80 FR 17083 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
NUCLEAR REGULATORY COMMISSION Federal Register Volume 80, Issue 61 (March 31, 2015)
Federal Register Volume 80, Issue 61 (March 31, 2015)
| Page Range | 17083-17109 | |
| FR Document | 2015-07192 |
[Federal Register Volume 80, Number 61 (Tuesday, March 31, 2015)] [Notices] [Pages 17083-17109] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2015-07192] ----------------------------------------------------------------------- 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: [email protected]. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document. Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-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 [email protected]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in 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 [[Page 17084]] 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 Sec. 50.92 of Title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period 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 [[Page 17085]] 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 [email protected], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a 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 [email protected], 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- [[Page 17086]] 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 [[Page 17087]] 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 [[Page 17088]] 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[emsp14][deg]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[emsp14][deg]F to <=102[emsp14][deg]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 [[Page 17089]] 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 [[Page 17090]] 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? [[Page 17091]] 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 [[Page 17092]] 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: [[Page 17093]] 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- [[Page 17094]] 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. [[Page 17095]] 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. [[Page 17096]] 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 [[Page 17097]] 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 [[Page 17098]] 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 reduction in a margin of safety? Response: No. The proposed changes to TS 5.5.10 are clarifications and are editorial and administrative in nature. No changes are made the LCOs for plant equipment, the time required for the TS Required Actions to be completed, or the out of service time for the components involved. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do 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 changes will not result in plant operation in a configuration outside the design basis. Therefore, the proposed changes do 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. 2.8 TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction Notes'' The proposed change revises several Specification 3.8.1, ``AC Sources--Operating,'' Surveillance Notes to allow full or partial performance of the SRs to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. These Surveillances currently have Notes prohibiting their performance in Modes 1 or 2, or in Modes 1, 2, or 3. SR 3.8.1.6 (ISTS SR 3.8.1.8), which tests the transfer of Alternating (AC) sources from normal to alternate offsite circuits, contains a Note prohibiting performance in Mode 1 or 2. The Note is modified to state that performance is normally prohibited in Mode 1 or 2 but may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.7 (ISTS SR 3.8.1.9), which tests the ability of the emergency diesel generator (DG) to reject a load greater than or equal to its associated single largest post-accident load, contains a Note prohibiting performance in Mode 1 or 2. An exception is provided for the swing DG. The Note is modified to state that performance is normally prohibited in Mode 1 or 2 but may be performed to re- establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.8 (ISTS SR 3.8.1.10), which tests emergency DG operation following a load rejection of greater than or equal to 2775 kW, contains a Note prohibiting performance in Mode 1 or 2. The Note is modified to state that performance is normally prohibited in Mode 1 or 2 but portions of the SR may be performed to re- establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.9 (ISTS SR 3.8.1.11), which tests the response to a loss of offsite power signal, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but portions of the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.10 (ISTS SR 3.8.1.12), which tests response to an Emergency Core Cooling System (ECCS) initiation signal, contains a Note prohibiting performance in Mode 1 or 2. The Note is modified to state that performance is normally prohibited in Mode 1 or 2, but the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.11 (ISTS SR 3.8.1.13), which tests that each DGs automatic trips are bypassed on a loss of voltage signal concurrent with an ECCS initiation signal, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.12 (ISTS SR 3.8.1.14), which performs a 24 hour loaded test run of the DG, contains a Note prohibiting performance in Mode 1 or 2. The Note is modified to state that performance is normally prohibited in Mode 1 or 2, but the SR may be performed to re- establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.14 (ISTS SR 3.8.1.16), which verifies transfer from DG to offsite power, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but portions of the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.15 (ISTS SR 3.8.1.17), which verifies than a DG operating in test mode will return to ready-to-load condition and energize the emergency load from offsite power on receipt of an ECCS initiation signal, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but portions of the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.16 (ISTS SR 3.8.1.18), which verifies the interval between each sequenced load, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. SR 3.8.1.17 (ISTS SR 3.8.1.19), which verifies the response to a loss of offsite power signal and Engineered Safety Features (ESF) actuation signal, contains a Note prohibiting performance in Mode 1, 2, or 3. The Note is modified to state that performance is normally prohibited in Mode 1, 2, or 3, but portions of the SR may be performed to re-establish Operability provided an assessment determines the safety of the plant is maintained or enhanced. 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: [[Page 17099]] 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 Mode restriction Notes on eleven emergency diesel generator (DG) Surveillances to allow performance of the Surveillance in whole or in part to re-establish emergency DG Operability. The emergency DGs and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result the probability of any accident previously evaluated is not increased. The proposed change allows Surveillance testing to be performed in whole or in part to re-establish Operability of an emergency DG. The consequences of an accident previously evaluated during the period that the emergency DG is being tested to re-establish Operability are no different from the consequences of an accident previously evaluated while the emergency DG is inoperable. As a result, the consequences of any accident previously evaluated are not increased. 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 purpose of Surveillances is to verify that equipment is capable of performing its assumed safety function. The proposed change will only allow the performance of the Surveillances to re- establish Operability and the proposed changes may not be used to remove an emergency DG from service. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the normally prohibited Modes. 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. 2.9 TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to Technical Specification 1.4, Frequency'' The change inserts a discussion paragraph into Specification 1.4, and two new examples are added to facilitate the use and application of SR Notes that utilize the terms ``met'' and ``perform.'' 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 changes insert a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of Surveillance Requirement (SR) Notes that utilize the terms ``met'' and ``perform''. The changes also modify SRs in multiple Specifications to appropriately use ``met'' and ``perform'' exceptions. The changes are administrative in nature because they provide clarification and correction of existing expectations, and therefore the proposed change does not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not significantly 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 administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes 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 reduction in a margin of safety? Response: No. The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, or alter their operation. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do 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 changes will not result in plant operation in a configuration outside the design basis. Therefore, the proposed changes do 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. 2.10 TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM Table to Separate Condition Entry for Each Penetration'' Specification 3.3.3.1, ``Post Accident Monitoring (PAM) Instrumentation,'' Function 6, is renamed from ``Primary Containment Isolation Valve Position'' to ``Penetration Flow Path Primary Containment Isolation Valve Position.'' 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 separate condition entry Note in Specification 3.3.3.1, ``Post Accident Monitoring (PAM) Instrumentation,'' for Function 6, ``Primary Containment Isolation Valve Position,'' and Function 9, ``Suppression Pool Water Temperature.'' The proposed change does not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. The actions taken for inoperable PAM channels are not changed. 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 [[Page 17100]] 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 PAM channels will continue to be operable or the existing, appropriate actions will be followed. 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. 2.11 TSTF-306-A, Revision 2, ``Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration'' The proposed change revises Specification 3.3.6.1, ``Primary Containment Isolation Instrumentation.'' An Actions Note is added allowing penetration flow paths to be unisolated intermittently under administrative controls. The traversing incore probe (TIP) isolation system is also segregated into a separate Function, allowing 12 hours to place the channel in trip and 24 hours to isolate the penetration. A new Condition G is added for the new TIP isolation system Function. Condition G is referenced from Required Action C.1 when Conditions A or B are not met. The subsequent Actions are renumbered. 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 revises Specification 3.3.6.1, ``Primary Containment Isolation Instrumentation.'' An Actions Note is added allowing penetration flow paths to be unisolated intermittently under administrative controls. The traversing incore probe (TIP) isolation system is segregated into a separate Function, allowing 12 hours to place the channel in trip and 24 hours to isolate the penetration. A new Action G is added which is referenced by the new TIP isolation system Function. The subsequent Actions are renumbered. The proposed change does not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. The allowance to unisolate a penetration flow path will not have a significant effect on mitigation of any accident previously evaluated because the penetration flow path can be isolated, if needed, by a dedicated operator. The option to isolate a TIP System penetration will ensure the penetration will perform as assumed in the accident analysis. 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 allowance to unisolate a penetration flow path will not have a significant effect on a margin of safety because the penetration flow path can be isolated manually, if needed. The option to isolate a TIP System penetration will ensure the penetration will perform as assumed in the accident analysis. 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. 2.12 TSTF-308-A, Revision 1, ``Determination of Cumulative and Projected Dose Contributions in RECP'' The proposed change revises Specification 5.5.4, ``Radioactive Effluent Controls Program,'' paragraph e, to describe the original intent of the dose projections. 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 revises Specification 5.5.4, ``Radioactive Effluent Controls Program,'' paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are not an assumption in any accident previously evaluated and have no effect on the 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 revises Specification 5.5.4, ``Radioactive Effluent Controls Program,'' paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are administrative tools to assure compliance with regulatory limits. The proposed change revises the requirement to clarify the intent, thereby improving the administrative control over this process. As a result, any effect on the margin of safety should be minimal. 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. 2.13 TSTF-318-A, Revision 0, ``Revise 3.5.1 for One LPCI Pump Inoperable in Each of Two ECCS Divisions'' The proposed change adds a provision to Condition A of Specification 3.5.1, ``ECCS--Operating,'' to allow one Low Pressure Coolant Injection (LPCI) pump to be inoperable in each subsystem for a period of seven days. 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 adds a provision to Condition A of Technical Specification (TS) 3.5.1 to allow one Low Pressure Coolant Injection (LPCI) pump to be inoperable in each subsystem for a period of seven days. The change to allow one LPCI pump to be inoperable in both subsystems is more reliable than what is currently allowed by Condition A, which requires entry into [[Page 17101]] shutdown limiting condition for operation (LCO) 3.0.3 under these conditions. The LPCI mode of the Residual Heat Removal system is not assumed to be initiator of any analyzed event sequence. The consequences of an accident previously evaluated under the proposed allowance are no different than the consequences under the existing requirements. 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 adds a provision to Condition A of Technical Specification TS 3.5.1 to allow one LPCI pump to be inoperable in each subsystem for a period of seven days. The change to allow one LPCI pump to be inoperable in both subsystems is more reliable than what is currently allowed by Condition A, which requires entry into shutdown LCO 3.0.3 under these conditions. The proposed change does not affect any safety analysis assumptions. 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. 2.14 TSTF-322-A, Revision 2, ``Secondary Containment and Shield Building Boundary Integrity SRs' The proposed change revises Specification 3.6.4.1, ``Secondary Containment,'' SRs 3.6.4.1.3 and 3.6.4.1.4 to clarify the intent of the Surveillances. 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 revises Specification 3.6.4.1, ``Secondary Containment,'' Surveillance Requirements (SRs) 3.6.4.1.3 and 3.6.4.1.4 to clarify the intent of the Surveillances. The secondary containment and the standby gas treatment (SGT) system are not initiators of any accident previously evaluated. 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 is an clarification of the intent of the surveillances to ensure that the secondary containment is not inappropriately declared inoperable when a SGT subsystem is inoperable. The safety functions of the secondary containment and the SGT system are not affected. This change is a correction that ensures that the intent of the secondary containment surveillances is clear. 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. 2.15 TSTF-323-A, Revision 0, ``EFCV Completion Time to 72 hours'' The proposed change revises Specification 3.6.1.3, ``Primary Containment Isolation Valves,'' Action C, to provide a 72 hour Completion Time instead of a 12 hour Completion Time to isolate an inoperable excess flow check valve (EFCV). 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 revises Specification 3.6.1.3, ``Primary Containment Isolation Valves,'' Action C, to provide a 72 hour Completion Time instead of a 12 hour Completion Time to isolate an inoperable excess flow check valve (EFCV). The primary containment isolation valves (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 penetration equipped with an excess flow check valve from 12 hours to 72 hours. 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 EFCVs are on penetrations that have been found to have acceptable barrier(s) in the event that the single isolation valve fails. 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. 2.16 TSTF-374-A, Revision 0, ``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil'' The proposed change revises Specification 5.5.9, ``Diesel Fuel Oil Testing Program,'' to remove references to the specific American Society for Testing and Materials (ASTM) Standard from the Administrative Controls Section of TS, and places them in a licensee-controlled document. Also, alternate criteria are added to establish the acceptability of new fuel oil for use prior to and following the addition to storage tanks. 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? [[Page 17102]] Response: No. The proposed changes remove the references to specific ASTM standards from the Administrative Controls Section of the Technical Specifications (TS) and place them in a licensee controlled document. Requirements to perform testing in accordance with the applicable ASTM standards is retained in the TS as are requirements to perform testing of both new and stored diesel fuel oil. Future changes to the licensee controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59 to ensure that these changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated. In addition, tests used to establish the acceptability of new fuel oil for use prior to and following the addition to storage tanks has been expanded to recognize more rigorous testing of water and sediment content. Relocating the specific ASTM standard references from the TS to a licensee controlled document and allowing a water and sediment content test to be performed to establish the acceptability of new fuel oil will not affect nor degrade the ability of the emergency diesel generators (EDGs) to perform their specified safety function. Fuel oil quality will continue to be tested and maintained to ASTM requirements. Diesel fuel oil testing is not an initiator of any accident previously evaluated, and the proposed changes do not adversely affect any accident initiators or precursors, or alter design assumptions, conditions, and configuration of the facility, or the manner in which the plant is operated. The proposed changes do not adversely affect the ability of structures, systems, and components to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. Therefore, the proposed changes do 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 changes remove the references to specific ASTM standards from the Administrative Controls Section of TS and place them in a licensee controlled document. In addition, the tests used to establish the acceptability of new fuel oil for use prior to and following the addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The requirements retained in the TS will continue to require testing of new and stored diesel fuel oil to ensure the proper functioning of the EDGs. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes remove the references to specific ASTM standards from the Administrative Controls Section of TS and place them in a licensee controlled document. Instituting the proposed changes will continue to ensure the use of applicable ASTM standards to evaluate the changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there is no significant reduction in a margin of safety. The margin of safety provided by the EDGs is unaffected by the proposed changes since TS requirements will continue to ensure fuel oil is of the appropriate quality. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies while maintaining sufficient controls to preserve the current margins 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. 2.17 TSTF-400-A, Revision 1, ``Clarify SR on Bypass of DG Automatic Trips'' The proposed change revises Specification 3.8.1, ``AC Sources-- Operating,'' Surveillance 3.8.1.11, to clarify that the intent of the SR is to test the non-critical emergency DG automatic trips. 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 change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This change clarifies the purpose of Surveillance Requirement (SR) 3.8.1.11, which is to verify that non-critical automatic emergency diesel generator (DG) trips are bypassed in an accident. The non-critical automatic DG trips and their bypasses are not initiators of any accident previously evaluated. Therefore, the probability of any accident is not significantly increased. Additionally, the function of the emergency DG in mitigating accidents is not changed. The revised SR continues to ensure the emergency DG will operate as assumed in the accident analysis. 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. This change clarifies the purpose of SR 3.8.1.11, which is to verify that non-critical automatic emergency DG trips are bypassed in an accident. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed), or a change in the methods governing normal plant operation. Thus, this 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. This change clarifies the purpose of SR 3.8.1.11, which is to verify that non-critical automatic DG trips are bypassed in an accident. This change clarifies the purpose of the SR, which is to verify that the emergency DG is capable of performing the assumed safety function. The safety function of the emergency DG is unaffected, so the change does not affect the margin of safety. Therefore, this change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed change presents no 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. 2.18 TSTF-439-A, Revision 2, ``Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO'' Specifications 3.1.7, ``Standby Liquid Control (SLC) System;'' 3.6.4.3, ``Standby Gas Treatment (SGT) System;'' 3.8.1, ``AC Sources--Operating;'' and 3.8.7, ``Distribution Systems-- Operating,'' contain Required Actions with a second Completion Time to establish a limit on the maximum time allowed for any combination of Conditions that result in a single continuous failure to meet the LCO. These Completion Times (henceforth referred to as ``second Completion Times'') are joined by an ``AND'' logical connector to the Condition-specific Completion Time and state ``X days from discovery of failure to meet the LCO'' (where ``X'' varies by specification). The proposed change deletes these second Completion Times from the affected Required Actions. It also revises ISTS Example 1.3-3 to remove the discussion of second Completion Times and to revise the discussion in that Example to state that alternating between Conditions in such a manner that operation could continue indefinitely without restoring systems to meet the LCO is inconsistent with the basis of the Completion Times and is inappropriate. Therefore, the licensee shall have administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. 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. [[Page 17103]] The proposed change eliminates certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the remaining Completion Time are no different than the consequences of the same accident during the removed Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. 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 to delete the second Completion Time does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside of 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. 2.19 TSTF-458-T, Revision 0, ``Removing Restart of Shutdown Clock for Increasing Suppression Pool Temperature'' The proposed change revises Specification 3.6.2.1, ``Suppression Pool Average Temperature,'' Required Actions D and E, to eliminate redundant requirements. 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 revises Specification 3.6.2.1, ``Suppression Pool Average Temperature,'' Required Actions D and E, to eliminate redundant requirements when suppression pool temperature is above the Limiting Conditions for Operation (LCO) limit. Suppression pool temperature is not an initiator to any accident previously evaluated. Suppression pool temperature may affect the mitigation of accidents previously evaluated. The proposed change reduces the time allowed to operate with suppression pool temperature above the limit. The consequences of an accident under the proposed change are no different than under the current requirements. 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 revises Specification 3.6.2.1, ``Suppression Pool Average Temperature,'' Required Actions D and E, to eliminate redundant requirements when suppression pool temperature is above the LCO limit. The proposed change reduces the time allowed to operate with suppression pool temperature above the limit. The proposed revision will not adversely affect the margin of safety as it corrects the Actions to provide appropriate compensatory measures when suppression pool temperature is greater than the limit. 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. 2.20 TSTF-464-T, Revision 0, ``Clarify the Control Rod Block Instrumentation Required Action'' The proposed change revises Specification 3.3.2.1, Required Action C.2.1.2 from ``Verify by administrative methods that startup with RWM inoperable has not been performed in the last calendar year'' to ``Verify by administrative methods that startup with RWM inoperable has not been performed in the last 12 months.'' 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 change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises a Required Action to limit startup with the Rod Worth Minimizer (RWM) inoperable from once per calendar year to once per 12 months. The RWM is used to minimize the possibility and consequences of a control rod drop accident. This change clarifies the intent of the limitation, but does not affect the requirement for the RWM to be operable. As, over time, the number of startups with the RWM inoperable will not increase, the probability of any accident previously evaluated is not significantly increased. As the RWM is still required to be operable, the consequences of an any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises a Required Action to limit startup with the Rod Worth Minimizer inoperable from once per calendar year to once per 12 months. No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. 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 revises a Required Action to limit startup with the Rod Worth Minimizer (RWM) inoperable from once per calendar year to once per 12 months. This clarifies the intent of the Required Action. The number of startups with RWM inoperable is not increased. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed change presents no 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. 2.21 ISTS Adoption #1--Revise the 5.5.7 Introductory Paragraph To Be Consistent With the ISTS The proposed change revises the introductory paragraph of Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' to be consistent with the ISTS. Specific requirements to perform testing after [[Page 17104]] structural maintenance on the HEPA filter or charcoal adsorber housing or following painting, fire or chemical release, and after every 720 hours of operation are relocated to the licensee- controlled program. The existing wording states, ``The VFTP will establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and C.5.d, or: (1) After any structural maintenance on the HEPA filter or charcoal adsorber housings, (2) following painting, fire or chemical release in any ventilation zone communicating with the system, or 3) after every 720 hours of charcoal adsorber operation.'' The proposed wording states, ``A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and C.5.d, and in accordance with Regulatory Guide 1.52, Revision 2.'' 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 revises the introductory paragraph of Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' to be consistent with the ISTS. Specific requirements to perform testing after structural maintenance on the HEPA filter or charcoal adsorber housing or following painting, fire or chemical release, and after every 720 hours of operation are retained as a reference to Regulatory Guide requirements and general requirements in Surveillance Requirement (SR) 3.0.1. Implementation of these requirements will be in the licensee-controlled VFTP. The VFTP will be maintained in accordance with 10 CFR 50.59. Since any changes to the VFTP will be evaluated under 10 CFR 50.59, no significant increase in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve 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 revises the introductory paragraph of Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' to be consistent with the ISTS. The proposed change will not reduce a margin of safety because it has no effect on any safety analysis assumption. In addition, no requirements are being removed, but are being replaced with references to an NRC Regulatory Guide and the requirements of SR 3.0.1. Therefore, this 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. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35201 NRC Branch Chief: Robert J. Pascarelli. Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama Date of amendment request: December 11, 2014 (ADAMS Accession No. ML14349A694). Description of amendment request: The amendment would revise Section 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' of the Technical Specifications (TSs) by replacing the current volume requirements with the number of continuous days the diesel generators (DGs) are required to run. The numerical volumes will be maintained in the licensee-controlled TSs Bases document so they may be modified under licensee control. The resulting requirements will specify an inventory of stored diesel fuel oil and lube oil sufficient for a 7-day supply for each DG. This proposed amendment is consistent with NRC's approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-501, Revision 1, ``Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.'' The availability of this TSs improvement was announced in the Federal Register on May 26, 2010 (75 FR 29588). The licensee also proposed additional changes to Section 3.8.3 and Section 5.5.9, ``Diesel Fuel Oil Testing Program,'' to support other related changes proposed by TSTF-501, Revision 1. These additional changes concern fuel oil quality and associated surveillance requirements (SRs). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to TS Section 3.8.3, Conditions A and B, and to SR 3.8.3.1 and SR 3.8.3.2 remove the volume of diesel fuel oil and lube oil required to support 7-day operation of each onsite diesel generator, and the volume equivalent to a 6-day supply, from the TS and replace them with the associated number of days. The numerical volumes will be maintained under licensee control. The specific volume of fuel oil equivalent to a 7 and 6-day supply is calculated using the NRC-approved methodology described in Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems for Standby Diesel Generators'' and ANSI [American National Standards Institute]-N195 1976, ``Fuel Oil Systems for Standby Diesel- Generators.'' The specific volume of lube oil equivalent to a 7-day and 6-day supply is based on the diesel generator manufacturer's consumption values for the run time of the diesel generator. Because the requirement to maintain a 7-day supply of diesel fuel oil and lube oil is not changed and is consistent with the assumptions in the accident analyses, and the actions taken when the volume of fuel oil and lube oil are less than a 6-day supply have not changed, neither the probability nor the consequences of any accident previously evaluated will be affected. The addition of a new Condition D provides a required action and completion time if new fuel oil properties are not within limits. The new SR 3.8.3.5 requires checking for and removing water from the 7-day storage tank every 31 days. The revised Section 5.5.9 adds testing requirements for new fuel oil to be completed prior to the addition of the new fuel oil to the 7-day storage tank, as well as additional testing to be completed prior or within 31 days of the addition. These requirements are more restrictive testing requirements and provide corrective action to be taken if the testing limits are not met. They are taken from the current NRC approved NUREG-1433, Revision 4, ``Standard Technical Specifications, General Electric BWR/4 Plants.'' Improved, more restrictive testing standards will neither change the probability or the consequences of any accident previously evaluated be affected. Therefore, the proposed changes do not involve a significant increase in the [[Page 17105]] 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 changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis but ensures that the diesel generator operates as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes to Section 3.8.3, Conditions A and B, and to SR 3.8.3.1 and SR 3.8.3.2 remove the numerical volume of diesel fuel oil and lube oil required to support 7-day operation of each onsite diesel generator, and the numerical volume equivalent to a 6- day supply from the TS and replaces them with the associated number of days. The numerical volumes will be maintained under licensee control. As the bases for the existing limits on diesel fuel oil volume and lube oil volume are not changed, no change is made to the accident analysis assumptions and no margin of safety is reduced as part of this change. The new, more restrictive, testing requirements, and the provision for corrective action to be taken if the testing limits are not met, are taken from the current NRC approved NUREG-1433, Revision 4, ``Standard Technical Specifications, General Electric BWR/4 Plants.'' These changes do not revise the accident analysis assumptions and no margin of safety is reduced as part of these changes. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902. NRC Branch Chief: Shana R. Helton. Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: November 20, 2014. A publicly-available version is in ADAMS under Accession No. ML14330A247. Description of amendment request: The amendment would revise the Technical Specification (TS) requirements to address NRC Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'' as described in Technical Specification Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.'' Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises or adds SRs [surveillance requirements] that require verification that the Emergency Core Cooling System (ECCS), the Residual Heat Removal (RHR) System, and the Containment Spray System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the RHR System, and the Containment Spray System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises or adds SRs that require verification that the ECCS, the RHR System, and the Containment Spray System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. III. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration. For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice. [[Page 17106]] Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington Date of amendment request: August 22, 2014. A publicly-available version is in ADAMS under Accession No. ML14237A729. Brief description of amendment request: The proposed amendment would revise the technical specification (TS) surveillance requirement (SR) for the ultimate heat sink (UHS) to clarify that spray pond level is the average of the level in both ponds. The design of the ultimate heat sink is such that it is difficult to meet the current SR when only one standby service water (SW) pump is in operation without overflowing a spray pond resulting in a net loss of water inventory, which may challenge the ability of the UHS to provide sufficient inventory for 30 days. However, if the SR is not met, a plant shutdown is required. Date of publication of individual notice in Federal Register: September 5, 2014 (79 FR 53085). Expiration date of individual notice: October 6, 2014 (public comments); November 4, 2014 (hearing requests). IV. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document. DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan Date of amendment request: April 23, 2013, as supplemented by letters dated June 19, and October 13, 2014. Brief description of amendment: The amendment revised the Fermi 2 technical specification (TS) surveillance requirements (SRs) associated with SR 3.8.4.2 and SR 3.8.4.5 to add a battery resistance limit; SR 3.8.6.3 to change the average electrolyte temperature of representative cells, and SR 3.8.4.8 to change the frequency of battery capacity testing. Date of issuance: March 16, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 199. A publicly-available version is in ADAMS under Accession No. ML15057A297; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-43: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 22, 2014 (79 FR 42542). The supplemental letters dated June 19, and October 13, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2015. No significant hazards consideration comments received: No. Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana Date of amendment request: June 13, 2013, as supplemented by letters dated August 28 and November 3, 2014, and January 22, 2015. Brief description of amendment: The amendment revised the Technical Specifications to risk-inform requirements regarding selected Required Action end states by adopting Technical Specification Task Force (TSTF)-423, Revision 1, ``Technical Specifications End States, NEDC- 32998-A,'' with some deviations as approved by the NRC staff. This technical specification improvement is part of the Consolidated Line Item Improvement Process (CLIIP). In addition, it approves a change to the facility operating license for the River Bend Station, Unit 1. The change deletes two license conditions that are no longer applicable and adds a new license condition for maintaining commitments required for the approval of this TSTF into the Updated Safety Analysis Report. Date of issuance: February 17, 2015. Effective date: As of the date of issuance and shall be implemented 90 days from the date of issuance. Amendment No.: 185. A publicly-available version is in ADAMS under Accession No. ML14106A167; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-47: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: August 20, 2013 (78 FR 51226). The supplemental letters dated August 28, and November 3, 2014, and January 22, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 17, 2015. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit 3, Westchester County, New York Date of amendment request: February 4, 2014, as supplemented by letter dated December 9, 2014. Brief description of amendment: The amendment revised Technical Specification 5.5.15, ``Containment Leakage Rate Testing Program,'' to allow a permanent extension of the Type A primary containment integrated leak [[Page 17107]] rate test frequency from once every 10 years to once every 15 years. Date of issuance: March 13, 2015. Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment No.: 256. A publicly-available version is in ADAMS under Accession No. ML15028A308; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. DPR-64: The amendment revised the Facility Operating License and the Technical Specifications. Date of initial notice in Federal Register: July 8, 2014 (79 FR 38587). The supplemental letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 13, 2015. No significant hazards consideration comments received: No Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit 3, Westchester County, New York Date of amendment request: April 1, 2014. Brief description of amendment: The amendment revised Technical Specification Figures 3.4.3-1, ``Heatup Limitations for Reactor Coolant System,'' 3.4.3-2, ``Cooldown Limitations for Reactor Coolant System,'' and 3.4.3-3, ``Hydrostatic and Inservice Leak Testing Limitations for Reactor Coolant System'' to address vacuum fill operations in the TSs. The proposed changes clarify that the figures are applicable for vacuum fill conditions where pressure limits are considered to be met for pressures that are below 0 pounds per square inch gauge (psig) (i.e., up to and including full vacuum conditions). Vacuum fill operations for the RCS can result in system pressures below 0 psig. Date of issuance: March 6, 2015. Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment No.: 255. A publicly-available version is in ADAMS under Accession No. ML15050A144; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. DPR-64: The amendment revised the Facility Operating License and the Technical Specifications. Date of initial notice in Federal Register: October 28, 2014 (79 FR 64223). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 6, 2015. No significant hazards consideration comments received: No Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of amendment request: April 5, 2013, as supplemented by letter dated March 20, 2014. Brief description of amendment: This amendment revised Technical Specification (TS) 2.1.1 and 2.1.2, ``Safety Limits,'' by reducing the reactor steam dome pressure from 785 pounds per square inch gauge (psig) to 685 psig to resolve the Pressure Regulator Failure-Open transient. Date of issuance: March 12, 2015. Effective date: As of the date of issuance, and shall be implemented within 60 days of issuance. Amendment No.: 242. A publicly-available version is in ADAMS under Accession No. ML14272A070; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-35: Amendment revised the License and TS. Date of initial notice in Federal Register: August 6, 2013 (78 FR 47788). The supplement dated March 20, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 2015. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1, Oswego County, New York Date of application for amendment: March 8, 2013, as supplemented by letter dated May 16, 2013, July 8, July16, August 29, 2014, and January 22, 2015. The public versions of these documents are available in ADAMS at the Accession Nos. ML13073A103, ML13144A068, ML14203A050, ML14199A384, ML14251A233, and ML15026A132, respectively. Brief description of amendment: The amendment to the Nine Mile Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63 modified Technical Specification (TS) Table 3.6.2i, ``Diesel Generator Initiation,'' by revising the existing 4.16kV Power Board (PB) 102/103 Emergency Bus Undervoltage (Degraded Voltage) Operating Time value and by updating the Set Point heading title. The TS revisions are being made to resolve the green non-cited violation (NCV) associated with the vital bus degraded voltage protection time delay documented in NRC Inspection Report (IR) 05000220/201101, ``Nine Mile Point Nuclear Station--NRC Unresolved Item Follow-up Inspection Report,'' dated January 23, 2012 (ADAMS Accession No. ML12023A119), specifically, NCV05000220/20 11011-01, ``Vital Bus Degraded Voltage Time Delay Not Maintained within LOCA Analysis Assumptions.'' Date of issuance: March 12, 2015. Effective date: effective as of the date of its issuance and shall be implemented within 60 days. Amendment No.: 217. Renewed Facility Operating License No. DPR-63: Amendment revised the License and Technical Specifications. Date of initial notice in Federal Register: June 11, 2013, (78 FR 35062). The supplements dated May 16, 2013, July 8, July16, August 29, 2014, and January 22, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed no significant hazards consideration determination noticed in the Federal Register on June 11, 2013 (78 FR 35062). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 2015. No significant hazards consideration comments received: No Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50- 277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of application for amendments: July 11, 2014, as supplemented by letter dated December 1, 2014. Brief description of amendments: The amendments incorporate several administrative changes to the Facility Operating Licenses (FOLs) and the Technical Specifications (TSs) such as deleting historical items that are no longer applicable, correcting errors, and removing references that are no longer valid. Date of issuance: March 11, 2015. [[Page 17108]] Effective date: As of the date of issuance, to be implemented within 60 days. Amendments Nos.: 296 and 299. A publicly-available version is in ADAMS under Accession No. ML14363A227; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the FOLs and the TSs. Date of initial notice in Federal Register: September 2, 2014 (79 FR 52062). The supplemental letter dated December 1, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 11, 2015. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania Date of amendment request: October 18, 2013, as supplemented by letters dated June 26, 2014, September 21, 2014, and February 4, 2015. Brief description of amendments: The amendment changes the Beaver Valley Power Station Technical Specifications (TS). Specifically, this change request involves the adoption of an approved change to the standard TS for Westinghouse plants (NUREG-1431), to allow relocation of specific TS surveillance frequencies to a licensee-controlled program. The proposed change is described in TS Task Force (TSTF) Traveler, TSTF-425, Revision 3, ``Relocation Surveillance Frequencies to Licensee Control--RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090850642). A Notice of Availability was published in the Federal Register on July 6, 2009 (74 FR 31996). The proposed change relocates surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program. This change is applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). Date of issuance: March 6, 2015. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 292 and 179. A publicly-available version is in ADAMS under Accession No. ML14322A461; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: January 21, 2014 (79 FR 3416). The supplemental letters dated June 26, 2014, September 21, 2014, and February 4, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 6, 2015. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, Georgia Date of amendment request: November 21, 2013, and supplemented by the letters dated March 5 and June 30, 2014. Brief description of amendment: The amendment authorizes changes to the VEGP Units 3 and 4 Updated Final Safety Analysis Report to revise the details of the effective thermal conductivity resulting from the oxidation of the inorganic zinc component of the containment vessel coating system. Date of issuance: February 26, 2015. Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. Amendment No.: 31. A publicly-available version is in ADAMS under Accession No. ML15028A358; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses. Date of initial notice in Federal Register: March 18, 2014 (79 FR 15150). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2015. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50- 364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of application for amendment: September 25, 2012; as supplemented on December 20, 2012; September 16, October 30, and November 12, 2013; April 23, May 23, July 3, August 11, August 29, and October 13, 2014; and January 16, 2015. Brief description of amendments: The amendment authorizes the transition of the Joseph M. Farley Nuclear Plant, Units 1 and 2, fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, ``Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition'' (NFPA 805), in accordance with 10 CFR 50.48(c). Date of issuance: March 10, 2015. Effective date: As of its date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: Unit 1-196, Unit 2-192. A publicly-available version is in ADAMS under Accession No. ML14308A048, documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: March 12, 2013 (78 FR 15750). The supplemental letters dated September 16, October 30, and November 12, 2013; April 23, May 23, July 3, August 11, August 29, and October 13, 2014; and January 16, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 10, 2015. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 23rd day of March 2015. [[Page 17109]] For the Nuclear Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2015-07192 Filed 3-30-15; 8:45 am] BILLING CODE 7590-01-P
![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17083
The Commission has determined for NUCLEAR REGULATORY Regulation, U.S. Nuclear Regulatory
these amendments that the application COMMISSION Commission, Washington DC 20555–
complies with the standards and 0001; telephone: 301–415–1506, email:
[NRC–2015–0073]
requirements of the Atomic Energy Act Kay.Goldstein@nrc.gov.
of 1954, as amended (the Act), and the Biweekly Notice; Applications and SUPPLEMENTARY INFORMATION:
Commission’s rules and regulations. Amendments to Facility Operating I. Obtaining Information and
The Commission has made appropriate Licenses and Combined Licenses Submitting Comments
findings as required by the Act and the Involving No Significant Hazards
Commission’s rules and regulations in Considerations A. Obtaining Information
10 CFR Chapter I, which are set forth in AGENCY: Nuclear Regulatory Please refer to Docket ID NRC–2015–
the license amendment. Commission. 0073 when contacting the NRC about
A notice of consideration of issuance ACTION: Biweekly notice.
the availability of information for this
of amendment to facility operating action. You may obtain publicly-
license or combined license, as SUMMARY: Pursuant to Section 189a. (2) available information related to this
applicable, proposed no significant of the Atomic Energy Act of 1954, as action by any of the following methods:
hazards consideration determination, amended (the Act), the U.S. Nuclear • Federal Rulemaking Web site: Go to
and opportunity for a hearing in Regulatory Commission (NRC) is http://www.regulations.gov and search
publishing this regular biweekly notice. for Docket ID NRC–2015–0073.
connection with these actions, was
The Act requires the Commission to • NRC’s Agencywide Documents
published in the Federal Register on Access and Management System
publish notice of any amendments
October 14, 2014 (79 FR 61662). The issued, or proposed to be issued and (ADAMS): You may obtain publicly-
September 23, 2014, application grants the Commission the authority to available documents online in the
revision, and the October 30 and issue and make immediately effective ADAMS Public Documents collection at
November 6, 2014, supplements had no any amendment to an operating license http://www.nrc.gov/reading-rm/
effect on the no significant hazards or combined license, as applicable, adams.html. To begin the search, select
consideration determination, and no upon a determination by the ‘‘ADAMS Public Documents’’ and then
comments were received during the 60- Commission that such amendment select ‘‘Begin Web-based ADAMS
day comment period. involves no significant hazards Search.’’ For problems with ADAMS,
The Commission has determined that consideration, notwithstanding the please contact the NRC’s Public
these amendments satisfy the criteria for pendency before the Commission of a Document Room (PDR) reference staff at
categorical exclusion in accordance request for a hearing from any person. 1–800–397–4209, 301–415–4737, or by
This biweekly notice includes all email to pdr.resource@nrc.gov. The
with 10 CFR 51.22. Therefore, pursuant
notices of amendments issued, or ADAMS accession number for each
to 10 CFR 51.22(b), no environmental
proposed to be issued from March 5, document referenced (if it is available in
impact statement or environmental 2015 to March 18, 2015. The last
assessment need be prepared for these ADAMS) is provided the first time that
biweekly notice was published on it is mentioned in the SUPPLEMENTARY
amendments. March 17, 2015. INFORMATION section.
IV. Conclusion DATES: Comments must be filed by April • NRC’s PDR: You may examine and
30, 2015. A request for a hearing must purchase copies of public documents at
Using the reasons set forth in the be filed by June 1, 2015. the NRC’s PDR, Room O1–F21, One
combined safety evaluation, the staff ADDRESSES: You may submit comments White Flint North, 11555 Rockville
granted the exemption and issued the by any of the following methods (unless Pike, Rockville, Maryland 20852.
amendment that the licensee requested this document describes a different
B. Submitting Comments
on August 22, 2014, and revised by method for submitting comments on a
letter dated September 23, 2014, and specific subject): Please include Docket ID NRC–2015–
supplemented by letters dated October • Federal Rulemaking Web site: Go to 0073, facility name, unit number(s),
30 and November 6, 2014. The http://www.regulations.gov and search application date, and subject in your
exemption and amendment were issued for Docket ID NRC–2015–0073. Address comment submission.
on December 23, 2014 as part of a questions about NRC dockets to Carol The NRC cautions you not to include
combined package to the licensee Gallagher; telephone: 301–415–3463; identifying or contact information that
(ADAMS Accession No. ML14323A609). email: Carol.Gallagher@nrc.gov. For you do not want to be publicly
technical questions, contact the disclosed in your comment submission.
Dated at Rockville, Maryland, this 23rd day individual listed in the FOR FURTHER The NRC posts all comment
of March 2015. INFORMATION CONTACT section of this submissions at http://
For the Nuclear Regulatory Commission. document. www.regulations.gov as well as entering
Lawrence Burkhart, • Mail comments to: Cindy Bladey, the comment submissions into ADAMS.
Chief, Licensing Branch 4, Division of New
Office of Administration, Mail Stop: The NRC does not routinely edit
Reactor Licensing, Office of New Reactors. OWFN–12–H08, U.S. Nuclear comment submissions to remove
[FR Doc. 2015–07277 Filed 3–30–15; 8:45 am]
Regulatory Commission, Washington, identifying or contact information.
DC 20555–0001. If you are requesting or aggregating
asabaliauskas on DSK5VPTVN1PROD with NOTICES
BILLING CODE 7590–01–P For additional direction on obtaining comments from other persons for
information and submitting comments, submission to the NRC, then you should
see ‘‘Obtaining Information and inform those persons not to include
Submitting Comments’’ in the identifying or contact information that
SUPPLEMENTARY INFORMATION section of they do not want to be publicly
this document. disclosed in their comment submission.
FOR FURTHER INFORMATION CONTACT: Kay Your request should state that the NRC
Goldstein, Office of Nuclear Reactor does not routinely edit comment
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![17086 Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices
available version is in ADAMS under Appropriate calculations were developed to Corporation, 526 South Church Street—
Accession No. ML14330A327. substantiate the revised TS parameters EC07H, Charlotte, NC 28202.
Description of amendment request: proposed in this LAR. There will be no NRC Branch Chief: Robert J.
impact on the source term or pathways Pascarelli.
The proposed amendments would
assumed in accidents previously evaluated.
modify the Technical Specifications No analysis assumptions will be violated and Duke Energy Carolinas, LLC, Docket
(TS) to correct non-conservative there will be no adverse effects on onsite or Nos. 50–269, 50–270, and 50–287,
setpoints. Specifically, modify the offsite doses as the result of an accident. Oconee Nuclear Station, Units 1, 2, and
Allowable Value parameter and the Adoption of the TSTF–493 footnotes for the 3, Oconee County, South Carolina
Nominal Trip Setpoint for the TS 3.3.2 respective SRs will ensure that the function’s
Table 3.3.2–1, ‘‘Engineered Safety channels will continue to behave in Date of amendment request: March
Feature Actuation System accordance with safety analysis assumptions 14, 2014. A publicly-available version is
Instrumentation’’ function for Auxiliary and the channel performance assumptions in in ADAMS under Accession No.
the setpoint methodology. ML14078A037.
Feedwater Loss of Offsite Power Therefore, the proposed amendments do
(Function 6.d.) and for the TS 3.3.5 Loss Description of amendment request:
not involve a significant increase in the The amendment would revise the
of Voltage function in Surveillance probability or consequences of an accident
Technical Specifications (TS) for the
Requirement (SR) 3.3.5.2. As part of the previously evaluated.
2. Does the proposed amendment create Inservice Testing Program to reflect the
change, the licensee is also proposing to
the possibility of a new or different kind of current edition of the American Society
add the applicable footnotes in
accident from any accident previously of Mechanical Engineers (ASME) Code
accordance with TSTF–493, Revision 4,
evaluated? that is referenced in 10 CFR 50.55a(b).
‘‘Clarify Application of Setpoint Basis for proposed no significant
Response: No.
Methodology for LSSS [limiting safety The proposed amendments do not change hazards consideration determination:
system set point] Functions.’’ the methods governing normal plant As required by 10 CFR 50.91(a), the
Basis for proposed no significant operation; nor are the methods utilized to licensee has provided its analysis of the
hazards consideration determination: respond to plant transients altered. In
issue of no significant hazards
As required by 10 CFR 50.91(a), the addition, the proposed changes to the
affected TS parameters and the adoption of consideration, which is presented
licensee has provided its analysis of the below:
issue of no significant hazards the TSTF–493 footnotes will not create the
potential for any new initiating events or 1. Does the proposed change involve a
consideration, which is presented below
transients to occur in the actual physical significant increase in the probability or
and staff’s changes/additions are plant. consequences of an accident previously
provided in [ ]: Therefore, the proposed amendments do evaluated?
1. Does the proposed amendment involve not create the possibility of a new or different Response: No.
a significant increase in the probability or kind of accident from any accident The proposed change corrects a
consequences of an accident previously previously evaluated. typographical error in TS 5.5.8, ‘‘Reactor
evaluated? 3. Does the proposed amendment involve Coolant Pump Flywheel Inspection
Response: No. a significant reduction in the margin of Program,’’ and revises TS 5.5.9, ‘‘lnservice
Duke Energy requests NRC review and safety? Testing Program,’’ for consistency with the
approval to revise the Allowable Value Response: No. requirements of 10 CFR 50.55a(f)(4) regarding
parameter for the Technical Specification Margin of safety is related to the the inservice testing of pumps and valves
(TS) 3.3.2 Table 3.3.2–1, ‘‘Engineered Safety confidence in the ability of the fission which are classified as ASME Code Class 1,
Feature Actuation System Instrumentation’’ product barriers to perform their design Class 2 and Class 3. The proposed change
function for Auxiliary Feedwater Loss of functions during and following an accident. incorporates revisions to the ASME Code that
Offsite Power (Function 6.d.) and for the TS These barriers include the fuel cladding, the result in a net improvement in the measures
3.3.5 Loss of Voltage function in Surveillance reactor coolant system, and the containment for testing pumps and valves.
Requirement (SR) 3.3.5.2 in order to make system. The proposed changes will assure the The proposed change does not impact any
this parameter more restrictive. The existing acceptable operation of the affected function accident initiators or analyzed events or
parameter was determined to be non- under all postulated transient and accident assumed mitigation of accident or transient
conservative and this parameter is presently conditions. This will ensure that all events. The proposed change does not
classified as Operable But Degraded in the applicable design and safety limits are involve the addition or removal of any
satisfied such that the fission product equipment, or any design changes to the
Catawba Corrective Action Program. In
barriers will continue to perform their design facility.
addition, the Nominal Trip Setpoint
functions. Therefore, the proposed change does not
parameter for this function is being slightly
Therefore, the proposed amendments do involve a significant increase in the
lowered in order to gain additional margin.
not involve a significant reduction in a probability or consequences of an accident
Finally, as part of this License Amendment previously evaluated.
Request (LAR), applicable footnotes are also margin of safety.
Based on the preceding discussion, Duke 2. Does the proposed change create the
being added to the affected TS 3.3.2 function possibility of a new or different kind of
in accordance with TS Task Force Traveler Energy concludes that the proposed
amendments do not involve a significant accident from any accident previously
[(TSTF)] TSTF–493, Revision 4, ‘‘Clarify evaluated?
Application of Setpoint Methodology for hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, Response: No.
LSSS Functions.’’ The more restrictive The proposed change corrects a
Allowable Value will preclude the potential accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified. typographical error in TS 5.5.8, ‘‘Reactor
for a double sequencing event to occur under Coolant Pump Flywheel Inspection
the condition of a Loss of Coolant Accident The NRC staff has reviewed the Program,’’ and revises TS 5.5.9, ‘‘lnservice
(LOCA) load sequencer actuation with a pre- licensee’s analysis and, based on this
asabaliauskas on DSK5VPTVN1PROD with NOTICES
Testing Program,’’ for consistency with the
existing degraded voltage condition on the review, it appears that the three requirements of 10 CFR 50.55a(f)(4) regarding
essential buses. These proposed changes will standards of 10 CFR 50.92(c) are the inservice testing of pumps and valves
not increase the probability of occurrence of which are classified as ASME Code Class 1,
any design basis accident since the affected
satisfied. Therefore, the NRC staff
Class 2 and Class 3. The proposed change
function, in and of itself, cannot initiate an proposes to determine that the incorporates revisions to the ASME Code that
accident. Should a LOCA occur, the amendment request involves no result in a net improvement in the measures
proposed changes will ensure that the significant hazards consideration. for testing pumps and valves.
sequencer operates properly in order to Attorney for licensee: Lara S. Nichols, The proposed change does not involve a
mitigate the consequences of the event. Associate General Counsel, Duke Energy modification to the physical configuration of
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![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17087
the plant (i.e., no new equipment will be Basis for proposed no significant 3. Does the proposed change involve a
installed), nor does it involve a change in the hazards consideration determination: significant reduction in a margin of safety?
methods governing normal plant operation. As required by 10 CFR 50.91(a), the Response: No.
The proposed change will not impose any The MCPR SLs have been evaluated in
licensee has provided its analysis of the
new or different requirements or introduce a accordance with Global Nuclear Fuels NRC-
new accident initiator, accident precursor, or
issue of no significant hazards approved cycle-specific safety limit
malfunction mechanism. Additionally, there consideration, which is presented methodology to ensure that during normal
is no change in the types or increases in the below: operation and during AOO’s, at least 99.9%
amounts of any effluent that may be released 1. Does the proposed change involve a of the fuel rods in the core are not expected
offsite and there is no increase in individual significant increase in the probability or to experience transition boiling. The
or cumulative occupational exposure. consequences of an accident previously proposed change to the [MCPR SL] accounts
Therefore, the proposed change does not evaluated? for requirements specified in the NRC Safety
create the possibility of a new or different Response: No. Evaluation limitations and conditions
kind of accident from any accident The Bases to TS 2.1.1.2 states that: ‘‘The associated with NEDC–33173P and NEDC–
previously evaluated. MCPR SL ensures sufficient conservatism in 33006P, which result in additional margin
3. Does the proposed change involve a the operating MCPR limit that, in the event above that specified in the TS Bases.
significant reduction in a margin of safety? of an AOO [Anticipated Operational Therefore, the proposed change to the
Response: No. Occurrence] from the limiting condition of MCPR SL does not involve a significant
The proposed change corrects a operation, at least 99.9% of the fuel rods in reduction in a margin of safety.
typographical error in TS 5.5.8, ‘‘Reactor the core would be expected to avoid boiling
Coolant Pump Flywheel Inspection
The NRC staff has reviewed the
transition.
Program,’’ and revises TS 5.5.9, ‘‘lnservice This condition is met in that the GGNS licensee’s analysis and, based on this
Testing Program,’’ for consistency with the Cycle 20 (C20) MCPR SL evaluation was review, it appears that the three
requirements of 10 CFR 50.55a(f)(4) regarding performed in accordance with Reference 4 standards of 10 CFR 50.92(c) are
the inservice testing of pumps and valves [NEDE–24011–P–A, ‘‘General Electric satisfied. Therefore, the NRC staff
which are classified as ASME Code Class 1, Standard Application for Reactor Fuel proposes to determine that the
Class 2 and Class 3. The proposed change (GESTAR–II’’)]. The resulting values amendment request involves no
incorporates revisions to the ASME Code that continue to ensure the conservatism significant hazards consideration.
result in a net improvement in the measures described in the Bases to TS 2.1.1.2. The Attorney for licensee: Joseph A.
for testing pumps and valves. The safety proposed changes also continue to ensure
sufficient conservatism in the operating
Aluise, Associate General Counsel—
function of the affected pumps and valves
will be maintained. Therefore, the proposed MCPR limit. The MCPR operating limits are Nuclear, Entergy Services, Inc., 639
change does not involve a significant presented and controlled in accordance with Loyola Avenue, New Orleans, Louisiana
reduction in a margin of safety. the GGNS Core Operating Limits Report 70113.
(COLR). NRC Branch Chief: Meena K. Khanna.
The NRC staff has reviewed the The requested Technical Specification
licensee’s analysis and, based on this change does not involve any plant Entergy Operations, Inc., System Energy
review, it appears that the three modifications or operational changes that Resources, Inc., South Mississippi
standards of 10 CFR 50.92(c) are could affect system reliability or performance Electric Power Association, and Entergy
satisfied. Therefore, the NRC staff or that could affect the probability of operator Mississippi, Inc., Docket No. 50–416,
proposes to determine that the error. The requested change does not affect Grand Gulf Nuclear Station, Unit 1
any postulated accident precursors, any (GGNS), Claiborne County, Mississippi
amendment request involves no
accident mitigating systems, or introduce any
significant hazards consideration. new accident initiation mechanisms. Date of amendment request:
Attorney for licensee: Lara S. Nichols, Therefore, the proposed change to increase November 21, 2014, as supplemented by
Deputy General Counsel, Duke Energy the MCPR SL values does not involve a letter dated February 18, 2015. Publicly-
Corporation, 550 South Tryon Street— significant increase in the probability or available versions are in ADAMS under
DEC45A, Charlotte, NC 28202–1802. consequences of an accident previously Accession Nos. ML14325A752 and
NRC Branch Chief: Robert J. evaluated. ML15049A536, respectively.
Pascarelli. 2. Does the proposed change create the
Description of amendment request:
possibility of a new or different kind of
accident from any accident previously The proposed amendment would revise
Entergy Operations, Inc., System Energy
evaluated? GGNS’s license basis to adopt a single
Resources, Inc., South Mississippi
Response: No. fluence methodology.
Electric Power Association, and Entergy
The proposed change does not involve any Basis for proposed no significant
Mississippi, Inc., Docket No. 50–416,
new modes of operation, any changes to hazards consideration determination:
Grand Gulf Nuclear Station, Unit 1 setpoints, or any plant modifications. The As required by 10 CFR 50.91(a), the
(GGNS), Claiborne County, Mississippi proposed change to the MCPR SL accounts licensee has provided its analysis of the
Date of amendment request: for requirements specified in the NRC Safety issue of no significant hazards
November 21, 2014. A publicly- Evaluation limitations and conditions
consideration, which is presented
associated with NEDC–33173P
available version is in ADAMS under [‘‘Applicability of GE Methods to Expanded below:
Accession No. ML14325A520. Operating Domains’’] and NEDC–33006P 1. Does the proposed change involve a
Description of amendment request: [‘‘Licensing Topical Report—General Electric significant increase in the probability or
The amendment would change the Boiling Water Reactor Maximum Extended consequences of an accident previously
GGNS Technical Specification (TS) Load Line Limit Analysis Plus’’]. Compliance evaluated?
2.1.1, ‘‘Reactor Core SLs [Safety with the criterion for incipient boiling Response: No.
asabaliauskas on DSK5VPTVN1PROD with NOTICES
Limits].’’ Specifically, the change would transition continues to be ensured. The core The proposed change adopts a single flux
revise the Minimum Critical Power operating limits will continue to be methodology. While Chapter 15, Accident
developed using NRC approved methods. Analysis, of the Standard Review Plan
Ratio (MCPR) SL stated in TS 2.1.1.2 for
The proposed [MCPR SL] does not result in (NUREG–0800, Standard Review Plan for the
two-loop operation from greater than or the creation of any new precursors to an Review of Safety Analysis Reports for
equal to (≥) 1.11 to ≥ 1.15. Additionally, accident. Nuclear Power Plants) assumes the pressure
the change would revise the MCPR SL Therefore, the proposed change does not vessel does not fail, the flux methodology is
stated in TS 2.1.1.2 for single-loop create of a new or different kind of accident not an initiator to any accident previously
operation from ≥ 1.14 to ≥ 1.15. from any accident previously evaluated. evaluated. Accordingly, the proposed change
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![17088 Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices
to the adoption of the flux methodology has consideration, which is presented kind of accident from any previously
no effect on the probability of any accident below: evaluated.
previously evaluated. 3. Does the Proposed Change Involve a
Therefore, the proposed change does not 1. Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?
involve a significant increase in the Significant Increase in the Probability or Response: No.
probability or consequences of an accident Consequences of an Accident Previously The proposed change continues to ensure
previously evaluated. Evaluated? that the maximum temperature of the cooling
2. Does the proposed change create the Response: No. water supplied to the plant SSCs during a
possibility of a new or different kind of The likelihood of a malfunction of any UHS design basis event remains within the
accident from any accident previously systems, structures or components (SSCs) evaluated equipment limits and capabilities
evaluated? supported by the UHS [ultimate heat sink] is assumed in the accident analysis. The
Response: No. not significantly increased by increasing the proposed change does not result in any
The proposed change adopts a flux allowable Ultimate Heat Sink (UHS) changes to plant equipment function,
methodology. The change does not involve a temperature from ≤100 °F to ≤102 °F. The including setpoints and actuations. All
physical alteration of the plant (i.e., no new UHS provides a heat sink for process and equipment will function as designed in the
or different type of equipment will be operating heat from safety related plant safety analysis without any physical
installed) or a change in the methods components during a transient or accident, as modifications. The proposed change does not
governing normal plant operations. The well as during normal operation. The alter a limiting condition for operation,
change does not alter assumptions made in proposed change does not make any physical limiting safety system setting, or safety limit
the safety analysis regarding fluence. changes to any plant SSCs, nor does it alter specified in the Technical Specifications.
Therefore, the proposed change does not any of the assumptions or conditions upon The proposed change does not adversely
create the possibility of a new or different which the UHS is designed. The UHS is not impact the UHS inventory required to be
kind of accident from any accident an initiator of any analyzed accident. All available for the UFSAR described design
previously evaluated. equipment supported by the UHS has been basis accident involving the worst case 30-
3. Does the proposed change involve a evaluated to demonstrate that their day period including losses for evaporation
significant reduction in a margin of safety? performance and operation remains as and seepage to support safe shutdown and
Response: No. described in the UFSAR [updated final safety cooldown of both Braidwood Station units.
The proposed change adopts a single analysis report] with no increase in Additionally, the structural integrity of the
fluence methodology. The proposed change probability of failure or malfunction. UHS is not impacted and remains acceptable
does not alter the manner in which safety The SSCs credited to mitigate the following the change, thereby ensuring that
limits, limiting safety system settings or consequences of postulated design basis the assumptions for both UHS temperature
limiting conditions for operation are accidents remain capable of performing their and inventory remain valid.
determined. The proposed change ensures design basis function. The change in Therefore, since there is no adverse impact
that the methodology used for fluence is in maximum UHS temperature has been of this change on the Braidwood Station
compliance with RG 1.190 requirements. evaluated using the UFSAR described safety analysis, there is no reduction in the
Therefore, the proposed change does not methods to demonstrate that the UHS margin of safety of the plant.
involve a significant reduction in a margin of remains capable of removing normal
safety. operating and post-accident heat. The change The NRC staff has reviewed the
in UHS temperature and resulting licensee’s analysis and, based on this
The NRC staff has reviewed the containment response following a postulated review, it appears that the three
licensee’s analysis and, based on this design basis accident has been demonstrated standards of 10 CFR 50.92(c) are
review, it appears that the three to not be impacted. Additionally, all the UHS satisfied. Therefore, the NRC staff
standards of 10 CFR 50.92(c) are supported equipment, credited in the proposes to determine that the
satisfied. Therefore, the NRC staff accident analysis to mitigate an accident, has
been shown to continue to perform their
requested amendments involve no
proposes to determine that the significant hazards consideration.
amendment request involves no design function as described in the UFSAR.
Therefore, the proposed change does not Attorney for licensee: Bradley J.
significant hazards consideration. involve a significant increase in the Fewell, Associate General Counsel,
Attorney for licensee: Joseph A. probability or consequences of an accident Exelon Nuclear, 4300 Winfield Road,
Aluise, Associate General Counsel— previously evaluated. Warrenville, IL 60555.
Nuclear, Entergy Services, Inc., 639 2. Does the Proposed Change Create the
Loyola Avenue, New Orleans, Louisiana NRC Branch Chief: Travis L. Tate
Possibility of a New or Different Kind of
70113. Accident from any Accident Previously Exelon Generation Company, LLC
NRC Branch Chief: Meena K. Khanna. Evaluated? (EGC), Docket Nos. STN 50–454 and
Response: No. STN 50–455, Byron Station, Units 1 and
Exelon Generation Company, LLC, The proposed change does not create the
Docket Nos. STN 50–456 and STN 50– 2, Ogle County, Illinois
possibility of a new or different kind of
457, Braidwood Station, Units 1 and 2, accident from any accident previously Date of amendment request:
Will County, Illinois evaluated. The proposed change does not November 24, 2014. A publicly-
introduce any new modes of plant operation, available version is in ADAMS under
Date of amendment request: August change the design function of any SSC,
19, 2014. A publicly-available version is Accession No. ML14328A800.
change the mode of operation of any SSC, or
in ADAMS under Accession No. change any actions required when the TS Description of amendment request:
ML14231A902. limit is exceeded. There are no new The proposed amendment would revise
Description of amendment request: equipment failure modes or malfunctions Condition I and surveillance
The proposed amendment would created as affected SSCs continue to operate requirement (SR) 3.7.9.3 associated with
increase the technical specification (TS) in the same manner as previously evaluated technical specification (TS) Section
asabaliauskas on DSK5VPTVN1PROD with NOTICES
surveillance requirement (SR) 3.7.9.2 and have been evaluated to perform as 3.7.9, ‘‘Ultimate Heat Sink (UHS),’’ to
allowable temperature to less than or designed at the increased UHS temperature reflect the current design basis flood
and as assumed in the accident analysis.
equal to 102 °F [degree Fahrenheit]. Additionally, accident initiators remain as
level.
Basis for proposed no significant described in the UFSAR and no new accident Basis for proposed no significant
hazards consideration determination: initiators are postulated as a result of the hazards consideration determination:
As required by 10 CFR 50.91(a), the increase in UHS temperature. As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the Therefore, the proposed change does not licensee has provided its analysis of the
issue of no significant hazards create the possibility of a new or different issue of no significant hazards
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![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17089
consideration, which is presented event flood level and TS 3.7.9, Condition I licensee has provided its analysis of the
below: action level for high river level. The issue of no significant hazards
proposed change will ensure the operability consideration, which is presented
EGC has evaluated whether or not a of the SX makeup pumps to meet TS 3.7.9
significant hazards consideration is involved below:
LCO and do not affect the ability of the SX
with the proposed amendment by focusing makeup pumps to provide the safety related EGC [Exelon Generation Company] has
on the three standards set forth in 10 CFR source makeup to the UHS. evaluated the proposed changes, using the
50.92(c), ‘‘Issuance of amendment,’’ as Therefore, the proposed change does not criteria in 10 CFR 50.92, and has determined
discussed below: involve a significant reduction in a margin of that the proposed changes do not involve a
1. Does the proposed amendment involve significant hazards consideration. The
safety.
a significant increase in the probability or following information is provided to support
Based on the above, EGC concludes that
consequences of an accident previously a finding of no significant hazards
the proposed amendment does not involve a
evaluated? consideration.
significant hazards consideration under the
Response: No. 1. Does the proposed change involve a
standards set forth in 10 CFR 50.92(c), and
The proposed changes to revise TS 3.7.9, significant increase in the probability or
accordingly, a finding of no significant
Condition I and SR 3.7.9.3 will ensure the consequences of an accident previously
hazards consideration is justified.
operability of the SX [service water] makeup evaluated?
pumps to meet TS 3.7.9 LCO [Limiting The NRC staff has reviewed the Response: No.
Condition for Operation] requirement. The licensee’s analysis and, based on this The proposed changes will not result in a
proposed change does not result in any review, it appears that the three significant change in the stored energy in the
physical changes to safety related structures, standards of 10 CFR 50.92(c) are reactor vessel during the performance of the
systems, or components. The probability of a testing. The probability of an accident is not
flood at the river screen house (RSH) is satisfied. Therefore, the NRC staff
proposes to determine that the significantly increased because the proposed
unchanged. Since the UHS itself is not an changes will not alter the method by which
accident initiator, the proposed change does requested amendments involve no
inservice leakage and hydrostatic testing is
not impact the initiators or assumptions of significant hazards consideration. performed or significantly change the
analyzed accidents, nor do they impact the Attorney for licensee: Bradley J. temperatures and pressures achieved to
mitigation of accidents or transient events. Fewell, Associate General Counsel, perform the test.
Consequently, the proposed change does not Exelon Nuclear, 4300 Winfield Road, The consequences of previously evaluated
increase the probability of occurrence for any Warrenville, IL 60555. accidents are not significantly increased
accident previously evaluated. NRC Branch Chief: Travis L. Tate. because the required testing conditions
The proposed change will ensure that
provide adequate assurance that the
actions to verify operability of the deep well Exelon Generation Company, LLC, consequences of a steam leak will be
pumps will be taken prior to the potential for Docket Nos. 50–237 and 50–249, conservatively bounded by the consequences
the SX makeup pumps to be adversely Dresden Nuclear Power Station, Units 2 of the postulated main system line break
affected by the combined event flood high
river level. Therefore, the UHS will be
and 3, Grundy County, Illinois outside of primary containment. Under these
proposed changes, the secondary
capable of performing its functions to Exelon Generation Company, LLC, containment, standby gas treatment system,
mitigate accidents by serving as the heat sink Docket Nos. 50–373 and 50–374, LaSalle
for safety related equipment. Thus, the and associated initiation instrumentation are
County Station, Units 1 and 2, LaSalle required to be operable during the
proposed change does not increase the
County, Illinois performance of inservice leakage and
consequences of any accident previously
evaluated. Exelon Generation Company, LLC, hydrostatic testing and would be capable of
Therefore, the proposed change does not mitigating any airborne radioactivity or steam
Docket Nos. 50–254 and 50–265, Quad leaks that could occur. In addition, the
involve a significant increase in the Cities Nuclear Power Station, Units 1
probability or consequences of an accident required Emergency Core Cooling subsystems
and 2, Rock Island County, Illinois will be more than adequate to ensure that a
previously evaluated.
2. Does the proposed amendment create Date of amendment request: significant increase in consequences will not
the possibility of a new or different kind of December 22, 2014. A publicly-available occur by ensuring that the potential for failed
accident from any accident previously fuel and a subsequent increase in coolant
version is in ADAMS under Accession
evaluated? activity above Technical Specification limits
No. ML14357A085. are minimized.
Response: No. Description of amendment request:
The proposed change to revise TS 3.7.9, Therefore, the proposed changes do not
The proposed amendment modifies the involve a significant increase in the
Condition I and SR 3.7.9.3 does not change
the design function or operation of the SX technical specifications (TSs) to add a probability or consequences of an accident
makeup pumps. The proposed change does new Limiting Condition for Operation previously evaluated.
not change or introduce the possibility of any (LCO) 3.10.8 to specifically permit 2. Does the proposed change create the
new or different type of equipment, modes of inservice leakage and hydrostatic testing possibility of a new or different kind of
system operation, failure mechanisms, at reactor coolant system (RCS) accident from any accident previously
malfunctions, or accident initiators. The temperatures greater than the average evaluated?
proposed change to lower the river level Response: No.
reactor coolant temperature for MODE 4 As the accumulated neutron fluence on the
value at which action is taken to verify basin
with the reactor shutdown. In addition, reactor vessel increases, the Pressure-
levels and deep well pumps are ready to
perform the UHS makeup function in the the proposed amendment includes an Temperature Limits in TS 3.4.9 for DNPS
place of the SX makeup pumps will not affect expanded scope of LCO 3.10.8 [Dresden Nuclear Power Station] and QCNPS
the operation or function of the UHS or the consistent with the NRC-approved [Quad Cities Nuclear Power Station and TS
deep well pumps. Revision 0 of Technical Specification [technical specification] 3.4.11 for LSCS
Therefore, the proposed change does not Task Force (TSTF) Improved Standard [LaSalle County Station] may eventually
asabaliauskas on DSK5VPTVN1PROD with NOTICES
create the possibility of a new or different Technical Specification Change require that inservice leakage and hydrostatic
kind of accident from any accident Traveler, TSTF–484, ‘‘Use of TS 3.10.1 testing be conducted at RCS [reactor coolant
previously evaluated. system] temperatures greater than the average
for Scram Time Testing Activities’’ reactor coolant temperature for MODE 4 with
3. Does the proposed amendment involve
a significant reduction in a margin of safety? available in ADAMS under Accession the reactor shutdown. However, even with
Response: No. No. ML062990425. the required minimum reactor coolant
The proposed change to revise TS 3.7.9, Basis for proposed no significant temperatures less than or equal to the average
Condition I and SR 3.7.9.3 reestablishes the hazards consideration determination: reactor coolant temperature for MODE 4 with
margin between the design bases combined As required by 10 CFR 50.91(a), the the reactor shutdown, maintaining RCS
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![17090 Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices
temperatures within a small band during current Note allows the licensee to does not involve a significant hazards
testing can be impractical. The proposed consider the low pressure coolant consideration under the standards set forth in
changes will not result in a significant injection (LPCI) subsystem associated 10 CFR 50.92(c), and, according a finding of
change in the stored energy in the reactor no significant hazards consideration is
with the residual heat removal (RHR)
vessel during the performance of the testing justified.
nor will it alter the way inservice leakage and system to be OPERABLE under
specified conditions. The NRC staff has reviewed the
hydrostatic testing is performed or
significantly change the temperatures and Basis for proposed no significant licensee’s analysis and, based on this
pressures achieved to perform the testing. hazards consideration determination: review, it appears that the three
Therefore, the proposed changes do not As required by 10 CFR 50.91(a), the standards of 10 CFR 50.92(c) are
create the possibility of a new or different licensee has provided its analysis of the satisfied. Therefore, the NRC staff
kind of accident from any previously issue of no significant hazards proposes to determine that the
evaluated. consideration, which is presented amendment request involves no
3. Does the proposed change involve a significant hazards consideration.
significant reduction in a margin of safety?
below:
Attorney for licensee: Tamra Domeyer,
Response: No. 1. Does the proposed amendment involve
The proposed changes and additions result a significant increase in the probability or
Exelon Generation Company, LLC, 4300
in increased system operability requirements consequences of an accident previously Winfield Road, Warrenville, IL, 60555.
above those that currently exist during the evaluated? Branch Chief: Travis L. Tate.
performance of inservice leakage and Response: No. FirstEnergy Nuclear Operating
hydrostatic testing. The incremental increase No physical changes to the facility will
Company, Docket No. 50–346, Davis-
in stored energy in the vessel during testing occur as a result of this proposed
will be conservatively bounded by the amendment. The proposed change will not Besse Nuclear Power Station, Unit 1,
consequences of the postulated main steam alter the physical design. Current TS note Ottawa County, Ohio
line break outside of primary containment could make LSCS susceptible to potential Date of amendment request:
and analyzed margins of safety are water hammer in the RHR system if in the December 31, 2014. A publicly-available
unchanged. SDC [shutdown cooling] Mode of RHR in version is in ADAMS under Accession
Therefore, the proposed changes do not Mode 3 when swapping from the SDC to
No. ML14365A080.
involve a significant reduction in a margin of LPCI mode of RHR. The proposed LAR
safety.
Description of amendment request:
[license amendment request] will eliminate
EGC has reviewed the no significant the risk for cavitation of the pump and The proposed amendment would revise
hazards determination published on August voiding in the suction piping, thereby the frequency for the technical
21, 2006 (71 FR 48561) [for Technical avoiding potential to damage the RHR specification surveillance to verify that
Specification Task Force traveler TSTF–484]. system, including water hammer. each containment spray system nozzle
The no significant hazards determination was Therefore, the proposed change does not is unobstructed from a frequency of 10
made available on October 27, 2006 (71 FR involve a significant increase in the years to an event-based frequency.
63050) as part of the CLIIP [Consolidated probability or consequences of an accident Basis for proposed no significant
Line Item Improvement Process] Notice of previously evaluated. hazards consideration determination:
Availability. EGC has concluded that the 2. Does the proposed amendment create As required by 10 CFR 50.91(a), the
determination presented in the notice is the possibility of a new or different kind of
applicable to DNPS, Units 2 and 3; LSCS, accident from any accident previously
licensee has provided its analysis of the
Units 1 and 2; and QCNPS, Units 1 and 2; evaluated? issue of no significant hazards
and the determination is hereby incorporated Response: No. consideration, which is presented
by reference to satisfy the requirements of 10 The proposed change does not alter the below:
CFR 50.91(a). physical design, safety limits, or safety 1. Does the proposed change involve a
analysis assumptions associated with the significant increase in the probability or
The NRC staff has reviewed the
operation of the plant. Accordingly, the consequences of an accident previously
licensee’s analysis and, based on this change does not introduce any new accident
review, it appears that the three evaluated?
initiators, nor does it reduce or adversely Response: No.
standards of 10 CFR 50.92(c) are affect the capabilities of any plant structure, The containment spray system and its
satisfied. Therefore, the NRC staff system, or component to perform their safety spray nozzles are not accident initiators and
proposes to determine that the function. Deletion of the TS note is therefore the proposed change does not
requested amendments involve no appropriate because current TSs could put involve a significant increase in the
significant hazards consideration. the plant at risk for potential cavitation of the probability of an accident. The revised
Attorney for licensee: Bradley Fewell, pump and voiding in the suction piping, surveillance requirement will require event-
Associate General Counsel, Exelon resulting in potential to damage the RHR based frequency verification in lieu of a fixed
system, including water hammer. frequency verification. The proposed change
Generation Company, LLC, 4300
Therefore, the proposed change does not does not have a detrimental impact on the
Winfield Road, Warrenville, IL 60555. create the possibility of a new or different integrity of any plant structure, system, or
NRC Branch Chief: Travis L. Tate. kind of accident from any previously component that may initiate an analyzed
Exelon Generation Company, LLC, evaluated. event. The proposed change will not alter the
Docket No. 50–373 and 50–374, LaSalle 3. Does the proposed amendment involve operation or otherwise increase the failure
a significant reduction in a margin of safety? probability of any plant equipment that can
County Station (LSCS), Units 1 and 2, Response: No. initiate an analyzed accident. Because the
LaSalle County, Illinois The proposed change conforms to NRC system will continue to be available to
Date of Amendment Request: January regulatory guidance regarding the content of perform its accident mitigation function, the
12, 2015. A publicly-available version is plant Technical Specifications. The proposed consequences of accidents previously
asabaliauskas on DSK5VPTVN1PROD with NOTICES
in ADAMS under Accession No. change does not alter the physical design, evaluated are not significantly increased.
ML15012A544. safety limits, or safety analysis assumptions Therefore, the proposed change does not
Description of amendment request: associated with the operation of the plant. involve a significant increase in the
Therefore, the proposed change does not probability or consequences of an accident
The proposed amendment would delete involve a significant reduction in a margin of previously evaluated.
the limiting condition for operation safety. 2. Does the proposed change create the
(LCO) Note for Technical Specification Based on the above evaluation, EGC possibility of a new or different kind of
(TS) Section 3.5.1, ‘‘ECCS [emergency [Exelon Generation Company, LLC] accident from any accident previously
core cooling system]—Operating.’’ The concludes that the proposed amendment evaluated?
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![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17091
Response: No. for Type C valves up to 75 months, and system, or component of the plant to perform
The proposed change will not physically corrects an editorial issue in the TS. a design function. In addition, this change
alter the plant (no new or different type of Basis for proposed no significant does not affect safety limits, limiting safety
equipment will be installed) or change the hazards consideration determination: system setpoints, or limiting conditions for
methods governing normal plant operation. operation. The specific requirements and
As required by 10 CFR 50.91(a), the
The proposed change does not introduce new conditions of the Technical Specification
accident initiators or impact assumptions licensee has provided its analysis of the Containment Leakage Rate Testing Program
made in the safety analysis. Testing issue of no significant hazards exist to ensure that the degree of containment
requirements continue to demonstrate that consideration, which is presented structural integrity and leak-tightness that is
the limiting conditions for operation are met below: considered in the plant safety analysis is
and the system components are functional. 1. Does the proposed amendment involve maintained.
Therefore, the proposed change does not a significant increase in the probability or The overall containment leak rate limit
create the possibility of a new or different consequences of an accident previously specified by Technical Specifications is
kind of accident from any previously evaluated? maintained, thus ensuring the margin of
evaluated. Response: No. safety in the plant safety analysis is
3. Does the proposed change involve a The proposed amendment adopts the NRG- maintained. The design, operation, testing
significant reduction in a margin of safety? accepted guidelines of [Nuclear Energy methods, and acceptance criteria for Type A,
Response: No. Institute] NEI 94–01, Revision 3–A, ‘‘Industry Type B, and Type C containment leakage
The safety function of the CSS Guideline for Implementing Performance- tests specified in applicable codes and
[containment spray system] is to spray water Based Option of 10 CFR part 50, Appendix standards would continue to be met with the
into the containment atmosphere in the event J,’’ for [Davis-Besse Nuclear Power Station] acceptance of this proposed change, since
of a loss-of-coolant accident to prevent DBNPS performance-based Type C these are not affected by this revision to the
containment pressure from exceeding the containment isolation valve testing. Revision performance-based containment testing
design value and to remove fission products 3–A of NEI 94–01 allows, based on previous program.
from the containment atmosphere. valve leak test performance, an extension of Therefore, the proposed amendment does
The CSS is not susceptible to corrosion- Type C containment isolation valve leak test not involve a significant reduction in a
induced obstruction or obstruction from intervals. Since the change involves only margin of safety.
sources external to the system. Maintenance performance-based Type C testing, the The NRC staff has reviewed the
activities that unexpectedly introduce proposed amendment does not involve either
unretrievable foreign material into the system licensee’s analysis and, based on this
a physical change to the plant or a change in
would require subsequent verification to the manner in which the plant is operated or review, it appears that the three
ensure there is no nozzle blockage. The spray controlled. standards of 10 CFR 50.92(c) are
header nozzles are expected to remain Implementation of these guidelines satisfied. Therefore, the NRC staff
unblocked and available in the event that a continues to provide adequate assurance that proposes to determine that the
safety function is required. Therefore, the during design basis accidents, the amendment request involves no
capacity of the system would remain components of the primary containment significant hazards consideration.
unaffected. The proposed change does not system will limit leakage rates to less than Attorney for licensee: David W.
relax any criteria used to establish safety the values assumed in the plant safety Jenkins, Attorney, FirstEnergy
limits and will not relax any safety system analyses.
Corporation, Mail Stop A–GO–15, 76
settings. The safety analysis acceptance The proposed amendment will not change
criteria are not affected by this change. the leakage rate acceptance requirements. As South Main Street, Akron, OH 44308.
Therefore, the proposed change does not such, the containment will continue to NRC Branch Chief: Travis L. Tate.
involve a significant reduction in the margin perform its design function as a barrier to Indiana Michigan Power Company
of safety. fission product releases. (IandM), Docket Nos. 50–315 and 50–
Therefore, the proposed amendment does
The NRC staff has reviewed the not significantly increase the probability or 316, Donald C. Cook Nuclear Plant,
licensee’s analysis and, based on this consequences of an accident previously Units 1 and 2, Berrien County, Michigan
review, it appears that the three evaluated. Date of amendment request:
standards of 10 CFR 50.92(c) are 2. Does the proposed amendment create November 14, 2014, as supplemented by
satisfied. Therefore, the NRC staff the possibility of a new or different kind of a letter dated February 12, 2015.
proposes to determine that the accident from any accident previously
evaluated?
Publicly-available versions are in
amendment request involves no ADAMS under Accession Nos.
Response: No.
significant hazards consideration. The proposed amendment to revise the ML14324A209, and ML15050A247,
Attorney for licensee: David W. extended frequency performance-based Type respectively.)
Jenkins, Attorney, FirstEnergy C testing program does not change the design Description of amendment request:
Corporation, Mail Stop A–GO–15, 76 or operation of structures, systems, or The proposed amendments would
South Main Street, Akron, OH 44308. components of the plant. replace the current Donald C. Cook
NRC Branch Chief: Travis L. Tate. The proposed amendment would continue Nuclear Plant (CNP) Units 1 and 2
to ensure containment operability and would technical specifications (TSs) limit on
FirstEnergy Nuclear Operating ensure operation within the bounds of
existing accident analyses. There are no reactor coolant system (RCS) gross
Company, et al., Docket No. 50–346,
accident initiators created or affected by the specific activity with a new limit on
Davis-Besse Nuclear Power Station, Unit
proposed amendment. RCS noble gas specific activity. The
1, Ottawa County, Ohio
Therefore, the proposed amendment does noble gas specific activity limit would
Date of amendment request: not create the possibility of a new or different be based on a new DOSE EQUIVALENT
December 19, 2014. A publicly-available kind of accident from any previously XE–133 definition that would replace
asabaliauskas on DSK5VPTVN1PROD with NOTICES
version is in ADAMS under Accession evaluated. the current E-Bar average disintegration
No. ML14353A349. 3. Does the proposed amendment involve energy definition. In addition, the
a significant reduction in a margin of safety?
Description of amendment request: current DOSE EQUIVALENT I–131
Response: No.
The proposed amendment would revise The proposed amendment to revise the definition would be revised to allow the
the technical specifications (TS) to extended frequency performance-based Type use of additional thyroid dose
adopt performance-based Type C testing C testing program does not affect plant conversion factors. The proposed RCS
for the reactor containment, which operations, design functions, or any analysis specific activity changes are consistent
would allow for extended test intervals that verifies the capability of a structure, with NRC-approved Industry Technical
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Specification Task Force (TSTF) increase the probability of an accident 10 CFR 50.67. Thus, by meeting the
Standard Technical Specification previously evaluated in Chapter 14 of the applicable regulatory limits for AST, there is
change traveler, TSTF–490, Revision 0, UFSAR. no significant reduction in a margin of safety.
Based on the AST analyses, there are no Therefore, because the proposed changes
‘‘Deletion of E-Bar Definition and
proposed changes to performance continue to result in dose consequences
Revision to Reactor Coolant System requirements and no proposed revision to the within the applicable regulatory limits, the
Specific Activity Technical parameters or conditions that could proposed amendment does not involve a
Specification,’’ with deviations. contribute to the initiation of an accident significant reduction in margin of safety.
Additionally, the proposed amendments previously discussed in Chapter 14 of the
would revise the CNP Units 1 and 2 UFSAR. Plant-specific radiological analyses The NRC staff has reviewed the
licensing basis and TSs to adopt the have been performed using the AST analysis and, based on this review, it
alternative source term (AST) as methodology and new X/Qs have been appears that the three standards of 10
established. Based on the results of these CFR 50.92(c) are satisfied. Therefore, the
allowed in 10 CFR 50.67. The proposed
analyses, it has been demonstrated that the NRC staff proposes to determine that the
amendments represent full CR [control room] and off-site dose
implementation of the AST as described amendments requested involve no
consequences of the limiting events
in the NRC’s Regulatory Guide 1.183, considered in the analyses meet the
significant hazards consideration.
‘‘Alternative Radiological Source Terms regulatory guidance provided for use with Attorney for licensee: Robert B.
for Evaluating Design Basis Accidents at the AST, and the doses are within the limits Haemer, Senior Nuclear Counsel, One
Nuclear Power Reactors,’’ Revision 0. established by 10 CFR 50.67. Cook Place, Bridgman, MI 49106.
Basis for proposed no significant Therefore, it is concluded that the NRC Branch Chief: David L. Pelton.
hazards consideration determination: proposed amendment does not involve a
significant increase in the probability or the Luminant Generation Company LLC,
As required by 10 CFR 50.91(a), the
consequences of an accident previously Docket Nos. 50–445 and 50–446,
licensee has provided its analysis of the evaluated. Comanche Peak Nuclear Power Plant
issue of no significant hazards Criterion 2—The Proposed Change Does (CPNPP), Units 1 and 2, Somervell
consideration. The licensee concluded Not Create the Possibility of a New or County, Texas
that the no significant hazards Different Kind of Accident from any
consideration determination published Previously Evaluated. Date of amendment request: January
on March 19, 2007 (72 FR 12838), The proposed change in specific activity 28, 2015. A publicly-available version is
‘‘Notice of Availability of the Model limits does not alter any physical part of the in ADAMS under Accession No.
Safety Evaluation,’’ is applicable. This plant nor does it affect any plant operating ML15036A032.
parameter. The change does not create the
determination is presented below, along Description of amendment request:
potential for a new or different kind of
with the licensee’s analysis of the accident from any previously calculated. The amendment would revise Technical
implementation of the AST. No new modes of operation are introduced Specification (TS) 5.5.16, ‘‘Containment
Criterion 1—The Proposed Change Does by the proposed changes. The proposed Leakage Rate Testing Program,’’ for
Not Involve a Significant Increase in the changes will not create any failure mode not CPNPP, Units 1 and 2, to allow an
Probability or Consequences of an Accident bounded by previously evaluated accidents. increase in the 10 CFR part 50,
Previously Evaluated. Implementation of AST and the associated appendix J, ‘‘Primary Reactor
Reactor coolant specific activity is not an proposed TS changes and new X/Qs have no Containment Leakage Testing for Water-
initiator for any accident previously impact to the initiation of any DBAs. These
changes do not affect the design function or
Cooled Power Reactors,’’ Type A
evaluated. The Completion Time when
primary coolant gross activity is not within modes of operation of structures, systems and Integrated Leak Rate Test (ILRT) interval
limit is not an initiator for any accident components in the facility prior to a from a 10-year frequency to a maximum
previously evaluated. The current variable postulated accident. Since structures, of 15 years and the extension of the
limit on primary coolant iodine systems and components are operated no containment isolation valves leakage
concentration is not an initiator to any differently after the AST implementation, no Type C tests from its current 60-month
accident previously evaluated. As a result, new failure modes are created by this frequency to 75 months in accordance
the proposed change does not significantly proposed change. The alternative source term with Nuclear Energy Institute (NEI) 94–
increase the probability of an accident. The change itself does not have the capability to 01, Revision 3–A, ‘‘Industry Guidance
proposed change will limit primary coolant initiate accidents.
noble gases to concentrations consistent with Consequently, the proposed changes do not
for Implementing Performance Based
the accident analyses. The proposed change create the possibility of a new or different Option of 10 CFR part 50, appendix J,’’
to the Completion Time has no impact on the kind of accident from any accident July 2012 (ADAMS Accession No.
consequences of any design basis accident previously evaluated. ML12221A202), and conditions and
since the consequences of an accident during Criterion 3—The Proposed Change Does limitations specified in NEI 94–01,
the extended Completion Time are the same Not Involve a Significant Reduction in a Revision 2–A, ‘‘Industry Guidance for
as the consequences of an accident during Margin of Safety. Implementing Performance Based
the Completion Time. As a result, the The proposed change revises the limits on Option of 10 CFR part 50, appendix J,’’
consequences of any accident previously noble gas radioactivity in the primary October 2008 (ADAMS Accession No.
evaluated are not significantly increased. coolant. The proposed change is consistent
There are no physical changes to the plant with the assumptions in the safety analyses
ML100620847), in addition to
being introduced by the proposed changes to and will ensure the monitored values protect limitations and conditions of NEI 94–01,
the accident source term. Implementation of the initial assumptions in the safety analyses. Revision 3–A. The proposed change
AST and the associated proposed TS changes The AST analyses have been performed would also delete the listing of one-time
and new atmospheric dispersion factors have using approved methodologies to ensure that exceptions previously granted to ILRT
asabaliauskas on DSK5VPTVN1PROD with NOTICES
no impact on the probability for initiation of analyzed events are bounding and safety frequencies.
any DBAs [Design Basis Accidents]. Once the margin has not been reduced. Also, new X/ Basis for proposed no significant
occurrence of an accident has been Qs, which are based on site specific
hazards consideration determination:
postulated, the new accident source term and meteorological data, were calculated in
atmospheric dispersion factors are an input accordance with the guidance of RG 1.194 to As required by 10 CFR 50.91(a), the
to analyses that evaluate the radiological utilize more recent data and improved licensee has provided its analysis of the
consequences. The proposed changes do not calculational methodologies. The dose issue of no significant hazards
involve a revision to the design or manner in consequences of these limiting events are consideration, which is presented
which the facility is operated that could within the acceptance criteria presented in below:
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1. Does the proposed amendment involve Engineers] Section XI, the Maintenance Rule, surveillance interval extension is bounded by
a significant increase in the probability or and TS requirements serve to provide a high the 15-year ILRT Interval and the 75-month
consequences of an accident previously degree of assurance that the containment Type C test interval currently authorized
evaluated? would not degrade in a manner that is within NEI 94–01, Revision 3–A. Industry
Response: No. detectable only by a Type A test. Based on experience supports the conclusion that Type
The proposed amendment to the TS the above, the proposed extensions do not B and C testing detects a large percentage of
involves the extension of the CPNPP, Units significantly increase the consequences of an containment leakage paths and that the
1 and 2 Type A containment test interval to accident previously evaluated. percentage of containment leakage paths that
15 years and the extension of the Type C test The proposed amendment also deletes are detected only by Type A testing is small.
interval to 75 months. The current Type A exceptions previously granted to allow one- The containment inspections performed in
test interval of 120 months (10 years) would time extensions of the ILRT test frequency for accordance with ASME Section Xl, TS and
be extended on a permanent basis to no both Units 1 and 2. These exceptions were the Maintenance Rule serve to provide a high
longer than 15 years from the last Type A for activities that have already taken place so degree of assurance that the containment
test. The current Type C test interval of 60 their deletion is solely an administrative would not degrade in a manner that is
months for selected components would be action that has no effect on any component detectable only by Type A testing. The
extended on a performance basis to no longer and no impact on how the units are operated. combination of these factors ensures that the
than 75 months. Extensions of up to nine Therefore, the proposed change does not margin of safety in the plant safety analysis
months (total maximum interval of 84 result in a significant increase in the is maintained. The design, operation, testing
months for Type C tests) are permissible only probability or consequences of an accident methods and acceptance criteria for Type A,
for non-routine emergent conditions. The previously evaluated. B, and C containment leakage tests specified
proposed extension does not involve either a 2. Does the proposed change create the in applicable codes and standards would
physical change to the plant or a change in possibility of a new or different kind of continue to be met, with the acceptance of
the manner in which the plant is operated or accident from any accident previously this proposed change, since these are not
controlled. The containment is designed to evaluated? affected by changes to the Type A and Type
provide an essentially leak tight barrier Response: No. C test intervals.
against the uncontrolled release of The proposed amendment to the TS The proposed amendment also deletes
radioactivity to the environment for involves the extension of the CPNPP, Unit 1 exceptions previously granted to allow one-
postulated accidents. The containment and and 2 Type A containment test interval to 15 time extensions of the ILRT test frequency for
the testing requirements invoked to years and the extension of the Type C test both Units 1 and 2. These exceptions were
periodically demonstrate the integrity of the interval to 75 months. The containment and for activities that would have already taken
containment exist to ensure the plant’s the testing requirements to periodically place by the time this amendment is
ability to mitigate the consequences of an demonstrate the integrity of the containment approved; therefore, their deletion is solely
accident, and do not involve the prevention exist to ensure the plant’s ability to mitigate an administrative action and does not change
or identification of any precursors of an the consequences of an accident do not how the units are operated and maintained.
accident. The change in dose risk for involve any accident precursors or initiators. Therefore, the proposed change does not
changing the Type A test frequency from The proposed change does not involve a involve a significant reduction in a margin of
three-per-ten years to once-per-fifteen-years, physical change to the plant (i.e., no new or safety.
measured as an increase to the total different type of equipment will be installed)
integrated dose risk for all internal events or a change to the manner in which the plant The NRC staff has reviewed the
accident sequences for CPNPP, of 1.00E–02 is operated or controlled. licensee’s analysis and, based on this
person rem/yr [roentgen equivalent man per The proposed amendment also deletes review, it appears that the three
year] to 6.51 person-rem/yr for Unit 1 and exceptions previously granted to allow one- standards of 10 CFR 50.92(c) are
6.53 person-rem/yr for Unit 2 using the EPRI time extensions of the ILRT test frequency for satisfied. Therefore, the NRC staff
[Energy Power Research Institute] guidance both Units 1 and 2. These exceptions were proposes to determine that the
with the base case corrosion included. for activities that would have already taken amendment request involves no
Therefore, this proposed extension does not place by the time this amendment is
significant hazards consideration.
involve a significant increase in the approved; therefore, their deletion is solely
an administrative action that does not result
Attorney for licensee: Timothy P.
probability of an accident previously
evaluated. in any change in how the units are operated. Matthews, Esq., Morgan, Lewis and
As documented in NUREG–1493 [, Therefore, the proposed change does not Bockius, 1111 Pennsylvania Avenue
‘‘Performance-Based Containment Leak-Test create the possibility of a new or different NW., Washington, DC 20004.
Program: Draft Report for Comment,’’ January kind of accident from any previously NRC Branch Chief: Michael T.
1995 (not publicly available)], Type B and C evaluated. Markley.
tests have identified a very large percentage 3. Does the proposed change involve a
of containment leakage paths, and the significant reduction in a margin of safety? South Carolina Electric and Gas
percentage of containment leakage paths that Response: No. Company Docket Nos.: 52–027 and 52–
are detected only by Type A testing is very The proposed amendment to TS 5.5.16 028, Virgil C. Summer Nuclear Station
small. The CPNPP, Units 1 and 2 Type A test involves the extension of the CPNPP, Units (VCSNS) Units 2 and 3, Fairfield
history supports this conclusion. 1 and 2 Type A containment test interval to County, South Carolina
The integrity of the containment is subject 15 years and the extension of the Type C test
to two types of failure mechanisms that can interval to 75 months for selected
Date of amendment request:
be categorized as: (1) Activity based, and; (2) components. This amendment does not alter December 4, 2014. A publicly-available
time based. Activity based failure the manner in which safety limits, limiting version is in ADAMs under Accession
mechanisms are defined as degradation due safety system set points, or limiting No. ML14339A637.
to system and/or component modifications or conditions for operation are determined. The Description of amendment request:
maintenance. Local leak rate test specific requirements and conditions of the The proposed change would amend
requirements and administrative controls TS Containment Leak Rate Testing Program Combined License (COL) Nos. NPF–93
asabaliauskas on DSK5VPTVN1PROD with NOTICES
such as configuration management and exist to ensure that the degree of containment and NPF–94 for the Virgil C. Summer
procedural requirements for system structural integrity and leak-tightness that is Nuclear Station (VCSNS) Units 2 and 3
restoration ensure that containment integrity considered in the plant safety analysis is
by changing the structure and layout of
is not degraded by plant modifications or maintained. The overall containment leak
maintenance activities. The design and rate limit specified by TS is maintained. various areas of the annex building. The
construction requirements of the The proposed change involves only the proposed amendment requires changes
containment combined with the containment extension of the interval between Type A to the Updated Final Safety Analysis
inspections performed in accordance with containment leak rate tests and Type C tests Report (UFSAR) in the form of
ASME [American Society of Mechanical for CPNPP, Units 1 and 2. The proposed departures from the incorporated plant-
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structural maintenance on the HEPA filter or 3. Does the proposed amendment involve proposed additional changes to Section
charcoal adsorber housing or following a significant reduction in a margin of safety? 3.8.3 and Section 5.5.9, ‘‘Diesel Fuel Oil
painting, fire or chemical release, and after Response: No. Testing Program,’’ to support other
every 720 hours of operation are relocated to The proposed change revises the
the licensee- controlled program. introductory paragraph of Specification 5.5.7,
related changes proposed by TSTF–501,
The existing wording states, ‘‘The VFTP ‘‘Ventilation Filter Testing Program (VFTP),’’ Revision 1. These additional changes
will establish the required testing of to be consistent with the ISTS. The proposed concern fuel oil quality and associated
Engineered Safety Feature (ESF) filter change will not reduce a margin of safety surveillance requirements (SRs).
ventilation systems at the frequencies because it has no effect on any safety analysis Basis for proposed no significant
specified in Regulatory Guide 1.52, Revision assumption. In addition, no requirements are hazards consideration determination:
2, Sections C.5.c and C.5.d, or: (1) After any being removed, but are being replaced with As required by 10 CFR 50.91(a), the
structural maintenance on the HEPA filter or references to an NRC Regulatory Guide and licensee has provided its analysis of the
charcoal adsorber housings, (2) following the requirements of SR 3.0.1.
painting, fire or chemical release in any Therefore, this proposed change does not
issue of no significant hazards
ventilation zone communicating with the involve a significant reduction in a margin of consideration, which is presented
system, or 3) after every 720 hours of safety. below:
charcoal adsorber operation.’’ Based on the above, SNC concludes that 1. Does the proposed change involve a
The proposed wording states, ‘‘A program the proposed amendment does not involve a significant increase in the probability or
shall be established to implement the significant hazards consideration under the consequences of an accident previously
following required testing of Engineered standards set forth in 10 CFR 50.92(c), and, evaluated?
Safety Feature (ESF) filter ventilation systems accordingly, a finding of ‘‘no significant Response: No.
at the frequencies specified in Regulatory hazards consideration’’ is justified. The proposed changes to TS Section 3.8.3,
Guide 1.52, Revision 2, Sections C.5.c and Conditions A and B, and to SR 3.8.3.1 and
C.5.d, and in accordance with Regulatory
The NRC staff has reviewed the
licensee’s analysis and, based on this SR 3.8.3.2 remove the volume of diesel fuel
Guide 1.52, Revision 2.’’ oil and lube oil required to support 7-day
Significant Hazards Consideration: SNC review, it appears that the three
operation of each onsite diesel generator, and
has evaluated whether or not a significant standards of 10 CFR 50.92(c) are the volume equivalent to a 6-day supply,
hazards consideration is involved with the satisfied. Therefore, the NRC staff from the TS and replace them with the
proposed amendment(s) by focusing on the proposes to determine that the associated number of days. The numerical
three standards set forth in 10 CFR 50.92, amendment request involves no volumes will be maintained under licensee
‘‘Issuance of amendment,’’ as discussed significant hazards consideration. control. The specific volume of fuel oil
below: Attorney for licensee: Jennifer M. equivalent to a 7 and 6-day supply is
1. Does the proposed amendment involve
a significant increase in the probability or Buettner, Associate General Counsel, calculated using the NRC-approved
Southern Nuclear Operating Company, methodology described in Regulatory Guide
consequences of an accident previously
40 Inverness Center Parkway, 1.137, Revision 1, ‘‘Fuel-Oil Systems for
evaluated?
Birmingham, AL 35201 Standby Diesel Generators’’ and ANSI
Response: No.
The proposed change revises the NRC Branch Chief: Robert J. [American National Standards Institute]-
introductory paragraph of Specification 5.5.7, N195 1976, ‘‘Fuel Oil Systems for Standby
Pascarelli. Diesel-Generators.’’ The specific volume of
‘‘Ventilation Filter Testing Program (VFTP),’’
to be consistent with the ISTS. Specific Tennessee Valley Authority, Docket lube oil equivalent to a 7-day and 6-day
requirements to perform testing after Nos. 50–259, 50–260, and 50–296, supply is based on the diesel generator
structural maintenance on the HEPA filter or Browns Ferry Nuclear Plant, Units 1, 2, manufacturer’s consumption values for the
charcoal adsorber housing or following run time of the diesel generator. Because the
and 3, Limestone County, Alabama requirement to maintain a 7-day supply of
painting, fire or chemical release, and after
every 720 hours of operation are retained as Date of amendment request: diesel fuel oil and lube oil is not changed and
a reference to Regulatory Guide requirements December 11, 2014 (ADAMS Accession is consistent with the assumptions in the
and general requirements in Surveillance No. ML14349A694). accident analyses, and the actions taken
Requirement (SR) 3.0.1. Implementation of Description of amendment request: when the volume of fuel oil and lube oil are
these requirements will be in the licensee- The amendment would revise Section less than a 6-day supply have not changed,
controlled VFTP. The VFTP will be neither the probability nor the consequences
3.8.3, ‘‘Diesel Fuel Oil, Lube Oil, and
maintained in accordance with 10 CFR 50.59. of any accident previously evaluated will be
Starting Air,’’ of the Technical affected.
Since any changes to the VFTP will be Specifications (TSs) by replacing the
evaluated under 10 CFR 50.59, no significant The addition of a new Condition D
increase in the probability or consequences
current volume requirements with the provides a required action and completion
of an accident previously evaluated will be number of continuous days the diesel time if new fuel oil properties are not within
allowed. generators (DGs) are required to run. limits. The new SR 3.8.3.5 requires checking
Therefore, this proposed change does not The numerical volumes will be for and removing water from the 7-day
represent a significant increase in the maintained in the licensee-controlled storage tank every 31 days. The revised
probability or consequences of an accident TSs Bases document so they may be Section 5.5.9 adds testing requirements for
previously evaluated. new fuel oil to be completed prior to the
modified under licensee control. The
2. Does the proposed amendment create addition of the new fuel oil to the 7-day
resulting requirements will specify an storage tank, as well as additional testing to
the possibility of a new or different kind of
accident from any accident previously
inventory of stored diesel fuel oil and be completed prior or within 31 days of the
evaluated? lube oil sufficient for a 7-day supply for addition. These requirements are more
Response: No. each DG. This proposed amendment is restrictive testing requirements and provide
The proposed change does not involve a consistent with NRC’s approved corrective action to be taken if the testing
physical alteration to the plant (i.e., no new Technical Specifications Task Force limits are not met. They are taken from the
asabaliauskas on DSK5VPTVN1PROD with NOTICES
or different type of equipment will be (TSTF) Improved Standard Technical current NRC approved NUREG–1433,
installed) or a change to the methods Specifications Change Traveler TSTF– Revision 4, ‘‘Standard Technical
governing normal plant operation. The Specifications, General Electric BWR/4
501, Revision 1, ‘‘Relocate Stored Fuel
changes do not alter the assumptions made Plants.’’ Improved, more restrictive testing
in the safety analysis.
Oil and Lube Oil Volume Values to standards will neither change the probability
Therefore, the proposed change does not Licensee Control.’’ The availability of or the consequences of any accident
create the possibility of a new or different this TSs improvement was announced previously evaluated be affected.
kind of accident from any accident in the Federal Register on May 26, 2010 Therefore, the proposed changes do not
previously evaluated. (75 FR 29588). The licensee also involve a significant increase in the
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![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17105
probability or consequences of an accident Description of amendment request: 3. Does the proposed change involve a
previously evaluated. The amendment would revise the significant reduction in a margin of safety?
2. Does the proposed change create the Technical Specification (TS) Response: No.
possibility of a new or different kind of The proposed change revises or adds SRs
accident from any accident previously
requirements to address NRC Generic
Letter 2008–01, ‘‘Managing Gas that require verification that the ECCS, the
evaluated?
Response: No. Accumulation in Emergency Core RHR System, and the Containment Spray
The changes do not involve a physical Cooling, Decay Heat Removal, and System are not rendered inoperable due to
alteration of the plant (i.e., no new or Containment Spray Systems,’’ as accumulated gas and to provide allowances
different type of equipment will be installed) described in Technical Specification which permit performance of the revised
or a change in the methods governing normal Task Force (TSTF) Traveler TSTF–523, verification. The proposed change adds new
plant operation. The change does not alter Revision 2, ‘‘Generic Letter 2008–01, requirements to manage gas accumulation in
assumptions made in the safety analysis but order to ensure the subject systems are
ensures that the diesel generator operates as
Managing Gas Accumulation.’’
capable of performing their assumed safety
assumed in the accident analysis. The Basis for proposed no significant functions. The proposed SRs are more
proposed change is consistent with the safety hazards consideration determination: comprehensive than the current SRs and will
analysis assumptions. As required by 10 CFR 50.91(a), the
Therefore, the proposed change does not ensure that the assumptions of the safety
licensee has provided its analysis of the analysis are protected. The proposed change
create the possibility of a new or different issue of no significant hazards
kind of accident from any accident does not adversely affect any current plant
previously evaluated.
consideration, which is presented safety margins or the reliability of the
3. Does the proposed change involve a below: equipment assumed in the safety analysis.
significant reduction in a margin of safety? 1. Does the proposed change involve a Therefore, there are no changes being made
Response: No. significant increase in the probability or to any safety analysis assumptions, safety
The proposed changes to Section 3.8.3, consequences of an accident previously limits or limiting safety system settings that
Conditions A and B, and to SR 3.8.3.1 and evaluated? would adversely affect plant safety as a result
SR 3.8.3.2 remove the numerical volume of Response: No. of the proposed change.
diesel fuel oil and lube oil required to The proposed change revises or adds SRs Therefore, the proposed change does not
support 7-day operation of each onsite diesel [surveillance requirements] that require
generator, and the numerical volume involve a significant reduction in a margin of
verification that the Emergency Core Cooling
equivalent to a 6-day supply from the TS and safety.
System (ECCS), the Residual Heat Removal
replaces them with the associated number of (RHR) System, and the Containment Spray
days. The numerical volumes will be The NRC staff has reviewed the
System are not rendered inoperable due to
maintained under licensee control. As the licensee’s analysis and, based on this
accumulated gas and to provide allowances
bases for the existing limits on diesel fuel oil which permit performance of the revised review, it appears that the three
volume and lube oil volume are not changed, verification. Gas accumulation in the subject standards of 10 CFR 50.92(c) are
no change is made to the accident analysis systems is not an initiator of any accident satisfied. Therefore, the NRC staff
assumptions and no margin of safety is previously evaluated. As a result, the proposes to determine that the
reduced as part of this change. probability of any accident previously amendment request involves no
The new, more restrictive, testing evaluated is not significantly increased. The significant hazards consideration.
requirements, and the provision for proposed SRs ensure that the subject systems
corrective action to be taken if the testing continue to be capable to perform their Attorney for licensee: Jay Silberg, Esq.,
limits are not met, are taken from the current assumed safety function and are not rendered Pillsbury Winthrop Shaw Pittman LLP,
NRC approved NUREG–1433, Revision 4, inoperable due to gas accumulation. Thus, 2300 N Street NW., Washington, DC
‘‘Standard Technical Specifications, General the consequences of any accident previously 20037.
Electric BWR/4 Plants.’’ These changes do evaluated are not significantly increased.
not revise the accident analysis assumptions NRC Branch Chief: Michael T.
Therefore, the proposed change does not
and no margin of safety is reduced as part of involve a significant increase in the
Markley.
these changes. probability or consequences of an accident
Therefore, the proposed change does not III. Previously Published Notices of
previously evaluated. Consideration of Issuance of
involve a significant reduction in a margin of 2. Does the proposed change create the
safety. possibility of a new or different kind of
Amendments to Facility Operating
The NRC staff has reviewed the accident from any accident previously Licenses and Combined Licenses,
licensee’s analysis and, based on this evaluated? Proposed No Significant Hazards
review, it appears that the three Response: No. Consideration Determination, and
standards of 10 CFR 50.92(c) are The proposed change revises or adds SRs Opportunity for a Hearing
that require verification that the ECCS, the
satisfied. Therefore, the NRC staff The following notices were previously
RHR System, and the Containment Spray
proposes to determine that the System are not rendered inoperable due to published as separate individual
amendment request involves no accumulated gas and to provide allowances notices. The notice content was the
significant hazards consideration. which permit performance of the revised same as above. They were published as
Attorney for licensee: General verification. The proposed change does not individual notices either because time
Counsel, Tennessee Valley Authority, involve a physical alteration of the plant (i.e., did not allow the Commission to wait
400 West Summit Hill Drive, 6A West no new or different type of equipment will
for this biweekly notice or because the
Tower, Knoxville, TN 37902. be installed) or a change in the methods
governing normal plant operation. In action involved exigent circumstances.
NRC Branch Chief: Shana R. Helton.
addition, the proposed change does not They are repeated here because the
asabaliauskas on DSK5VPTVN1PROD with NOTICES
Wolf Creek Nuclear Operating impose any new or different requirements biweekly notice lists all amendments
Corporation, Docket No. 50–482, Wolf that could initiate an accident. The proposed issued or proposed to be issued
Creek Generating Station, Coffey change does not alter assumptions made in involving no significant hazards
County, Kansas the safety analysis and is consistent with the consideration.
safety analysis assumptions.
Date of amendment request: Therefore, the proposed change does not For details, see the individual notice
November 20, 2014. A publicly- create the possibility of a new or different in the Federal Register on the day and
available version is in ADAMS under kind of accident from any accident page cited. This notice does not extend
Accession No. ML14330A247. previously evaluated. the notice period of the original notice.
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![17108 Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices
Effective date: As of the date of Informed Method for Control of Safety Evaluation dated February 26,
issuance, to be implemented within 60 Surveillance Frequencies’’ (ADAMS 2015.
days. Accession No. ML071360456). No significant hazards consideration
Amendments Nos.: 296 and 299. A Date of issuance: March 6, 2015. comments received: No.
publicly-available version is in ADAMS Effective date: As of the date of
under Accession No. ML14363A227; issuance and shall be implemented Southern Nuclear Operating Company,
documents related to these amendments within 120 days of issuance. Inc., Docket Nos. 50–348 and 50–364,
are listed in the Safety Evaluation Amendment Nos.: 292 and 179. A Joseph M. Farley Nuclear Plant, Units 1
enclosed with the amendments. publicly-available version is in ADAMS and 2, Houston County, Alabama
Renewed Facility Operating License under Accession No. ML14322A461;
Nos. DPR–44 and DPR–56: The documents related to these amendments Date of application for amendment:
amendments revised the FOLs and the are listed in the Safety Evaluation September 25, 2012; as supplemented
TSs. enclosed with the amendments. on December 20, 2012; September 16,
Date of initial notice in Federal Renewed Facility Operating License October 30, and November 12, 2013;
Register: September 2, 2014 (79 FR Nos. DPR–66 and NPF–73: Amendments April 23, May 23, July 3, August 11,
52062). The supplemental letter dated revised the Renewed Facility Operating August 29, and October 13, 2014; and
December 1, 2014, provided additional Licenses and Technical Specifications. January 16, 2015.
information that clarified the Date of initial notice in Federal Brief description of amendments: The
application, did not expand the scope of Register: January 21, 2014 (79 FR amendment authorizes the transition of
the application as originally noticed, 3416). The supplemental letters dated the Joseph M. Farley Nuclear Plant,
and did not change the staff’s original June 26, 2014, September 21, 2014, and Units 1 and 2, fire protection program
proposed no significant hazards February 4, 2015, provided additional to a risk-informed, performance-based
consideration determination as information that clarified the program based on National Fire
published in the Federal Register. application, did not expand the scope of Protection Association (NFPA) 805,
The Commission’s related evaluation the application as originally noticed, ‘‘Performance-Based Standard for Fire
of the amendments is contained in a and did not change the staff’s original Protection for Light Water Reactor
Safety Evaluation dated March 11, 2015. proposed no significant hazards Electric Generating Plants, 2001
No significant hazards consideration consideration determination as Edition’’ (NFPA 805), in accordance
comments received: No. published in the Federal Register.
with 10 CFR 50.48(c).
The Commission’s related evaluation
FirstEnergy Nuclear Operating of the amendment is contained in a Date of issuance: March 10, 2015.
Company, et al., Docket Nos. 50–334 Safety Evaluation dated March 6, 2015. Effective date: As of its date of
and 50–412, Beaver Valley Power No significant hazards consideration issuance and shall be implemented
Station, Units 1 and 2 (BVPS–1 and 2), comments received: No. within 60 days from the date of
Beaver County, Pennsylvania issuance.
Southern Nuclear Operating Company,
Date of amendment request: October Docket Nos. 52–025 and 52–026, Vogtle Amendment Nos.: Unit 1–196, Unit
18, 2013, as supplemented by letters Electric Generating Plant (VEGP) Units 2–192. A publicly-available version is in
dated June 26, 2014, September 21, 3 and 4, Burke County, Georgia ADAMS under Accession No.
2014, and February 4, 2015. ML14308A048, documents related to
Brief description of amendments: The Date of amendment request:
November 21, 2013, and supplemented these amendments are listed in the
amendment changes the Beaver Valley
by the letters dated March 5 and June Safety Evaluation enclosed with the
Power Station Technical Specifications
30, 2014. amendments.
(TS). Specifically, this change request
involves the adoption of an approved Brief description of amendment: The Facility Operating License Nos. NPF–
amendment authorizes changes to the 2 and NPF–8: The amendments revised
change to the standard TS for
VEGP Units 3 and 4 Updated Final the Renewed Facility Operating
Westinghouse plants (NUREG–1431), to
Safety Analysis Report to revise the Licenses and Technical Specifications.
allow relocation of specific TS
details of the effective thermal
surveillance frequencies to a licensee- Date of initial notice in Federal
conductivity resulting from the
controlled program. The proposed Register: March 12, 2013 (78 FR
oxidation of the inorganic zinc
change is described in TS Task Force 15750). The supplemental letters dated
component of the containment vessel
(TSTF) Traveler, TSTF–425, Revision 3, September 16, October 30, and
coating system.
‘‘Relocation Surveillance Frequencies to Date of issuance: February 26, 2015. November 12, 2013; April 23, May 23,
Licensee Control—RITSTF [Risk- Effective date: As of the date of July 3, August 11, August 29, and
Informed Technical Specifications Task issuance and shall be implemented October 13, 2014; and January 16, 2015,
Force] Initiative 5b’’ (Agencywide within 30 days of issuance. provided additional information that
Documents Access and Management Amendment No.: 31. A publicly- clarified the application, did not expand
System (ADAMS) Accession No. available version is in ADAMS under the scope of the application as originally
ML090850642). A Notice of Availability Accession No. ML15028A358; noticed, and did not change the staff’s
was published in the Federal Register documents related to these amendments original proposed no significant hazards
on July 6, 2009 (74 FR 31996). are listed in the Safety Evaluation consideration determination as
The proposed change relocates
asabaliauskas on DSK5VPTVN1PROD with NOTICES
enclosed with the amendments. published in the Federal Register.
surveillance frequencies to a licensee- Facility Combined Licenses Nos. NPF– The Commission’s related evaluation
controlled program, the Surveillance 91 and NPF–92: Amendment revised the of the amendments is contained in a
Frequency Control Program. This Facility Combined Licenses. Safety Evaluation dated March 10, 2015.
change is applicable to licensees using Date of initial notice in Federal
probabilistic risk guidelines contained Register: March 18, 2014 (79 FR No significant hazards consideration
in NRC-approved NEI 04–10, Revision 15150). comments received: No.
1, ‘‘Risk-Informed Technical The Commission’s related evaluation Dated at Rockville, Maryland, this 23rd day
Specifications Initiative 5b, Risk- of the amendment is contained in a of March 2015.
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![Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17109
For the Nuclear Regulatory Commission. Ellmers at 301–415–0442 or via email at regulations and operational experience;
Michele G. Evans, Glenn.Ellmers@nrc.gov. announces guidelines for a formal
Director, Division of Operating Reactor * * * * * process for interested DoD civilian
Licensing, Office of Nuclear Reactor The NRC Commission Meeting acquisition organizations to use to
Regulation. Schedule can be found on the Internet request approval to participate in
[FR Doc. 2015–07192 Filed 3–30–15; 8:45 am] at: http://www.nrc.gov/public-involve/ AcqDemo; and provides notice of
BILLING CODE 7590–01–P public-meetings/schedule.html. expansion of coverage to new or
* * * * * realigned organizations.
The NRC provides reasonable DATES: The amendments will become
NUCLEAR REGULATORY accommodation to individuals with effective as of March 31, 2015.
COMMISSION disabilities where appropriate. If you FOR FURTHER INFORMATION CONTACT: (1)
[NRC–2015–0001] need a reasonable accommodation to DoD: Darryl R. Burgan, Civilian
participate in these public meetings, or Acquisition Workforce Personnel
Sunshine Act Meeting Notice need this meeting notice or the Demonstration Project Program Office,
transcript or other information from the 9820 Belvoir Road, Ft. Belvoir, VA
DATE:March 30, April 6, 13, 20, 27, public meetings in another format (e.g. 22060, (703) 805–5050; (2) OPM: Zelma
May 4, 2015. braille, large print), please notify Moore, U.S. Office of Personnel
PLACE: Commissioners’ Conference Kimberly Meyer, NRC Disability Management, 1900 E Street NW., Room
Room, 11555 Rockville Pike, Rockville, Program Manager, at 301–287–0727, by 7456, Washington, DC 20415, (202) 606–
Maryland. videophone at 240–428–3217, or by 1157.
STATUS: Public and Closed. email at Kimberly.Meyer-Chambers@ SUPPLEMENTARY INFORMATION:
nrc.gov. Determinations on requests for
Week of March 30, 2015 reasonable accommodation will be A. Background
There are no meetings scheduled for made on a case-by-case basis. The AcqDemo Project was established
the week of March 30, 2015. * * * * * under the authority of the Secretary of
Members of the public may request to Defense, with the approval of OPM.
Week of April 6, 2015—Tentative
receive this information electronically. Subject to the authority, direction, and
There are no meetings scheduled for If you would like to be added to the control of the Secretary, the Under
the week of April 6, 2015. distribution, please contact the Nuclear Secretary of Defense for Acquisition,
Week of April 13, 2015—Tentative Regulatory Commission, Office of the Technology, and Logistics (USD(AT&L))
Secretary, Washington, DC 20555 (301– carries out the powers, functions, and
Tuesday, April 14, 2015 415–1969), or email duties of the Secretary concerning the
9:30 a.m. Meeting with the Advisory Brenda.Akstulewicz@nrc.gov or DoD acquisition workforce. As stated in
Committee on the Medical Uses of Patricia.Jimenez@nrc.gov. the most recent legislative
Isotopes Dated: March 26, 2015. authorization, the purpose of the
(Public Meeting) demonstration project is ‘‘to determine
Glenn Ellmers,
(Contact: Nima Ashkeboussi, 301- the feasibility or desirability of one or
Policy Coordinator, Office of the Secretary.
415–5775) more proposals for improving the
[FR Doc. 2015–07384 Filed 3–27–15; 11:15 am] personnel management policies or
This meeting will be webcast live at
BILLING CODE 7590–01–P procedures that apply with respect to
the Web address—http://www.nrc.gov/.
the acquisition workforce of the [DoD]
Thursday, April 16, 2015 and supporting personnel assigned to
9:30 a.m. Meeting with the Organization OFFICE OF PERSONNEL work directly with the acquisition
of Agreement States and the MANAGEMENT workforce.’’
Conference of Radiation Control This demonstration project was
Civilian Acquisition Workforce originally authorized under section
Program Directors
(Public Meeting) Personnel Demonstration Project; 4308 of the National Defense
(Contact: Nima Ashkeboussi, 301– Department of Defense Authorization Act (NDAA) for Fiscal
415–5775) AGENCY: U.S. Office of Personnel Year (FY) 1996 (Pub. L. 104–106, 110
This meeting will be webcast live at Management (OPM). Stat. 669; 10 United States Code
the Web address—http://www.nrc.gov/. ACTION: Notice of amendments to the Annotated (U.S.C.A.) 1701 note), as
project plan for the Department of amended by section 845 of NDAA for
Week of April 20, 2015—Tentative FY 1998 (Pub. L. 105–85, 111 Stat.1845);
Defense (DoD) Civilian Acquisition
There are no meetings scheduled for Workforce Personnel Demonstration section 813 of NDAA for FY 2003 (Pub.
the week of April 20, 2015. Project (AcqDemo). L. 107–314, 116 Stat. 2609); and section
1112 of NDAA for FY 2004 (Pub. L.
Week of April 27, 2015—Tentative
SUMMARY: The DoD, with the approval of 108–136, 117 Stat. 1634). Section 1113
There are no meetings scheduled for OPM, received authority to conduct a of NDAA for FY 2010 (Pub. L. 111–84,
the week of April 27, 2015. personnel demonstration project within 123 Stat. 2190) repealed the National
DoD’s civilian acquisition workforce Security Personnel System and directed
asabaliauskas on DSK5VPTVN1PROD with NOTICES
Week of May 4, 2015—Tentative
and among those supporting personnel conversion of all NSPS employees to
There are no meetings scheduled for assigned to work directly with it. This their previous pay system by January 1,
the week of May 4, 2015. notice announces the repeal and 2012. All NSPS employees formerly in
* * * * * replacement of AcqDemo’s original legal AcqDemo were transitioned back to
The schedule for Commission authorization and modifies the project AcqDemo during the month of May
meetings is subject to change on short plan to include new provisions; updates 2011. On January 7, 2011, the original
notice. For more information or to verify the project plan to address changes demonstration project authority was
the status of meetings, contact Glenn resulting from new General Schedule repealed and codified at section 1762 of
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Document Created: 2015-12-18 11:48:28
Document Modified: 2015-12-18 11:48:28
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by April 30, 2015. A request for a hearing must be filed by June 1, 2015. | |
| Contact | Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: [email protected] | |
| FR Citation | 80 FR 17083 |
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