation detectors. The relocated and new VAS radiation detectors monitor the same parameters, as well as perform the same design functions, as the existing radiation detectors. The proposed changes to the system do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes. The proposed changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment. The proposed changes do not allow for a new fission product release path, result in a new fission product barrier failure mode, or 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. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not change the codes or standards for the radiation detectors, or functionality of the ductwork in the auxiliary building and annex building. The proposed changes have no adverse effect on the nonsafety-related system design functions of the VAS for the prevention of the unmonitored release of airborne radioactivity to the atmosphere or adjacent plant areas by providing monitoring of the VAS exhaust from radiologically controlled areas of the auxiliary building and annex building, and to automatically isolate the selected building areas and start the VFS upon detection of high radioactivity. The proposed changes do not affect safety-related equipment or equipment whose failure could initiate an accident. The proposed changes to relocate and add radiation detectors do not adversely interface with safety-related equipment or fission product barriers. Therefore, the proposed changes do not affect any safety-related equipment, design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes, thus, no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:*
                     Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
                    ( print page 7841) *Acting NRC Branch Chief:*
                     John McKirgan. *Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia* *Date of amendment request:*
                     December 22, 2015. A publicly-available version is in ADAMS. under Accession No.ML15356A656. *Description of amendment request:*
                     The proposed change would amend Combined License Nos. NPF-91 and NPF-92 for VEGP, Units 3 and 4, respectively. The requested amendment proposes to depart from approved AP1000 Design Control Documents (DCD) Tier 2 information (text, tables, and figures) and involved Tier 2* information (as incorporated into the Updated Final Safety Analysis Report (UFSAR) as plant specific DCD information), and also involves a change to a license condition. Specifically, the requested amendment proposes changes to the design of auxiliary building Wall 11 and proposes other changes to the licensing basis for use of Seismic Category II structures. *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 do not adversely affect the operation of any systems or equipment inside or outside the auxiliary building that could initiate or mitigate abnormal events, 
                        *e.g.,*
                         accidents, anticipated operational occurrences, earthquakes, floods, tornado missiles, and turbine missiles, or their safety or design analyses, evaluated in the UFSAR. The changes do not adversely affect any design function of the auxiliary building or the systems and equipment contained therein. The ability of the affected auxiliary building [Main Steam Isolation Valve] MSIV compartments to withstand the pressurization effects from the design basis pipe rupture is not adversely affected by the removal of the Wall 11 upper vent openings, because vents at these locations are not credited in the subcompartment pressurization analysis. MSIV compartment temperatures following the limiting one square foot pipe rupture with the vent openings removed remain acceptably within the envelope for environmental qualification of equipment in the compartments. The credit of seismic Category II Wall 11.2 as a [high energy line break] HELB barrier and the seismic Category II turbine building first bay and associated missile barriers to protect Wall 11 openings from tornado missiles continues to provide adequate protection of structures, systems, and components (SSCs) required to safely shut down the plant, as these structures are designed to the same requirements as seismic Category I structures, and with the additional HELB loadings assumed, remain well within the applicable acceptance criteria. Therefore, the proposed amendment does not involve an 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 change the design function of the auxiliary building or of any of the systems or equipment in the auxiliary building or elsewhere within the Nuclear Island structure. These proposed changes do not introduce any new equipment or components that would result in a new failure mode, malfunction or sequence of events that could affect safety-related or nonsafety-related equipment. This activity will not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that would result in significant fuel cladding failures. Therefore, this activity 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 margin of safety for the design of the auxiliary building is maintained through continued use of the current codes and standards as stated in the UFSAR and adherence to the assumptions used in the analyses of this structure and the events associated with this structure. The auxiliary building will continue to maintain a seismic Category I rating which preserves the current structural safety margins. The 3-hour fire rating requirements for the impacted auxiliary building walls are maintained. The Wall 11 upper vents are not credited in the subcompartment pressurization analysis and the remaining vents and pressure relief devices provide sufficient venting to maintain the MSIV compartment pressures below the design limit and design basis. The credit of turbine building Wall 11.2 as a HELB barrier provides protection of Wall 11 from selected dynamic effects, which in turn provides that essential SSCs remain protected from the effects of postulated HELB events. The credit of the seismic Category II turbine building first bay and associated missile barriers to provide protection of Wall 11 openings from tornado missiles provides sufficient protection for the essential SSCs located in the auxiliary building in the vicinity of Wall 11 from the effects of external missiles. Thus, the requested changes will not adversely affect any safety-related equipment, design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change, thus, no margin of safety is reduced. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:*
                     Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015. *Acting NRC Branch Chief:*
                     John McKirgan. *Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia* *Date of amendment request:*
                     December 15, 2015. A publicly-available version is in ADAMS under Accession No. ML15351A023. *Description of amendment request:*
                     The amendments would modify the Technical Specifications (TSs) to risk-inform the requirements regarding selected Required Action end states by incorporating TS Task Force (TSTF) traveler TSTF-423, Revision 1, “Technical Specification End States, NEDC-32988-A.” Additionally, it would modify the TS Required Actions with a Note prohibiting the use of limiting condition for operation 3.0.4.a when entering the preferred end state (Mode 3) on startup. The Notice of Availability for TSTF-423, Revision 1, was published in the 
                    **Federal Register**
                     on February 18, 2011 (76 FR 9614). *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 allows a change to certain required end states when the TS Completion Times for remaining in power operation will be exceeded. Most of the requested technical specification (TS) changes are to permit an end state of hot shutdown (Mode 3) rather than an end state of cold shutdown (Mode 4) contained in the current TS. The request was limited to: (1) 
                        ( print page 7842)
                        Those end states where entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Section 6 of topical report NEDC-32988-A, Revision 2, “Technical Justification to Support Risk Informed Modification to Selected Required Action End States for BWR Plants.” They provide an integrated discussion of deterministic and probabilistic issues, focusing on specific TSs, which are used to support the proposed TS end state and associated restrictions. The NRC staff finds that the risk insights support the conclusions of the specific TS assessments. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting TSTF-423 are no different than the consequences of an accident prior to adopting TSTF-423. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. 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 involve a physical alteration of the plant (no new or different type of equipment will be installed). If risk is assessed and managed, allowing a change to certain required end states when the TS Completion Times for remaining in power operation are exceeded (
                        *i.e.,*
                         entry into hot shutdown rather than cold shutdown to repair equipment) will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in TSTF-IG-05-02, “Implementation Guidance for TSTF-423, Revision 1, `Technical Specifications End States, NEDC-32988-A,' ” will further minimize possible concerns. Thus, based on the above, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The Boiling Water Reactor Owners' Group's risk assessment approach is comprehensive and follows NRC staff guidance as documented in Regulatory Guides (RG) 1.174 and 1.177. In addition, the analyses show that the criteria of the three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based upon the reasoning presented above, SNC concludes that the requested change involves no significant hazards consideration, as set forth in 10 CFR 50.92(c), “Issuance of Amendment.” The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:*
                     Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35242. *NRC Branch Chief:*
                     Michael T. Markley. *Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama* *Date of amendment request:*
                     November 20, 2015, as supplemented by letter dated January 12, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15324A297 and ML16012A457, respectively. *Description of amendment request:*
                     The proposed change would revise the setpoint requirements in Technical Specification (TS) 3.3.5, “Loss of Power Diesel Generator Start Instrumentation.” *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 license amendment request changes the TS 3.3.5 requirements for loss of power diesel generator start instrumentation to enable elimination of manual actions for protection of safety-related equipment from degraded voltage conditions during design basis events. Elimination of these manual actions is required to fulfill an existing License Condition on each unit. The proposed change increases the Allowable Value (AV) for the 4.16 kV Emergency Bus Degraded Grid Voltage Actuation function. Installation of new, higher precision Degraded Voltage Relays (DVRs) makes possible an increase in the DVR actuation setpoint (encompassed by the AV) to a level which provides fully automatic protection of safety-related equipment while minimizing the chance of unwanted disconnection from the preferred offsite power source, which is itself an analyzed condition. Based on the above, 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 license change request changes the TS 3.3.5 requirements for loss of power diesel generator start instrumentation to enable elimination of manual actions for protection of safety-related equipment from degraded voltage conditions during design basis events. Elimination of these manual actions is required to fulfill an existing License Condition on each unit. The proposed changes to TS 3.3.5 do not change the methods of normal plant operation nor the methods of response to transient conditions, save that the range of automatic action provided by the DVRs is expanded. This change will eliminate the need for manual action from the degraded voltage protection scheme, as required by a License Condition for each unit, to achieve compliance with 10 CFR 50.55a(h)(2) and 10 CFR part 50, Appendix A, General Design Criterion 17—Electric Power Systems. Accordingly, 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 provided by the performance capability of plant equipment in preventing or mitigating challenges to fission product barriers under postulated operational transient and accident conditions. Since the proposed license amendment request changes the TS 3.3.5 requirements for loss of power diesel generator start instrumentation to enable elimination of manual actions for protection of safety-related equipment from degraded voltage conditions during design basis events, it will tend to increase the margin of safety by better protecting the safety-related plant equipment. Based on the above, 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.
                    ( print page 7843) *Attorney for licensee:*
                     Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201. *NRC Branch Chief:*
                     Michael T. Markley. *STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499, South Texas Project (STP), Units 1 and 2, Matagorda County, Texas* *Date of amendment request:*
                     June 19, 2013, as supplemented by letters dated October 3, October 31, November 13, November 21, and December 23, 2013 (two letters); January 9, February 13, February 27, March 17, March 18, May 15, May 22, June 25, and July 15, 2014; and March 10, March 25, and August 20, 2015. For the convenience of the reader, the ADAMS accession numbers of the amendment request, supplements, and additional documents (if publicly available) are provided below in a table in the “Availability of Documents” section. *Description of amendment request:*
                     The amendments would revise the Technical Specifications (TSs) and licensing basis for Facility Operating License Nos. NPF-76 and NPF-80, for STP, Units 1 and 2, as documented in the Updated Final Safety Analysis Report (UFSAR). The changes incorporate use of both a deterministic and a risk-informed approach to address safety issues discussed in Generic Safety Issue (GSI)-191, “Assessment of Debris Accumulation on PWR [Pressurized-Water Reactor] Sump Performance,” and to close Generic Letter (GL) 2004-02, “Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors,” dated September 13, 2004 (ADAMS Accession No. ML042360586), for STP. *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 a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the STP licensing basis, changes to the [emergency core cooling system (ECCS)] and [containment spray system (CSS)] TS to extend the required completion time for potential [loss-of-coolant accident (LOCA)] debris related effects and associated administrative TS changes. The methodology change concludes that the ECCS and CSS will have sufficient defense-in-depth and safety margin and will operate with high probability following a LOCA when considering the impacts and effects of debris accumulation on containment emergency sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss of coolant accidents. The methodology change also supports the changes to the TS. There is no significant increase in the probability of an accident previously evaluated. The proposed changes address mitigation of loss of coolant accidents and have no effect on the probability of the occurrence of a loss of coolant accident. The proposed methodology and TS changes do not implement any physical changes to the facility or any [structures, systems, and components (SSCs)], and do not implement any changes in plant operation that could lead to a different kind of accident. The proposed changes do not involve a significant increase in the consequences of an accident previously evaluated. The methodology change confirms that required SSCs supported by the containment sumps will perform their safety functions with a high probability, as required, and does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits. The safety analysis acceptance criteria in the UFSAR continue to be met for the proposed methodology change. The evaluation of the changes determined that containment integrity will be maintained. The dose consequences were considered in the assessment and quantitative evaluation of the effects on dose using input from the risk-informed approach shows the increase in dose consequences is small. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any the accident previously evaluated in the UFSAR. 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 a methodology change for assessment of debris effects from LOCAs that are already evaluated in the STP UFSAR, an extension of TS required completion time for potential LOCA debris related effects on ECCS and CSS, and associated administrative changes to the TS. No new or different kind accident is being evaluated. None of the changes install or remove any plant equipment, or alter the design, physical configuration, or mode of operation of any plant structure, system or component. The proposed changes do not introduce any new failure mechanisms or malfunctions that can initiate an accident. The proposed changes do not introduce failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident. Therefore, the proposed changes do not create the possibility for 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 changes are a methodology change for assessment of debris effects from LOCAs that are already evaluated in the STP UFSAR, an extension of TS required completion time for potential LOCA debris related effects on ECCS and CSS, and associated administrative changes to the TS. The effects from a full spectrum of LOCAs, including double-ended guillotine breaks for all piping sizes up to and including the largest pipe in the reactor coolant system, are analyzed. Appropriate redundancy and consideration of loss of offsite power and worst case single failure are retained, such that defense-in-depth is maintained. Application of the risk-informed methodology showed that the increase in risk from the contribution of debris effects is very small as defined by [NRC Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis”] and that there is adequate defense in depth and safety margin. Consequently, STP determined that the risk-informed method demonstrates the containment sumps will continue to support the ability of safety related components to perform their design functions when the effects of debris are considered. The proposed change does not alter the manner in which safety limits are determined or acceptance criteria associated with a safety limit. The proposed change does not implement any changes to plant operation, and does not significantly affect SSCs that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The proposed change does not significantly affect the existing safety margins in the barriers for the release of radioactivity. There are no changes to any of the safety analyses in the UFSAR. Defense in depth and safety margin was extensively evaluated for the methodology change and the associated TS changes. The evaluation determined that there is substantial defense in depth and safety margin that provide a high level of confidence that the calculated risk for the methodology and TS changes is conservative and that the actual risk is likely much lower. 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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Availability of Documents For further details with respect to this action, see the application for license amendment dated June 19, 2013, listed below in the table, in addition to supplements, requests for additional information responses, and other relevant documents.
                    ( print page 7844) Title
Date
ADAMS Accession No.SECY-12-0093, “Closure Options for Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance.”
07/09/2012
ML121320270STP Pilot Submittal and Request for Exemption for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191
01/31/2013
ML13043A013NRC Letter to STPNOC, “South Texas Project, Units 1 and 2—Supplemental Information Needed for Acceptance of Requested Licensing Action Re: Request for Exemption for a Risk-Informed Approach to Resolve Generic Safety Issue 191”
04/01/2013
ML13066A519Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed approach to Resolving Generic Safety Issue (GSI)-191
06/19/2013
ML131750250 
                             (package)NRC Letter to STPNOC, “South Texas Project, Units 1 and 2—Acceptance of Requests for Exemptions and License Amendment Request for Approval of a Risk-Informed Approach to Resolve Generic Safety Issue GSI-191”
08/13/2013
ML13214A031Corrections to Information Provided in Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Resolving Generic Safety Issue (GSI)-191
10/03/2013
ML13295A222Submittal of GSI-191 Chemical Effects Test Reports
10/31/2013
ML13323A673
                             (package)Supplement 1 to Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Resolving Generic Safety Issue (GSI)-191
11/13/2013
ML13323A128 
                             (package)Supplement 1 to Revised STP Pilot Submittal for a Risk-Informed Approach to Resolving Generic Safety Issue (GSI)-191 to Supersede and Replace the Revised Pilot Submittal
11/21/2013
ML13338A165Response to STP-GSI-191-EMCB-RAI-1
12/23/2013
ML14015A312Response to NRC Request for Reference Document for STP Risk-Informed GSI-191 Application
12/23/2013
ML14015A311Response to Request for Additional Information re Use of RELAP5 in Analyses for Risk-Informed GSI-191 Licensing Application
01/09/2014
ML14029A533Submittal of CASA Grande Code and Analyses for STP's Risk-Informed GSI-191 Licensing Application
02/13/2014
ML14052A110 
                             (package, portions redacted)Submittal of GSI-191 Chemical Effects Test Reports
02/27/2014
ML14072A075 
                             (package)Response to NRC Accident Dose Branch Request for Additional Information Regarding STP Risk-Informed GSI-191 Application
03/17/2014
ML14086A383 
                             (package)Submittal of CASA Grande Source Code for STP's Risk-Informed GSI-191 Licensing Application
03/18/2014
(proprietary, 
                             non-public)Second Submittal of CASA Grande Source Code for STP's Risk-Informed GSI-191 Licensing Application
05/15/2014
ML14149A354First Set of Responses to April, 2014, Requests for Additional Information Regarding STP Risk-Informed GSl-191 Licensing Application—Revised
05/22/2014
ML14149A439 
                             (package)Second Set of Responses to April, 2014, Requests for Additional Information Regarding STP Risk-Informed GSI-191 Licensing Application
06/25/2014
ML14178A467 (package)Third Set of Responses to April, 2014, Requests for Additional Information Regarding STP Risk-Informed GSI-191 Licensing Application
07/15/2014
ML14202A045Submittal of Updated CASA Grande Input for STP's Risk-Informed GSI-191 Licensing Application
03/10/2015
ML15072A092Description of Revised Risk-Informed Methodology and Responses to Round 2 Requests for Additional Information Regarding STP Risk-Informed GSI-191 Licensing Application
03/25/2015
ML15091A440Supplement 2 to STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSI)-191 and Respond to Generic Letter (GL) 2004-02
08/20/2015
ML15246A125
 *Attorney for licensee:*
                     Steve Frantz, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004. *NRC Branch Chief:*
                     Robert J. Pascarelli. *Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant (WBN), Unit 2,*
                     Rhea County, Tennessee *Date of amendment request:*
                     December 15, 2015. A publicly-available version is in ADAMS under Accession No. ML15362A023. *Description of amendment request:*
                     The amendment would revise Technical Specifications (TSs) 3.4.17, “Steam Generator (SG) Tube Integrity”; 5.7.2.12, “Steam Generator (SG) Program”; and 5.9.9, “Steam Generator Tube Inspection Report,” to exclude portions of the SG tubes below the top of the tube sheet from needing to be plugged. *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. Allowing the use of an alternate repair criteria as proposed in this amendment request does not involve a significant increase in the probability or consequence of an accident previously evaluated. The presence of the tubesheet enhances the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter. The resistance to both tube rupture and tube collapse is strengthened by the presence of the tubesheet in that region. Hardrolling of the tube into the tubesheet results in an interference fit between the tube and the tubesheet. Tube rupture cannot occur because the contact between the tube and tubesheet does not permit sufficient movement of tube material. In a similar manner, the tubesheet does not permit sufficient movement of tube material to permit buckling collapse of the tube during postulated loss-of-coolant-accident (LOCA) loadings. The type of degradation for which the F* [the length of mechanical expansion required to prevent pullout for all normal operating and postulated accident conditions] has been developed (cracking with a circumferential orientation) can theoretically lead to a postulated tube rupture event, provided that the postulated through-wall circumferential crack exists near the top of the tubesheet. An evaluation including analysis and testing has 
                        ( print page 7845)
                        been performed to determine the resistive strength of roll expanded tubes within the tubesheet. That evaluation provides the basis for the acceptance criteria for tube degradation subject to the F* criterion. The F* length of roll expansion is sufficient to preclude tube pullout from tube degradation located below the F* distance, regardless of the extent of the tube degradation. The existing technical specification leakage rate requirements and accident analysis assumptions remain unchanged in the unlikely event that significant leakage from this region does occur. As noted above, tube rupture and pullout are not expected for tubes using the ARC [alternative repair criterion]. Any leakage out of the tube from within the tubesheet at any elevation in the tubesheet is fully bounded by the existing Main Steam Line Break (MSLB) analysis included in the WBN Unit 2 Final Safety Analysis Report (FSAR). Therefore, the proposed ARC does not adversely impact any other previously evaluated design basis accident. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Implementation of the proposed ARC does not introduce any significant changes to the plant design basis. Use of the criterion does not provide a mechanism to result in an accident initiated outside of the region of the tubesheet expansion. A hypothetical accident as a result of any tube degradation in the expanded portion of the tube would be bounded by the existing tube rupture accident analysis. Tube bundle structural integrity and leak tightness are expected to be maintained. 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 use of the ARC has been demonstrated to maintain the integrity of the tube bundle commensurate with the requirements of Regulatory Guide 1.121, “Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes,” for indications in the free span of tubes and the primary to secondary pressure boundary under normal and postulated accident conditions. Acceptable tube degradation for the F* criterion is any degradation indication in the tubesheet region, more than the F* distance below either the bottom of the transition between the roll expansion and the unexpanded tube, or the top of the tubesheet, whichever is lower. The safety factors used in the verification of the strength of the degraded tube are consistent with the safety factors in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code used in SG design. The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude both tube pullout and significant leakage during normal and postulated accident conditions. Resistance to tube pullout is based upon the primary to secondary pressure differential as it acts on the surface area of the tube, which includes the tube wall cross-section, in addition to the inside diameter-based area of the tube. The leak testing acceptance criteria are based on the primary to secondary leakage limit in the technical specifications and the leakage assumptions used in the UFSAR [Updated FSAR] accident analyses. Implementation of the ARC will decrease the number of tubes which must be taken out of service with tube plugs. Plugs reduce the RCS flow margin; thus, implementation of the ARC will maintain the margin of flow that would otherwise be reduced in the event of increased plugging. Based on the above, it is concluded that the proposed change does not result in a significant reduction in or a loss of margin with respect to plant safety as defined in the FSAR or the bases of the WBN Unit 2 technical specifications. 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:*
                     Ralph E. Rodgers, General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., 6A West Tower, Knoxville, TN 37902. *NRC Branch Chief:*
                     Benjamin G. Beasley. III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (AEA), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the 
                    **Federal Register**
                     as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document. *Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit No. 2 (MPS2) and Unit No. 3 (MPS3), New London County, Connecticut* *Date of amendment request:*
                     January 15, 2015, as supplemented by letters dated April 15, July 16, July 30, November 2, and December 1, 2015. *Brief description of amendment:*
                     The amendments revised the MPS2 and MPS3 Technical Specifications (TSs) to adopt NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.” *Date of issuance:*
                     January 29, 2016. *Effective date:*
                     As of the date of issuance and shall be implemented within 90 days from the date of issuance. *Amendment Nos.:*
                     325 and 267. A publicly-available version is in ADAMS under Accession No. ML16011A400; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. *Renewed Facility Operating License Nos. DPR-65 and NPF-49:*
                     Amendments revised the Renewed Operating License and TSs. *Date of initial notice in**Federal Register*
                    : July 21, 2015 (80 FR 43126). The supplemental letter dated April 15, 2015, was published with the January 15, 2015, application, in the initial FR notice. The supplemental letters dated July 16, July 30, November 2, and December 1, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the 
                    **Federal Register**
                    . The Commission's related evaluation of the amendment is contained in a 
                    ( print page 7846)
                    Safety Evaluation dated January 29, 2016. *No significant hazards consideration comments received:*
                     No. *Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania* *Date of amendment request:*
                     February 2, 2015, as supplemented by letters dated August 11, 2015, and October 20, 2015. *Brief description of amendments:*
                     The amendments modified the technical specifications (TSs) to allow for brief, inadvertent, simultaneous opening of redundant secondary containment personnel access doors during normal entry and exit conditions. *Date of issuance:*
                     January 28, 2016. *Effective date:*
                     As of the date of issuance and shall be implemented within 60 days of issuance. *Amendment Nos.:*
                     220 and 182. A publicly-available version is in ADAMS under Accession No. ML15356A140; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. *Renewed Facility Operating License Nos. NPF-39 and NPF-85:*
                     Amendments revised the Renewed Facility Operating Licenses and TSs. *Date of initial notice in**Federal Register*
                    : April 14, 2015 (80 FR 20022). The supplemental letters dated August 11, 2015, and October 20, 2015, 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 January 28, 2016. *No significant hazards consideration comments received:*
                     Yes. *Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania* *Date of amendment request:*
                     February 23, 2015, as supplemented by letters dated August 12, 2015, and October 20, 2015. *Brief description of amendments:*
                     The amendments modified the technical specifications (TSs) to allow for brief, inadvertent, simultaneous opening of redundant secondary containment personnel access doors during normal entry and exit conditions. *Date of issuance:*
                     February 1, 2016. *Effective date:*
                     As of the date of issuance and shall be implemented within 60 days of issuance. *Amendments Nos.:*
                     303 and 307. A publicly-available version is in ADAMS under Accession No. ML15350A179; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. *Renewed Facility Operating License Nos. DPR-44 and DPR-56:*
                     The amendments revised the Renewed Facility Operating Licenses and the TSs. *Date of initial notice in**Federal Register*
                    : April 14, 2015 (80 FR 20023). The supplemental letters dated August 12, 2015, and October 20, 2015, 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 February 1, 2016. *No significant hazards consideration comments received:*
                     Yes. *Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska* *Date of amendment request:*
                     January 15, 2015, as supplemented by letters dated May 4, 2015, June 9, 2015, and January 12, 2016. *Brief description of amendment:*
                     The amendment revised the technical specifications (TSs) to add a limiting condition for operation, applicability, required actions, completion times, and surveillance requirements for the residual heat removal containment spray and associated interlock permissive instrumentation. A new TS Section 3.6.1.9, “Residual Heat Removal (RHR) Containment Spray,” has been added to reflect the reliance on containment spray to maintain the drywell within design temperature limits during a small steam line break. In addition, the “Drywell Pressure—High” function that serves as an interlock permissive to allow RHR containment spray mode alignment has been relocated from the Technical Requirements Manual to TS 3.3.5.1, “Emergency Core Cooling System (ECCS) Instrumentation.” *Date of issuance:*
                     January 22, 2016. *Effective date:*
                     As of the date of issuance and shall be implemented within 60 days of issuance. *Amendment No.:*
                     253. A publicly-available version is in ADAMS under Accession No. ML15343A301; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. *Renewed Facility Operating License No. DPR-46:*
                     The amendment revised the Facility Operating License and Technical Specifications. *Date of initial notice in**Federal Register*
                    : March 17, 2015 (80 FR 13910). The supplemental letters dated May 4, 2015, June 9, 2015, and January 12, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the 
                    **Federal Register**
                    . The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 22, 2016. *No significant hazards consideration comments received:*
                     No. *South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina* *Date of amendment request:*
                     November 26, 2013, as supplemented by letter dated June 3, 2015. *Brief description of amendment:*
                     The amendments are to Combined License Nos. NPF-93 and NPF-94 for VCSNS, Units 2 and 3. The amendments authorized changes to the VCSNS, Units 2 and 3, Updated Final Safety Analysis Report to revise the details of the effective thermal conductivity resulting from the oxidation of the inorganic zinc component of the containment vessel coating system. *Date of issuance:*
                     October 9, 2015. *Effective date:*
                     As of the date of issuance and shall be implemented within 90 days of issuance. *Amendment No.:*
                     34. A publicly-available version is in ADAMS under Accession No. ML15272A417; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. *Facility Combined Licenses No. NPF-93 and NPF-94:*
                     Amendments revised the Facility Combined Licenses. *Date of initial notice in**Federal Register*
                    : February 19, 2014 (79 FR 9490). The supplemental letter dated June 3, 2015, 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**
                    .
                    ( print page 7847) The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 9, 2015. *No significant hazards consideration comments received:*
                     No. *Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear Plant (HNP), Unit No. 1, Appling County, Georgia* *Date of application for amendment:*
                     September 1, 2015. *Brief description of amendments:*
                     The amendment revised the Technical Specification value of the Safety Limit Minimum Critical Power Ratio to support operation in the next fuel cycle. *Date of issuance:*
                     January 29, 2016. *Effective date:*
                     As of the date of issuance and shall be implemented prior to reactor startup following the HNP, Unit 1, spring 2016, refueling outage. *Amendment No.:*
                     275. A publicly-available version is in ADAMS under Accession No. ML15342A398; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. *Renewed Facility Operating License No. DPR-57:*
                     Amendment revised the license and the Technical Specifications. *Date of initial notice in**Federal Register*
                    : November 3, 2015 (80 FR 67802). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 29, 2016. *No significant hazards consideration comments received:*
                     No. *Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri* *Date of application for amendment:*
                     April 29, 2015. *Brief description of amendment:*
                     The amendment revised the Cyber Security Plan Implementation Milestone 8 completion date and the physical protection license condition. *Date of issuance:*
                     January 28, 2016. *Effective date:*
                     As of its date of issuance and shall be implemented within 90 days from the date of issuance. *Amendment No.:*
                     214. A publicly-available version is in ADAMS under Accession No. ML15328A059; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. *Renewed Facility Operating License No. NPF-30:*
                     The amendment revised the Operating License. *Date of initial notice in**Federal Register*
                    : July 7, 2015 (80 FR 38778). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2016. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 8th day of February 2016. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.


[FR Doc. 2016-02916 Filed 2-12-16; 8:45 am] BILLING CODE 7590-01-P