# Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3A2015-07192 ## Record - **Collection:** Federal Register - **Document type:** Notice - **Published:** March 31, 2015 - **Citation:** 80 FR 17083 ## Text NUCLEAR REGULATORY COMMISSION [NRC-2015-0073] Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission. ACTION: Biweekly notice. SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 5, 2015 to March 18, 2015. The last biweekly notice was published on March 17, 2015. DATES: Comments must be filed by April 30, 2015. A request for a hearing must be filed by June 1, 2015. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject): • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0073. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. • Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2015-0073 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-2015-0073. • 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. • 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-2015-0073, 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, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period 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 by the above date, 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 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 identify 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. If a hearing is requested, 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. 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 ten 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 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, 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. Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: November 24, 2014. A publicly- available version is in ADAMS under Accession No. ML14330A327. Description of amendment request: The proposed amendments would modify the Technical Specifications (TS) to correct non-conservative setpoints. Specifically, modify the Allowable Value parameter and the Nominal Trip Setpoint for the TS 3.3.2 Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation” function for Auxiliary Feedwater Loss of Offsite Power (Function 6.d.) and for the TS 3.3.5 Loss of Voltage function in Surveillance Requirement (SR) 3.3.5.2. As part of the change, the licensee is also proposing to add the applicable footnotes in accordance with TSTF-493, Revision 4, “Clarify Application of Setpoint Methodology for LSSS [limiting safety system set point] Functions.” 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 and staff's changes/additions are provided in [ ]: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Duke Energy requests NRC review and approval to revise the Allowable Value parameter for the Technical Specification (TS) 3.3.2 Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation” function for Auxiliary Feedwater Loss of Offsite Power (Function 6.d.) and for the TS 3.3.5 Loss of Voltage function in Surveillance Requirement (SR) 3.3.5.2 in order to make this parameter more restrictive. The existing parameter was determined to be non-conservative and this parameter is presently classified as Operable But Degraded in the Catawba Corrective Action Program. In addition, the Nominal Trip Setpoint parameter for this function is being slightly lowered in order to gain additional margin. Finally, as part of this License Amendment Request (LAR), applicable footnotes are also being added to the affected TS 3.3.2 function in accordance with TS Task Force Traveler [(TSTF)] TSTF-493, Revision 4, “Clarify Application of Setpoint Methodology for LSSS Functions.” The more restrictive Allowable Value will preclude the potential for a double sequencing event to occur under the condition of a Loss of Coolant Accident (LOCA) load sequencer actuation with a pre-existing degraded voltage condition on the essential buses. These proposed changes will not increase the probability of occurrence of any design basis accident since the affected function, in and of itself, cannot initiate an accident. Should a LOCA occur, the proposed changes will ensure that the sequencer operates properly in order to mitigate the consequences of the event. Appropriate calculations were developed to substantiate the revised TS parameters proposed in this LAR. There will be no impact on the source term or pathways assumed in accidents previously evaluated. No analysis assumptions will be violated and there will be no adverse effects on onsite or offsite doses as the result of an accident. Adoption of the TSTF-493 footnotes for the respective SRs will ensure that the function's channels will continue to behave in accordance with safety analysis assumptions and the channel performance assumptions in the setpoint methodology. Therefore, the proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendments do not change the methods governing normal plant operation; nor are the methods utilized to respond to plant transients altered. In addition, the proposed changes to the affected TS parameters and the adoption of the TSTF-493 footnotes will not create the potential for any new initiating events or transients to occur in the actual physical plant. Therefore, the proposed amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed changes will assure the acceptable operation of the affected function under all postulated transient and accident conditions. This will ensure that all applicable design and safety limits are satisfied such that the fission product barriers will continue to perform their design functions. Therefore, the proposed amendments do not involve a significant reduction in a margin of safety. Based on the preceding discussion, Duke Energy concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202. NRC Branch Chief: Robert J. Pascarelli. Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: March 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14078A037. Description of amendment request: The amendment would revise the Technical Specifications (TS) for the Inservice Testing Program to reflect the current edition of the American Society of Mechanical Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b). 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 corrects a typographical error in TS 5.5.8, “Reactor Coolant Pump Flywheel Inspection Program,” and revises TS 5.5.9, “lnservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. 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 corrects a typographical error in TS 5.5.8, “Reactor Coolant Pump Flywheel Inspection Program,” and revises TS 5.5.9, “lnservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed change does not involve a modification to the physical configuration of the plant ( i.e., no new equipment will be installed), nor does it involve a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change corrects a typographical error in TS 5.5.8, “Reactor Coolant Pump Flywheel Inspection Program,” and revises TS 5.5.9, “lnservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. 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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street—DEC45A, Charlotte, NC 28202-1802. NRC Branch Chief: Robert J. Pascarelli. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: November 21, 2014. A publicly-available version is in ADAMS under Accession No. ML14325A520. Description of amendment request: The amendment would change the GGNS Technical Specification (TS) 2.1.1, “Reactor Core SLs [Safety Limits].” Specifically, the change would revise the Minimum Critical Power Ratio (MCPR) SL stated in TS 2.1.1.2 for two-loop operation from greater than or equal to (≥) 1.11 to ≥ 1.15. Additionally, the change would revise the MCPR SL stated in TS 2.1.1.2 for single-loop operation from ≥ 1.14 to ≥ 1.15. 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 Bases to TS 2.1.1.2 states that: “The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO [Anticipated Operational Occurrence] from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. This condition is met in that the GGNS Cycle 20 (C20) MCPR SL evaluation was performed in accordance with Reference 4 [NEDE-24011-P-A, “General Electric Standard Application for Reactor Fuel (GESTAR-II”)]. The resulting values continue to ensure the conservatism described in the Bases to TS 2.1.1.2. The proposed changes also continue to ensure sufficient conservatism in the operating MCPR limit. The MCPR operating limits are presented and controlled in accordance with the GGNS Core Operating Limits Report (COLR). The requested Technical Specification change does not involve any plant modifications or operational changes that could affect system reliability or performance or that could affect the probability of operator error. The requested change does not affect any postulated accident precursors, any accident mitigating systems, or introduce any new accident initiation mechanisms. Therefore, the proposed change to increase the MCPR SL values 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 any new modes of operation, any changes to setpoints, or any plant modifications. The proposed change to the MCPR SL accounts for requirements specified in the NRC Safety Evaluation limitations and conditions associated with NEDC-33173P [“Applicability of GE Methods to Expanded Operating Domains”] and NEDC-33006P [“Licensing Topical Report—General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus”]. Compliance with the criterion for incipient boiling transition continues to be ensured. The core operating limits will continue to be developed using NRC approved methods. The proposed [MCPR SL] does not result in the creation of any new precursors to an accident. Therefore, the proposed change does not create of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The MCPR SLs have been evaluated in accordance with Global Nuclear Fuels NRC-approved cycle-specific safety limit methodology to ensure that during normal operation and during AOO's, at least 99.9% of the fuel rods in the core are not expected to experience transition boiling. The proposed change to the [MCPR SL] accounts for requirements specified in the NRC Safety Evaluation limitations and conditions associated with NEDC-33173P and NEDC-33006P, which result in additional margin above that specified in the TS Bases. Therefore, the proposed change to the MCPR SL 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. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: November 21, 2014, as supplemented by letter dated February 18, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML14325A752 and ML15049A536, respectively. Description of amendment request: The proposed amendment would revise GGNS's license basis to adopt a single fluence methodology. 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 adopts a single flux methodology. While Chapter 15, Accident Analysis, of the Standard Review Plan (NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants) assumes the pressure vessel does not fail, the flux methodology is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the adoption of the flux methodology has no effect on the probability of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change adopts a flux methodology. The change does not involve a physical alteration of the plant ( i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. The change does not alter assumptions made in the safety analysis regarding fluence. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change adopts a single fluence methodology. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the methodology used for fluence is in compliance with RG 1.190 requirements. 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. Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Date of amendment request: August 19, 2014. A publicly-available version is in ADAMS under Accession No. ML14231A902. Description of amendment request: The proposed amendment would increase the technical specification (TS) surveillance requirement (SR) 3.7.9.2 allowable temperature to less than or equal to 102 °F [degree Fahrenheit]. 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 likelihood of a malfunction of any systems, structures or components (SSCs) supported by the UHS [ultimate heat sink] is not significantly increased by increasing the allowable Ultimate Heat Sink (UHS) temperature from ≤100 °F to ≤102 °F. The UHS provides a heat sink for process and operating heat from safety related components during a transient or accident, as well as during normal operation. The proposed change does not make any physical changes to any plant SSCs, nor does it alter any of the assumptions or conditions upon which the UHS is designed. The UHS is not an initiator of any analyzed accident. All equipment supported by the UHS has been evaluated to demonstrate that their performance and operation remains as described in the UFSAR [updated final safety analysis report] with no increase in probability of failure or malfunction. The SSCs credited to mitigate the consequences of postulated design basis accidents remain capable of performing their design basis function. The change in maximum UHS temperature has been evaluated using the UFSAR described methods to demonstrate that the UHS remains capable of removing normal operating and post-accident heat. The change in UHS temperature and resulting containment response following a postulated design basis accident has been demonstrated to not be impacted. Additionally, all the UHS supported equipment, credited in the accident analysis to mitigate an accident, has been shown to continue to perform their design function as described in the UFSAR. 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 create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not introduce any new modes of plant operation, change the design function of any SSC, change the mode of operation of any SSC, or change any actions required when the TS limit is exceeded. There are no new equipment failure modes or malfunctions created as affected SSCs continue to operate in the same manner as previously evaluated and have been evaluated to perform as designed at the increased UHS temperature and as assumed in the accident analysis. Additionally, accident initiators remain as described in the UFSAR and no new accident initiators are postulated as a result of the increase in UHS temperature. 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 continues to ensure that the maximum temperature of the cooling water supplied to the plant SSCs during a UHS design basis event remains within the evaluated equipment limits and capabilities assumed in the accident analysis. The proposed change does not result in any changes to plant equipment function, including setpoints and actuations. All equipment will function as designed in the plant safety analysis without any physical modifications. The proposed change does not alter a limiting condition for operation, limiting safety system setting, or safety limit specified in the Technical Specifications. The proposed change does not adversely impact the UHS inventory required to be available for the UFSAR described design basis accident involving the worst case 30-day period including losses for evaporation and seepage to support safe shutdown and cooldown of both Braidwood Station units. Additionally, the structural integrity of the UHS is not impacted and remains acceptable following the change, thereby ensuring that the assumptions for both UHS temperature and inventory remain valid. Therefore, since there is no adverse impact of this change on the Braidwood Station safety analysis, there is no reduction in the margin of safety of the plant. 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 requested amendments involve no significant hazards consideration. Attorney for licensee: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Travis L. Tate Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle County, Illinois Date of amendment request: November 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14328A800. Description of amendment request: The proposed amendment would revise Condition I and surveillance requirement (SR) 3.7.9.3 associated with technical specification (TS) Section 3.7.9, “Ultimate Heat Sink (UHS),” to reflect the current design basis flood level. 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: EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92(c), “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to revise TS 3.7.9, Condition I and SR 3.7.9.3 will ensure the operability of the SX [service water] makeup pumps to meet TS 3.7.9 LCO [Limiting Condition for Operation] requirement. The proposed change does not result in any physical changes to safety related structures, systems, or components. The probability of a flood at the river screen house (RSH) is unchanged. Since the UHS itself is not an accident initiator, the proposed change does not impact the initiators or assumptions of analyzed accidents, nor do they impact the mitigation of accidents or transient events. Consequently, the proposed change does not increase the probability of occurrence for any accident previously evaluated. The proposed change will ensure that actions to verify operability of the deep well pumps will be taken prior to the potential for the SX makeup pumps to be adversely affected by the combined event flood high river level. Therefore, the UHS will be capable of performing its functions to mitigate accidents by serving as the heat sink for safety related equipment. Thus, the proposed change does not increase the consequences of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to revise TS 3.7.9, Condition I and SR 3.7.9.3 does not change the design function or operation of the SX makeup pumps. The proposed change does not change or introduce the possibility of any new or different type of equipment, modes of system operation, failure mechanisms, malfunctions, or accident initiators. The proposed change to lower the river level value at which action is taken to verify basin levels and deep well pumps are ready to perform the UHS makeup function in the place of the SX makeup pumps will not affect the operation or function of the UHS or the deep well pumps. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to revise TS 3.7.9, Condition I and SR 3.7.9.3 reestablishes the margin between the design bases combined event flood level and TS 3.7.9, Condition I action level for high river level. The proposed change will ensure the operability of the SX makeup pumps to meet TS 3.7.9 LCO and do not affect the ability of the SX makeup pumps to provide the safety related source makeup to the UHS. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified. 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 requested amendments involve no significant hazards consideration. Attorney for licensee: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: December 22, 2014. A publicly-available version is in ADAMS under Accession No. ML14357A085. Description of amendment request: The proposed amendment modifies the technical specifications (TSs) to add a new Limiting Condition for Operation (LCO) 3.10.8 to specifically permit inservice leakage and hydrostatic testing at reactor coolant system (RCS) temperatures greater than the average reactor coolant temperature for MODE 4 with the reactor shutdown. In addition, the proposed amendment includes an expanded scope of LCO 3.10.8 consistent with the NRC-approved Revision 0 of Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-484, “Use of TS 3.10.1 for Scram Time Testing Activities” available in ADAMS under Accession No. ML062990425. 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: EGC [Exelon Generation Company] has evaluated the proposed changes, using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration. 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes will not result in a significant change in the stored energy in the reactor vessel during the performance of the testing. The probability of an accident is not significantly increased because the proposed changes will not alter the method by which inservice leakage and hydrostatic testing is performed or significantly change the temperatures and pressures achieved to perform the test. The consequences of previously evaluated accidents are not significantly increased because the required testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main system line break outside of primary containment. Under these proposed changes, the secondary containment, standby gas treatment system, and associated initiation instrumentation are required to be operable during the performance of inservice leakage and hydrostatic testing and would be capable of mitigating any airborne radioactivity or steam leaks that could occur. In addition, the required Emergency Core Cooling subsystems will be more than adequate to ensure that a significant increase in consequences will not occur by ensuring that the potential for failed fuel and a subsequent increase in coolant activity above Technical Specification limits are minimized. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. As the accumulated neutron fluence on the reactor vessel increases, the Pressure-Temperature Limits in TS 3.4.9 for DNPS [Dresden Nuclear Power Station] and QCNPS [Quad Cities Nuclear Power Station and TS [technical specification] 3.4.11 for LSCS [LaSalle County Station] may eventually require that inservice leakage and hydrostatic testing be conducted at RCS [reactor coolant system] temperatures greater than the average reactor coolant temperature for MODE 4 with the reactor shutdown. However, even with the required minimum reactor coolant temperatures less than or equal to the average reactor coolant temperature for MODE 4 with the reactor shutdown, maintaining RCS temperatures within a small band during testing can be impractical. The proposed changes will not result in a significant change in the stored energy in the reactor vessel during the performance of the testing nor will it alter the way inservice leakage and hydrostatic testing is performed or significantly change the temperatures and pressures achieved to perform the testing. Therefore, the proposed changes do 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 changes and additions result in increased system operability requirements above those that currently exist during the performance of inservice leakage and hydrostatic testing. The incremental increase in stored energy in the vessel during testing will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment and analyzed margins of safety are unchanged. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. EGC has reviewed the no significant hazards determination published on August 21, 2006 (71 FR 48561) [for Technical Specification Task Force traveler TSTF-484]. The no significant hazards determination was made available on October 27, 2006 (71 FR 63050) as part of the CLIIP [Consolidated Line Item Improvement Process] Notice of Availability. EGC has concluded that the determination presented in the notice is applicable to DNPS, Units 2 and 3; LSCS, Units 1 and 2; and QCNPS, Units 1 and 2; and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a). 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 requested amendments involve no significant hazards consideration. Attorney for licensee: Bradley Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Travis L. Tate. Exelon Generation Company, LLC, Docket No. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of Amendment Request: January 12, 2015. A publicly-available version is in ADAMS under Accession No. ML15012A544. Description of amendment request: The proposed amendment would delete the limiting condition for operation (LCO) Note for Technical Specification (TS) Section 3.5.1, “ECCS [emergency core cooling system]—Operating.” The current Note allows the licensee to consider the low pressure coolant injection (LPCI) subsystem associated with the residual heat removal (RHR) system to be OPERABLE under specified conditions. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. No physical changes to the facility will occur as a result of this proposed amendment. The proposed change will not alter the physical design. Current TS note could make LSCS susceptible to potential water hammer in the RHR system if in the SDC [shutdown cooling] Mode of RHR in Mode 3 when swapping from the SDC to LPCI mode of RHR. The proposed LAR [license amendment request] will eliminate the risk for cavitation of the pump and voiding in the suction piping, thereby avoiding potential to damage the RHR system, including water hammer. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function. Deletion of the TS note is appropriate because current TSs could put the plant at risk for potential cavitation of the pump and voiding in the suction piping, resulting in potential to damage the RHR system, including water hammer. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change conforms to NRC regulatory guidance regarding the content of plant Technical Specifications. The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above evaluation, EGC [Exelon Generation Company, LLC] concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, according a finding of no significant hazards consideration is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Tamra Domeyer, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL, 60555. Branch Chief: Travis L. Tate. FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio Date of amendment request: December 31, 2014. A publicly-available version is in ADAMS under Accession No. ML14365A080. Description of amendment request: The proposed amendment would revise the frequency for the technical specification surveillance to verify that each containment spray system nozzle is unobstructed from a frequency of 10 years to an event-based frequency. 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 containment spray system and its spray nozzles are not accident initiators and therefore the proposed change does not involve a significant increase in the probability of an accident. The revised surveillance requirement will require event-based frequency verification in lieu of a fixed frequency verification. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that may initiate an analyzed event. The proposed change will not alter the operation or otherwise increase the failure probability of any plant equipment that can initiate an analyzed accident. Because the system will continue to be available to perform its accident mitigation function, the consequences of accidents 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. The proposed change will not physically alter the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. The proposed change does not introduce new accident initiators or impact assumptions made in the safety analysis. Testing requirements continue to demonstrate that the limiting conditions for operation are met and the system components are functional. 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 safety function of the CSS [containment spray system] is to spray water into the containment atmosphere in the event of a loss-of-coolant accident to prevent containment pressure from exceeding the design value and to remove fission products from the containment atmosphere. The CSS is not susceptible to corrosion-induced obstruction or obstruction from sources external to the system. Maintenance activities that unexpectedly introduce unretrievable foreign material into the system would require subsequent verification to ensure there is no nozzle blockage. The spray header nozzles are expected to remain unblocked and available in the event that a safety function is required. Therefore, the capacity of the system would remain unaffected. The proposed change does not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this change. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Travis L. Tate. FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio Date of amendment request: December 19, 2014. A publicly-available version is in ADAMS under Accession No. ML14353A349. Description of amendment request: The proposed amendment would revise the technical specifications (TS) to adopt performance-based Type C testing for the reactor containment, which would allow for extended test intervals for Type C valves up to 75 months, and corrects an editorial issue in the TS. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment adopts the NRG-accepted guidelines of [Nuclear Energy Institute] NEI 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” for [Davis-Besse Nuclear Power Station] DBNPS performance-based Type C containment isolation valve testing. Revision 3-A of NEI 94-01 allows, based on previous valve leak test performance, an extension of Type C containment isolation valve leak test intervals. Since the change involves only performance-based Type C testing, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the components of the primary containment system will limit leakage rates to less than the values assumed in the plant safety analyses. The proposed amendment will not change the leakage rate acceptance requirements. As such, the containment will continue to perform its design function as a barrier to fission product releases. Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to revise the extended frequency performance-based Type C testing program does not change the design or operation of structures, systems, or components of the plant. The proposed amendment would continue to ensure containment operability and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by the proposed amendment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment to revise the extended frequency performance-based Type C testing program does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation. The specific requirements and conditions of the Technical Specification Containment Leakage Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by Technical Specifications is maintained, thus ensuring the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods, and acceptance criteria for Type A, Type B, and Type C containment leakage tests specified in applicable codes and standards would continue to be met with the acceptance of this proposed change, since these are not affected by this revision to the performance-based containment testing program. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Travis L. Tate. Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: November 14, 2014, as supplemented by a letter dated February 12, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML14324A209, and ML15050A247, respectively.) Description of amendment request: The proposed amendments would replace the current Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 technical specifications (TSs) limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new DOSE EQUIVALENT XE-133 definition that would replace the current E-Bar average disintegration energy definition. In addition, the current DOSE EQUIVALENT I-131 definition would be revised to allow the use of additional thyroid dose conversion factors. The proposed RCS specific activity changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification change traveler, TSTF-490, Revision 0, “Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification,” with deviations. Additionally, the proposed amendments would revise the CNP Units 1 and 2 licensing basis and TSs to adopt the alternative source term (AST) as allowed in 10 CFR 50.67. The proposed amendments represent full implementation of the AST as described in the NRC's Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” Revision 0. 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. The licensee concluded that the no significant hazards consideration determination published on March 19, 2007 (72 FR 12838), “Notice of Availability of the Model Safety Evaluation,” is applicable. This determination is presented below, along with the licensee's analysis of the implementation of the AST. Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated. Reactor coolant specific activity is not an initiator for any accident previously evaluated. The Completion Time when primary coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit primary coolant noble gases to concentrations consistent with the accident analyses. The proposed change to the Completion Time has no impact on the consequences of any design basis accident since the consequences of an accident during the extended Completion Time are the same as the consequences of an accident during the Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased. There are no physical changes to the plant being introduced by the proposed changes to the accident source term. Implementation of AST and the associated proposed TS changes and new atmospheric dispersion factors have no impact on the probability for initiation of any DBAs [Design Basis Accidents]. Once the occurrence of an accident has been postulated, the new accident source term and atmospheric dispersion factors are an input to analyses that evaluate the radiological consequences. The proposed changes do not involve a revision to the design or manner in which the facility is operated that could increase the probability of an accident previously evaluated in Chapter 14 of the UFSAR. Based on the AST analyses, there are no proposed changes to performance requirements and no proposed revision to the parameters or conditions that could contribute to the initiation of an accident previously discussed in Chapter 14 of the UFSAR. Plant-specific radiological analyses have been performed using the AST methodology and new X/Qs have been established. Based on the results of these analyses, it has been demonstrated that the CR [control room] and off-site dose consequences of the limiting events considered in the analyses meet the regulatory guidance provided for use with the AST, and the doses are within the limits established by 10 CFR 50.67. Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or the consequences of an accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated. The proposed change in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter. The change does not create the potential for a new or different kind of accident from any previously calculated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Implementation of AST and the associated proposed TS changes and new X/Qs have no impact to the initiation of any DBAs. These changes do not affect the design function or modes of operation of structures, systems and components in the facility prior to a postulated accident. Since structures, systems and components are operated no differently after the AST implementation, no new failure modes are created by this proposed change. The alternative source term change itself does not have the capability to initiate accidents. Consequently, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety. The proposed change revises the limits on noble gas radioactivity in the primary coolant. The proposed change is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses. The AST analyses have been performed using approved methodologies to ensure that analyzed events are bounding and safety margin has not been reduced. Also, new X/Qs, which are based on site specific meteorological data, were calculated in accordance with the guidance of RG 1.194 to utilize more recent data and improved calculational methodologies. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67. Thus, by meeting the applicable regulatory limits for AST, there is no significant reduction in a margin of safety. Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the proposed amendment does not involve a significant reduction in margin of safety. The NRC staff has reviewed the 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 amendments requested involve no significant hazards consideration. Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: David L. Pelton. Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, Somervell County, Texas Date of amendment request: January 28, 2015. A publicly-available version is in ADAMS under Accession No. ML15036A032. Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.16, “Containment Leakage Rate Testing Program,” for CPNPP, Units 1 and 2, to allow an increase in the 10 CFR part 50, appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” Type A Integrated Leak Rate Test (ILRT) interval from a 10-year frequency to a maximum of 15 years and the extension of the containment isolation valves leakage Type C tests from its current 60-month frequency to 75 months in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A, “Industry Guidance for Implementing Performance Based Option of 10 CFR part 50, appendix J,” July 2012 (ADAMS Accession No. ML12221A202), and conditions and limitations specified in NEI 94-01, Revision 2-A, “Industry Guidance for Implementing Performance Based Option of 10 CFR part 50, appendix J,” October 2008 (ADAMS Accession No. ML100620847), in addition to limitations and conditions of NEI 94-01, Revision 3-A. The proposed change would also delete the listing of one-time exceptions previously granted to ILRT frequencies. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the CPNPP, Units 1 and 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all internal events accident sequences for CPNPP, of 1.00E-02 person rem/yr [roentgen equivalent man per year] to 6.51 person-rem/yr for Unit 1 and 6.53 person-rem/yr for Unit 2 using the EPRI [Energy Power Research Institute] guidance with the base case corrosion included. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated. As documented in NUREG-1493 [, “Performance-Based Containment Leak-Test Program: Draft Report for Comment,” January 1995 (not publicly available)], Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CPNPP, Units 1 and 2 Type A test history supports this conclusion. The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and; (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers] Section XI, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated. The proposed amendment also deletes exceptions previously granted to allow one-time extensions of the ILRT test frequency for both Units 1 and 2. These exceptions were for activities that have already taken place so their deletion is solely an administrative action that has no effect on any component and no impact on how the units are operated. Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the CPNPP, Unit 1 and 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant ( i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. The proposed amendment also deletes exceptions previously granted to allow one-time extensions of the ILRT test frequency for both Units 1 and 2. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the units are operated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed amendment to TS 5.5.16 involves the extension of the CPNPP, Units 1 and 2 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for CPNPP, Units 1 and 2. The proposed surveillance interval extension is bounded by the 15-year ILRT Interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals. The proposed amendment also deletes exceptions previously granted to allow one-time extensions of the ILRT test frequency for both Units 1 and 2. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action and does not change how the units are operated and maintained. 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: Timothy P. Matthews, Esq., Morgan, Lewis and Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004. NRC Branch Chief: Michael T. Markley. 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 4, 2014. A publicly-available version is in ADAMs under Accession No. ML14339A637. Description of amendment request: The proposed change would amend Combined License (COL) Nos. NPF-93 and NPF-94 for the Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3 by changing the structure and layout of various areas of the annex building. The proposed amendment requires changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant- specific Design Control Document (DCD) Tier 2 information and involves changes to related plant-specific Tier 2* and Tier 1 information, with corresponding changes to the associated COL Appendix C information. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's Advanced Passive 1000 DCD, the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed additions of a new nonsafety-related battery, battery room and battery equipment room, the room height increase, the floor thickness changes, the relocation of a non-structural internal wall, and the associated wall, room and corridor changes within the annex building do not adversely affect the fire loading analysis durations of the affected fire zones and areas ( i.e., the calculated fire durations remain less than their design values). Thus, the fire loads analysis is not adversely affected ( i.e., analysis results remain acceptable). The safe shutdown fire analysis is not affected. The proposed changes to the structural configuration, including anticipated equipment loading, room height, and floor thickness are accounted for in the updated structural configuration model that was used to analyze the Annex Building for safe shutdown earthquake (SSE) and other design loads and load combinations, thus the structural analysis is not adversely affected. The structural analysis description and results in the UFSAR are unchanged. The relocated internal Annex Building wall is non-structural, thus this change does not affect the structural analyses for the Annex Building. The proposed changes do not involve any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. The rooms affected by the proposed changes do not contain or interface with safety-related equipment, thus the proposed changes would not affect any safety-related equipment or accident mitigating function. The radioactive material source terms and release paths used in the safety analyses are unchanged, thus the radiological releases in the accident analyses are not affected. With the conversion of an annex building room to a battery room, the building volume serviced by nuclear island nonradioactive ventilation system decreases by approximate five percent. This reduced volume is used in the post-accident main control room dose portion of the UFSAR LOCA radiological analysis. However, the volume decrease is not sufficient to change the calculated main control room dose reported in the UFSAR, and control room habitability is not affected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed additions of a new nonsafety-related battery, battery room and battery equipment room, the room height increase, the floor thickness changes, the relocation of a non-structural internal wall, and their associated wall, room and corridor changes do not change fire barrier performance, and the fire loading analyses results remain acceptable. The room height and floor thickness changes are consistent with the annex building configuration used in the building's structural analysis. The relocated internal wall is non-structural, thus the structural analyses for the annex building are not affected. The affected rooms and associated equipment do not interface with components that contain radioactive material. The affected rooms do not contain equipment whose failure could initiate an accident. The proposed changes do not create a new fault or sequence of events that could result in a radioactive material release. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed additions of a new nonsafety-related battery, battery room and battery equipment room, the room height increase, the floor thickness changes, the relocation of a non-structural internal wall, and their associated wall, room and corridor changes do not change the fire barrier performance of the affected fire areas. The affected rooms do not contain safety-related equipment, and the safe shutdown fire analysis is not affected. Because the proposed change does not alter compliance with the construction codes to which the annex building is designed and constructed, the proposed changes to the structural configuration, including anticipated equipment loading, room height, and floor thickness do not adversely affect the safety margins associated with the seismic Category II structural capability of the annex building. The floor areas and amounts of combustible material loads in affected fire zones and areas do not significantly change, such that their fire duration times remain within their two-hour design value, thus the safety margins associated with the fire loads analysis are not affected. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus no margin of safety is reduced. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. NRC Branch Chief: Lawrence Burkhart. 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: February 10, 2015. A publicly-available version is in ADAMS under Accession No. ML15041A698. Description of amendment request: The proposed change would amend Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3 by revising Tier 2* information contained within the Human Factors Engineering Design Verification, Task Support Verification and Integrated System Validation plans. These documents are incorporated by reference into the VCSNS Units 2 and 3 Updated Final Safety Analysis Report and will additionally require changes to be made to affected Tier 2 information. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment includes changes to Integrated System Validation (ISV) activities, which are performed on the AP1000 plant simulator to validate the adequacy of the AP1000 human systems interface design and confirm that it meets human factors engineering principles. The proposed changes involve administrative details related to performance of the ISV, and no plant hardware or equipment is affected whose failure could initiate an accident, or that interfaces with a component that could initiate an accident, or that contains radioactive material. Therefore, these changes have no effect on any accident initiator in the Updated Final Safety Analysis Report (UFSAR), nor do they affect the radioactive material releases in the UFSAR accident analysis. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment includes changes to ISV activities, which are performed on the AP1000 plant simulator to validate the adequacy of the AP1000 human system interface design and confirm that it meets human factors engineering principles. The proposed changes involve administrative details related to performance of the ISV, and no plant hardware or equipment is affected whose failure could initiate an accident, or that interfaces with a component that could initiate an accident, or that contains radioactive material. Although the ISV may identify a need to initiate changes to add, modify, or remove plant structures, systems, or components, these changes will not be made directly as part of the ISV. 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 amendment includes changes to ISV activities, which are performed on the AP1000 plant simulator to validate the adequacy of the AP1000 human system interface design and confirm that it meets human factors engineering principles. The proposed changes involve administrative details related to performance of the ISV, and do not affect any safety-related equipment, design code compliance, design function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed 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. NRC Branch Chief: Lawrence Burkhart. Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-366, Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Date of amendment request: October 10, 2014. A publicly-available version is in ADAMS under Accession No. ML14288A226. Description of amendment request: The licensee requested 21 revisions to the Technical Specifications. The licensee states the changes were chosen to increase the consistency between the Hatch Technical Specifications, the Improved Standard Technical Specifications, and the Technical Specifications of other plants in the Southern Nuclear Operating Company fleet. A list of the requested revisions is included in Enclosure 1 of the application. 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 for each of the 24 changes requested, which is presented below: 2.1 TSTF-30-A, Revision 3, “Extend the Completion Time for Inoperable Isolation Valve to a Closed System to 72 Hours.” Specification 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” Action C, TS page 3.6-9, is revised to provide a 72 hour Completion Time for penetration flow paths with one inoperable PCIV with a closed system. Significant Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change extends the Completion Time to isolate an inoperable primary containment isolation valve (PCIV) from 4 hours to 72 hours when the PCIV is associated with a closed system. The PCIVs are not an initiator of any accident previously evaluated. The consequences of a previously evaluated accident during the extended Completion Time are the same as the consequences during the existing Completion Time. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration to the plant ( i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the 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 accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change extends the Completion Time to isolate an inoperable primary containment isolation valve (PCIV) from 4 hours to 72 hours when the PCIV is associated with a closed system. The PCIVs serve to mitigate the potential for radioactive release from the primary containment following an accident. The design and response of the PCIVs to an accident are not affected by this change. The revised Completion Time is appropriate given the isolation capability of the closed system. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.2 TSTF-45-A, Revision 2, “Exempt Verification of CIVs that are Locked, Sealed or Otherwise Secured” The proposed change revises SRs 3.6.1.3.2 and 3.6.1.3.3 in Specification 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” to exempt manual PCIVs and blind flanges which are locked, sealed, or otherwise secured in position from position verification requirements. The proposed change also revises SR 3.6.4.2.1 in Specification 3.6.4.2, “Secondary Containment Isolation Valves (SCIVs),” to exempt manual SCIVs and blind flanges which are locked, sealed, or otherwise secured in position from position verification requirements. Signification Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change exempts manual primary containment isolation valves and blind flanges located inside and outside of containment, and manual secondary containment isolation valves and blind flanges, that are locked, sealed, or otherwise secured in position from the periodic verification of valve position required by Surveillance Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. The exempted valves and devices are verified to be in the correct position upon being locked, sealed, or secured. Because the valves and devices are in the condition assumed in the accident analysis, the proposed change will not affect the initiators or mitigation of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration to the plant ( i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the 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 accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change exempts manual primary containment isolation valves and blind flanges located inside and outside of containment, and manual secondary containment isolation valves and blind flanges, that are locked, sealed, or otherwise secured in position from the periodic verification of valve position required by Surveillance Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. These valves and devices are administratively controlled and their operation is a non-routine event. The position of a locked, sealed or secured blind flange or valve is verified at the time it is locked, sealed or secured, and any changes to their position is performed under administrative controls. Industry experience has shown that these valves are generally found to be in the correct position. Since the change impacts only the frequency of verification for blind flange and valve position, the proposed change will provide a similar level of assurance of correct position as the current frequency of verification. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.3 TSTF-46-A, Revision 1, “Clarify the CIV Surveillance to Apply Only to Automatic Isolation Valves” The proposed change modifies SR 3.6.1.3.5 in Specification 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” and SR 3.6.4.2.2, in Specification 3.6.4.2, “Secondary Containment Isolation Valves (SCIVs),” including their associated Bases, to delete the requirement to verify the isolation time of “each power operated” containment isolation valve and only require verification of each “power operated automatic isolation valve.” Signification Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises the requirements in Technical Specification Surveillance Requirements (SRs) 3.6.1.3.5 and 3.6.4.2.2, and their associated Bases, to delete the requirement to verify the isolation time of “each power operated” PCIV and SCIV and only require verification of closure time for each “automatic power operated isolation valve.” The closure times for PCIVs and SCIVs that do not receive an automatic closure signal are not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The PCIVs and SCIVs are used to respond to accidents previously evaluated. Power operated PCIVs and SCIVs that do not receive an automatic closure signal are not assumed to close in a specified time. The proposed change does not change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not result in a change in the manner in which the PCIVs and SCIVs provide plant protection or introduce any new or different operational conditions. Periodic verification that the closure times for PCIVs and SCIVs that receive an automatic closure signal are within the limits established by the accident analysis will continue to be performed under SRs 3.6.1.3.5 and 3.6.4.2.2. The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. There are also no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant ( i.e., no new or different type of equipment will be installed). Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change provides clarification that only PCIVs and SCIVs that receive an automatic isolation signal are within the scope of SRs 3.6.1.3.5 and 3.6.4.2.2. The proposed change does not result in a change in the manner in which the PCIVs and SCIVs provide plant protection. Periodic verification that closure times for PCIVs and SCIVs that receive an automatic isolation signal are within the limits established by the accident analysis will continue to be performed. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.4 TSTF-222-A, Revision 1, “Control Rod Scram Time Testing” Specification 3.1.4, “Control Rod Scram Times,” SRs 3.1.4.1 and 3.1.4.4, are revised to only require scram time testing of control rods that are in an affected core cell. The SR 3.1.4.1 Frequency “Prior to exceeding 40% RTP after fuel movement within the reactor vessel,” is eliminated and a new Frequency is added to SR 3.1.4.4 which states, “Prior to exceeding 40% RTP after fuel movement within the affected core cell.” Significant Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change clarifies the intent of Surveillance testing in Specification 3.1.4, “Control Rod Scram Times.” The existing Specification wording requires control rod scram time testing of all control rods whenever fuel is moved within the reactor pressure vessel, even though the Technical Specification Bases state that control rod scram time testing is only required in the affected core cells. The Frequency of Surveillances 3.1.4.1 and 3.1.4.4 are revised to implement the Bases statement in the Specifications. The proposed change does not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. The control rods will continue to perform their function as designed. Operation in accordance with the proposed Technical Specifications will ensure that all analyzed accidents will continue to be mitigated as previously analyzed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration to the plant ( i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the 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 accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change clarifies the intent of Surveillance testing in Specification 3.1.4, “Control Rod Scram Times.” The existing Specification wording requires control rod scram time testing of all control rods whenever fuel is moved within the reactor pressure vessel, even though the Technical Specification Bases state that the control rod scam time testing is only required in the affected core cells. The proposed change will not affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. Control rod scram time testing will be performed following any fuel movement that could affect the scram time. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.5 TSTF-264-A, Revision 0, “3.3.9 and 3.3.10—Delete Flux Monitors Specific Overlap Requirement SRs” The proposed change revises Specification 3.3.1.1, “RPS Instrumentation,” by deleting Surveillances 3.3.1.1.6 and 3.3.1.1.7, which verify the overlap between the source range monitor (SRM) and the intermediate range monitor (IRM), and between the IRM and the average power range monitor (APRM). Significant Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change eliminates two Surveillances Requirements (SRs) (SRs 3.3.1.1.6 and 3.3.1.1.7) which verify the overlap between the source range monitor (SRM) and intermediate range monitor (IRM) and between the IRM and the average power range monitor (APRM). The testing requirement is incorporated in the existing Channel Check Surveillance (SR 3.3.1.1.1). The proposed change does not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. The SRM, IRM, and APRM will continue to perform their function as designed. Operation in accordance with the proposed Technical Specifications will ensure that all analyzed accidents will continue to be mitigated as previously analyzed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration to the plant ( i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the 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 accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change eliminates SRs 3.3.1.1.6 and 3.3.1.1.7 which verify the overlap between the SRM and IRM and between the IRM and the APRM. The testing requirement is incorporated in the existing Channel Check Surveillance (SR 3.3.1.1.1). The proposed change will not affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. Instrument channel overlap will continue to be verified under the existing Channel Check surveillance. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.6 TSTF-269-A, Revision 2, “Allow Administrative Means of Position Verification for Locked or Sealed Valves” The proposed change modifies Specification 3.6.1.3, “Primary Containment Isolation Valves,” and Specification 3.6.4.2, “Secondary Containment Isolation Valves.” The specifications require penetrations with an inoperable isolation valve to be isolated and periodically verified to be isolated. A Note is added to Specification 3.6.1.3, Actions A and C, and Specification 3.6.4.2, Action A, to allow isolation devices that are locked, sealed, or otherwise secured to be verified by use of administrative means. Significant Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change modifies Specification 3.6.1.3, “Primary Containment Isolation Valves,” and Specification 3.6.4.2, “Secondary Containment Isolation Valves.” The specifications require penetrations with an inoperable isolation valve to be isolated and periodically verified to be isolated. A Note is added to Specification 3.6.1.3, Actions A and C, and Specification 3.6.4.2, Action A, to allow isolation devices that are locked, sealed, or otherwise secured to be verified by use of administrative means. The proposed change does not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. The inoperable containment penetrations will continue to be isolated, and hence perform their isolation function. Operation in accordance with the proposed Technical Specifications will ensure that all analyzed accidents will continue to be mitigated as previously analyzed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a physical alteration to the plant ( i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the 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 accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change will not affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. The primary and secondary containment isolation valves will continue to be operable or will be isolated as required by the existing specifications. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 2.7 TSTF-273-A, Revision 2, “Safety Function Determination Program Clarifications” The proposed Technical Specification (TS) changes add explanatory text to the Bases for limiting condition for operation (LCO) 3.0.6 clarifying the “appropriate LCO for loss of function,” and that consideration does not have to be made for a loss of power in determining loss of function. Explanatory text is also added to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.12 to provide clarification of these same issues. Signification Hazards Consideration: SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed Technical Specification (TS) changes add explanatory text to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.10 to clarify in the requirements that consideration does not have to be made for a loss of power in determining loss of function. The Bases for limiting condition for operations (LCO) 3.0.6 are revised to provide clarification of the “appropriate LCO for loss of function,” and that consideration does not have to be made for a loss of power in determining loss of function. The changes are editorial and administrative in nature, and therefore do not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are editorial and administrative in nature and do not result in a change in the manner in which the plant operates. The loss of function of any specific component will continue to be addressed in its specific TS LCO and plant configuration will be governed by the required actions of those LCOs. The proposed changes are clarifications that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant ( i.e., no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator. 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 reduct [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A2015-07192. Public record. Not legal advice.