| Page Range | 61476-61492 | |
| FR Document | 2015-25860 |
[Federal Register Volume 80, Number 197 (Tuesday, October 13, 2015)] [Notices] [Pages 61476-61492] From the Federal Register Online [www.thefederalregister.org] [FR Doc No: 2015-25860] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION [NRC-2015-0236] Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission. ACTION: Biweekly notice. ----------------------------------------------------------------------- SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from September 15 to September 28, 2015. The last biweekly notice was published on September 29, 2015. DATES: Comments must be filed by November 12, 2015. A request for a hearing must be filed by December 14, 2015. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different [[Page 61477]] method for submitting comments on a specific subject):Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0236. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415- 3463; email: [email protected]. 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: Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1384, email: [email protected]. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2015-0236 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-2015-0236. 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-2015-0236, facility name, unit number(s), application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS. II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 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 [[Page 61478]] 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) 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. Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3), New London County, Connecticut Date of amendment request: June 30, 2015. A publicly-available version is in ADAMS under Accession No. ML15183A022. Description of amendment request: The amendments would revise the MPS2 and MPS3 Final Safety Analysis Reports (FSARs) to: (1) Delete the information pertaining to the severe line outage detection (SLOD) special protection system; (2) update the description of the tower structures associated with the four offsite transmission lines feeding Millstone Power Station; and (3) describe how the current offsite power source configuration and design satisfies the requirements of General Design Criteria (GDC)-17, ``Electric Power Systems,'' and GDC-5, ``Sharing of Structures, Systems, and Components.'' The amendments also request NRC approval of a new Technical Requirements Manual (TRM) requirement, ``Offsite Line Power Sources,'' for MPS2 and MPS3. With one offsite transmission line nonfunctional, the TRM requirement would allow 72 hours to restore the nonfunctional line with a provision to allow up to 14 days if specific TRM action requirements are met. 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 any accident previously evaluated? Response: No. The post-modification configuration of the offsite 345 [kilovolt (kV)] transmission system (four lines separately supported and SLOD disabled) improves overall grid reliability and continues to meet the requirements for two independent sources of offsite power (GDC- 17). Therefore, the post-modification configuration does not significantly increase the probability or consequences of a loss of offsite power event. Likewise, the associated proposed changes to the MPS2 and MPS3 FSARs to document the revised 345 kV transmission line tower design and disabling of SLOD, do not increase the probability or consequences of an accident previously evaluated in the FSARs. The grid (offsite power) is by design, the preferred power source for the affected units. The grid provides a reliable source of power to MPS2 and MPS3 while the units are at power, in the event of unit trips, and when the units are shut down for maintenance. New TRM requirements are proposed that will maintain adequate defense in depth to ensure grid reliability and stability are preserved. A loss of offsite power event is an anticipated operational occurrence. The proposed changes do not significantly increase the probability of this event. Additionally, as described in Chapter 14 (MPS2) and Chapter 15 (MPS3), several events are assumed to occur coincident with a loss of offsite power. Sufficient onsite power sources are available to mitigate these events and ensure the consequences of the existing analyses for these events remain bounding. The proposed new TRM requirements for offsite line power sources will not change the plant design or design requirements. The design criteria for the offsite power system remain unchanged. Therefore, the safety analyses as documented in the MPS2 and MPS3 FSARs remain unchanged. Temporary reductions in the number of offsite lines from four to three, in accordance with the proposed TRM action requirements, will not adversely affect offsite power system availability in the event of a loss of either MPS2, MPS3, the largest other unit on the grid, or the most critical transmission line. Use of the proposed TRM requirements will not cause an accident to occur and will not change how accident mitigation equipment is operated. Allowing one offsite line to be nonfunctional for up to 14 days does not increase the probability of any previously evaluated accidents. Therefore, the proposed changes to the offsite 345 kV transmission system (four lines separately supported and SLOD disabled) and proposed new TRM requirements does not significantly increase the probability or consequences of an accident previously evaluated. [[Page 61479]] 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed amendments do not change the design function or operation of the offsite power system and do not affect the offsite power systems ability to perform its design function. The proposed amendments do not conflict with the design criteria, codes, or standards committed to in the licensing basis. The existing codes and standards, as they apply to the onsite emergency power systems, remain unchanged. The design criteria for the offsite power system remain unchanged. Therefore, the safety analyses as documented in the MPS2 and MPS3 FSARs remain unchanged. No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing basis are created by the proposed amendment. The offsite power system is assumed to be available during several FSAR Chapter 14 (MPS2) and Chapter 15 (MPS3) events. The new TRM requirements would allow 72 hours to restore a nonfunctional line, and up to 14 days to restore a nonfunctional line if specific TRM action requirements are met. Use of these TRM requirements does not impact offsite power availability and does not create the possibility for a new or different kind of accident from any previously evaluated. Temporary reductions in the number of offsite lines from four to three, in accordance with the proposed TRM requirements, will continue to ensure offsite power system availability in the event of a loss of either MPS2, MPS3, the largest other unit on the grid, or the most critical transmission line. The proposed amendments have no adverse effect on plant operation or accident mitigation equipment. The response of the plants and the operators following a design basis accident will not be different. In addition, the proposed amendments do not create the possibility of a new failure mode associated with any equipment or personnel failures. Therefore, the proposed amendments will 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. The post-modification configuration of the offsite 345 kV transmission system (four lines separately supported and SLOD disabled) improves overall grid reliability and continues to meet the requirements for two independent sources of offsite power (GDC- 17). Likewise, the addition of TRM requirements that limit the unavailability of offsite lines provides acceptable assurance that line outages will not result in a significant reduction to grid stability and hence also to the margin of safety. The offsite power systems are assumed to be available during several FSAR Chapter 14 (MPS2) and Chapter 15 (MPS3) events. The loss of the offsite power system is an anticipated operational occurrence. Additionally, as described in Chapter 14 (MPS2) and Chapter 15 (MPS3), several events are assumed to occur coincident with a loss of offsite power. Sufficient onsite power sources are available to mitigate these events and ensure the consequences of the existing analyses for these events remain bounding. The proposed amendments do not affect the assumptions in the safety analyses or the ability to safely shutdown the reactors and mitigate accident conditions. Station structures, systems, and components will continue to be able to mitigate the design basis accidents as assumed in the safety analyses and ensure proper operation of accident mitigation equipment. In addition, the proposed amendment will not affect equipment design or operation of station structures, systems, and components and there are no changes being made to the safety limits or safety system settings required by technical specifications. Therefore, the proposed amendments will not result in a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Benjamin Beasley. Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: July 9, 2015. A publicly-available version is in ADAMS under Accession No. ML15198A151. Description of amendment request: The amendments would change the reactor coolant pump (RCP) under-frequency trip setpoint Allowable Value (AV) and add footnotes. The proposed license amendment request affects Technical Specification (TS) 3.3.1, ``Reactor Trip System Instrumentation,'' for McGuire Nuclear Station, Units 1 and 2. Basis for proposed no significant hazards 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 TS changes involve lowering the existing RCP under- voltage ALLOWABLE VALUE and adopting [Technical Specification Task Force (TSTF)-493] provisions for as-found and as-left calibration tolerances. The proposed TS changes serve to further ensure the Reactor Trip RCP under-frequency and under-voltage trip instrumentation will properly function as credited in the safety analyses. The proposed changes do not alter any assumptions previously made in the radiological consequences evaluations nor do they affect mitigation of the radiological consequences of an accident previously evaluated. The proposed TS changes do not affect the probability of accident initiation. In summary, the proposed changes will not involve any 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 TS changes involve lowering the existing RCP under- voltage ALLOWABLE VALUE and adopting TSTF-493 provisions for as- found and as-left calibration tolerances. No new accident scenarios, failure mechanisms, or single failures are introduced as a result of any of the proposed changes. The Reactor Trip System is not an accident initiator. No changes to the overall manner in which the plant is operated are being proposed. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The proposed TS changes serve to ensure proper operation of the Reactor Trip RCP under-frequency and under-voltage trip instrumentation and that the instrumentation will properly function as credited in the safety analyses. The proposed TS changes will not have any effect on the margin of safety of fission product barriers. No accident mitigating equipment will be adversely impacted as a result of the modification. Therefore, existing safety margins will be preserved. None of the proposed changes will 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, Associate General Counsel, Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 28202. NRC Branch Chief: Robert J. Pascarelli. [[Page 61480]] Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina; and 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: July 15, 2015. A publicly-available version is available at ADAMS Accession No. ML15196A093. Description of amendment request: The proposed amendments would revise the facilities Updated Final Safety Analysis Reports (UFSARs) to provide gap release fractions for high-burnup fuel rods that exceed the linear heat generation rate limit detailed in Table 3 of Regulatory Guide (RG) 1.183, ``Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession No. ML003716792). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 [gigawatt days per metric ton unit (GWD/MTU)] that exceed the 6.3 [kiloWatt per foot (kW/ft)] linear heat generation rate (LHGR) limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined and accounted for in the dose analysis for Catawba Nuclear Station (CNS), Units 1 and 2; McGuire Nuclear Station (MNS), Units 1 and 2; and Oconee Nuclear Station (ONS), Units 1, 2, and 3. The dose consequence reported in each site's Updated Final Safety Analysis Report (UFSAR) were reanalyzed for fuel handling-type accidents only. Dose consequences were not reanalyzed for other non- fuel-handling accidents since no fuel rod that is predicted to enter departure from nuclear boiling (DNB) will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11. The current NRC requirements, as described in 10 CFR 50.67, specifies dose acceptance criteria in terms of Total Effective Dose Equivalent (TEDE). The revised dose consequence analysis for fuel handling-type events at CNS, MNS, and ONS meet the applicable TEDE dose acceptance criteria (specified also in RG 1.183). A slight increase in dose consequences is exhibited. However, the increase is not significant and the new TEDE results are below regulatory acceptance criteria. The changes proposed do not affect the precursors for fuel handling-type accidents analyzed in Chapter 15 of the CNS, MNS, or ONS UFSARs. The probability remains unchanged since the accident analyses performed and discussed in the basis for the UFSAR changes, involve no change to a system, structure, or component that affects initiating events for any UFSAR Chapter 15 accident evaluated. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously analyzed. 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 involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined and accounted for in the dose analysis for CNS, MNS, and ONS. The dose consequences reported in each site's UFSAR were reanalyzed for fuel handling-type accidents only. Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter departure from nucleate boiling (DNB) will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11. The proposed change does not involve the addition or modification of any plant equipment. The proposed change has the potential to affect future core designs for CNS, MNS, and ONS. However, the impact will not be beyond the standard function capabilities of the equipment. The proposed change involves using gap release fractions that would allow high-burnup fuel rods (i.e., greater than 54 GWD/MTU) to exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Accounting for these new gap release fractions in the dose analysis for CNS, MNS, and ONS does not create the possibility of a new accident. Therefore, the proposed change does no 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. The proposed change involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined and accounted for in the dose analysis for CNS, MNS, and ONS. The dose consequences reported in each site's UFSAR were reanalyzed for fuel handling-type accidents only. Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter departure from nucleate boiling (DNB) will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11. The proposed change has the potential for an increased postulated accident dose at CNS, MNS or ONS. However, the analysis demonstrates that the resultant doses are within the appropriate acceptance criteria. The margin of safety, as described by 10 CFR 50.67 and Regulatory Guide 1.183, has been maintained. Furthermore, the assumptions and input used in the gap release and dose consequences calculations are conservative. These conservative assumptions ensure that the radiation doses calculated pursuant to Regulatory Guide 1.183 and cited in this license amendment requires are the upper bounds to radiological consequences of the fuel handling-type accidents analyzed. The analysis shows that with increased gap release fractions accounted for in the dose consequences calculations there is margin between the offsite radiation doses calculated and the dose limits of 10 CFR 50.67 and acceptance criteria of Regulatory Guide 1.183. The proposed change will not degrade the plant protective boundaries, will not cause a release of fission products to the public and will not degrade the performance of any structures, systems and components important to safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 28202. NRC Branch Chief: Robert J. Pascarelli. Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant (JAF), Oswego County, New York Date of amendment request: August 20, 2015. A publicly-available version is in ADAMS under Accession No. ML15232A761. Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.6, ``Primary Containment Leak Rate Testing Program,'' to allow permanent extension of the Type A Primary Containment Integrated Leak Rate Test (ILRT) interval to 15 years and to allow extension of Type C Local Leak Rate Test (LLRT) testing interval up to 75 months. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or [[Page 61481]] consequences of an accident previously evaluated? Response: No. The proposed amendment to the TS involves the extension of the JAF 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. As such, 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 plant risk for those accident sequences influenced by Type A testing, is 0.0087 person-[roentgen equivalent man (rem)]/year. [Electric Power Research Institute (EPRI)] Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of <= 1.0 person-rem per year, or <= 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of the risk assessment for this amendment meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. 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''], 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 JAF 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 [American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code] 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 JAF. 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 has no effect on any component and no impact on how the unit is 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 JAF 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 JAF. 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 unit is 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 5.5.6 involves the extension of the JAF 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 JAF. The proposed surveillance interval extension is bounded by the 15- year ILRT Interval and the 75-month Type C test interval currently authorized within [Nuclear Energy Institute (NEI) 94-01, Revision 3- A [``Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,'' July 2012 (ADAMS Accession No. ML12221A202)]. 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 JAF. 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 unit is operated and maintained. Thus, there is no reduction in any margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Benjamin G. Beasley. 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: June 29, 2015. A publicly-available version is in ADAMS under Accession No. ML15180A376. Description of amendment request: The amendment proposes a change to [[Page 61482]] the GGNS Cyber Security Plan (CSP) Milestone 8 full implementation date as set forth in the CSP Implementation Schedule. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change to the CSP Implementation Schedule is administrative in nature. This change does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems and components relied upon to mitigate the consequences of postulated accidents and has no impact on the probability or consequences of an accident previously evaluated. Therefore, the proposed 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 to the CSP Implementation Schedule is administrative in nature. This proposed change does not alter accident analysis assumptions, add any initiators or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. The proposed change does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the CSP Implementation Schedule is administrative in nature. In addition, the milestone date delay for full implementation of the CSP has no substantive impact because other measures have been taken which provide adequate protection during this period of time. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Joseph A. Aluise, Associate General Counsel--Legal, Nuclear and Environmental, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, LA 70113. NRC Branch Chief: Meena K. Khanna. Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant (Ginna), Wayne County, New York Date of amendment request: June 4, 2015. A publicly-available version is in ADAMS under Accession No. ML15166A075. Description of amendment request: The amendment would modify Ginna's technical specifications (TS) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, [Rev. 1, ``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,'' April 2007 (ADAMS Accession No. ML071360456)]. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated? Response: No. The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program [SFCP]. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. No new or different accidents result from utilizing the proposed changes. 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. In addition, the changes do not impose any new or different requirements. 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 changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The design, operation, testing methods, and acceptance criteria for systems, structures, and components, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04- 10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177 [``An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications'']. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: J. Bradley Fewell, Senior Vice President, Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Branch Chief: Benjamin G. Beasley. [[Page 61483]] Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2 (SL-1 and 2), St. Lucie County, Florida Date of amendment request: March 10, 2015. A publicly-available version is in ADAMS under Accession No. ML15084A141. Description of amendment request: The amendments would remove Technical Specification (TS) Limiting Condition for Operation (LCO) 3/ 4.9.5, ``Communications,'' from the SL-1 and 2 TSs; remove LCO 3/4.9.6, ``Manipulator Crane Operability,'' from the SL-1 TSs; and remove LCO 3/ 4.9.6, ``Manipulator Crane,'' from the SL-2 TSs. Each of these TS requirements will be relocated to the Updated Final Safety Analysis Report (UFSAR) for SL-1 and 2. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes act to remove the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59. The probability of occurrence of a previously evaluated accident is not increased because these changes do not introduce any new potential accident initiating conditions. The consequences of accidents previously evaluated in the UFSAR are not affected because the ability of the components to perform their required functions is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes act to remove the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59. The proposed changes do not introduce new modes of plant operation and do not involve physical modifications to the plant (no new or different type of equipment will be installed). There are no changes in the method by which any safety related plant structure, system, or component (SSC) performs its specified safety function. As such, the plant conditions for which the design basis accident analyses were performed remain valid. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of the proposed changes. There will be no adverse effect or challenges imposed on any SSC as a result of the proposed changes. 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. Margin of safety is related to confidence in the ability of the fission product barriers to perform their accident mitigation functions. The proposed changes act to remove the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59. The proposed changes do not physically alter any SSC. There will be no effect on those SSCs necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, loss of cooling accident peak cladding temperature (LOCA PCT), or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: William S. Blair, Managing Attorney-- Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, FL 33408-0420. NRC Branch Chief: Shana R. Helton. Northern States Power Company--Minnesota Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: September 2, 2015. A publicly-available version is in ADAMS under Accession No. ML15246A530. Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling System]--Operating,'' to correct the current non-conservative value specified for minimum Alternate Nitrogen System pressure. The proposed change would revise the TS surveillance requirement (SR) 3.5.1.3.b pressure limit for determining operability of the Alternate Nitrogen System from greater than or equal to (>=) 410 pounds per square inch gauge (psig) to a corrected value of >=1060 psig. 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 provided below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The proposed TS change does not introduce new equipment or new equipment operating modes, nor does the proposed change alter existing system relationships. The proposed change does not affect plant operation[.] Further, the proposed change does not increase the likelihood of the malfunction of any SSC [structure, system or component] or impact any analyzed accident. Consequently, the probability of an accident previously evaluated is not affected and there is no significant increase in the consequences of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The [[Page 61484]] 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 in the methods governing normal plant operations. The proposed change does not alter assumptions made in the safety analysis for the components supplied by the Alternate Nitrogen System. Further, the proposed change does not introduce new accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis assumptions and acceptance criteria are not affected by this change. Therefore, the proposed change does not involve a significant reduction in 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401. NRC Branch Chief: David L. Pelton. Northern States Power Company--Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of amendment request: July 15, 2015. A publicly-available version is in ADAMS under Accession No. ML15196A576. Description of amendment request: The proposed amendment would revise or add technical specification (TS) surveillance requirements (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Residual Heat Removal (RHR) System/Shutdown Cooling (SDC) System, the Containment Spray (CS) System, and the Reactor Core Isolation Cooling (RCIC) System are not rendered inoperable due to gas accumulation and to provide allowances which permit performance of the revised verification. The changes are being made to address the concerns discussed in NRC Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems.'' The proposed changes are based on Revision 2 of NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-523, ``Generic Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 (ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of Availability for TSTF-523, Revision 2, for plant- specific adoption using the consolidated line item improvement process, in the Federal Register on January 15, 2014 (79 FR 2700). 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 provided 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 Surveillance Requirements (SRs) that require verification that the Emergency Core Cooling Systems (ECCS), the Residual Heat Removal (RHR) System/Shutdown Cooling (SDC) System, the Containment Spray (CS) System, and the Reactor Core Isolation Cooling (RCIC) 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/SDC System, the CS System, and the RCIC 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/SDC System, the CS System, and the RCIC System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change clarifies requirements for management of 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401. NRC Branch Chief: David L. Pelton. Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50- 306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of amendment request: June 29, 2015. A publicly-available version is in ADAMS under Accession No. ML15187A259. Description of amendment request: The proposed amendment would revise or add technical specification (TS) surveillance requirements (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Residual Heat Removal (RHR) System, and the Containment Spray (CS) System are not rendered inoperable due to gas accumulation and to provide allowances which permit performance of the revised verification. The changes are being made to address the concerns discussed in NRC Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems.'' The proposed changes are [[Page 61485]] based on Revision 2 of NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-523, ``Generic Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 (ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of Availability for TSTF- 523, Revision 2, for plant-specific adoption using the consolidated line item improvement process, in the Federal Register on January 15, 2014 (79 FR 2700). 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 provided 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 or adds Surveillance Requirements (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Residual Heat Removal (RHR) System, and the Containment Spray (CS) 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 licensing basis 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 [revises or] adds SRs that require verification that the ECCS, the RHR System, and the CS 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 licensing basis 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 or] adds SRs that require verification that the ECCS, the RHR System, and the CS 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 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 licensing basis 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401. NRC Branch Chief: David L. Pelton. Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: July 24, 2015. A publicly-available version is in ADAMS under Accession No. ML15205A276. Description of amendment request: The amendment would revise the Technical Specification (TS) Surveillance Requirements (SRs), which currently require operating ventilation systems with charcoal filters for a 10-hour period at a monthly frequency. The SRs would be revised to require operation of the systems for 15 continuous minutes at a monthly frequency. The proposed amendment is consistent with NRC- approved Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month,'' as published in the Federal Register on September 20, 2012 (77 FR 58428), with variations due to plant- specific nomenclature. The changes would revise TS 3.2, Table 3-5; SR Items 10a.3.a, ``Control Room Air Filtration System (CRAFS)''; 10b.3.a, ``Spent Fuel Pool Storage Area Filtration System (SFPSAFS)''; and 10c.3.a, ``Safety Injection Pump Room Air Filtration System (SIPRAFS),'' and TS 3.6(3)c, ``Containment Recirculating Air Cooling and Filtering System,'' also known as the Containment Air Cooling and Filtering System (CACFS). 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 replaces an existing SR to operate the CRAFS for ten (10) continuous hours every month with heaters operating with a requirement to operate the system for 15 continuous minutes every month with heaters operating. The proposed change also replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the CACFS for ten (10) hours every month with a requirement to operate these systems for 15 continuous minutes every month. These systems are not accident initiators and therefore, these changes do not involve a significant increase in the probability of an accident. The proposed system and filter testing changes are consistent with current regulatory guidance for these systems. The proposed changes continue to ensure that these systems perform their design function, which may include mitigating accidents. Thus, the change does not involve a significant increase in the consequences of an accident. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change replaces an existing SR to operate the CRAFS for ten (10) continuous hours every month with heaters operating with a requirement to operate the system for 15 continuous minutes every month with heaters operating. The proposed change also replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the CACFS for ten (10) hours every month with a requirement to operate these systems for 15 continuous minutes every month. The change proposed for these ventilation systems does not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and/or the system components are capable of performing their intended safety functions. The change does not create new failure modes or mechanisms and no new accident precursors are generated. Therefore, it is concluded that this change does not create the possibility of a new or [[Page 61486]] 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 replaces an existing SR to operate the CRAFS for ten (10) continuous hours every month with heaters operating with a requirement to operate the system for 15 continuous minutes every month with heaters operating. The proposed change also replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the CACFS for ten (10) hours every month with a requirement to operate these systems for 15 continuous minutes every month. The design basis for the CRAFS heaters is to heat the incoming air, which reduces the relative humidity. The heater testing change proposed for the CRAFS will continue to demonstrate that the heaters are capable of heating the air and will perform their design function. The SFPSAFS, and the SIPRAFS are tested for adsorption at a relative humidity of [95 percent (%)] in accordance with RG [Regulatory Guide] 1.52, Revision 3, and do not require heaters for these systems to perform their specified safety function. The CACFS does not need to be tested similarly because the CACFS charcoal filters are not credited for the removal of radioiodines. The proposed change is consistent with regulatory guidance. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817. NRC Branch Chief: Michael T. Markley. Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: August 20, 2015. A publicly-available version is in ADAMS under Accession No. ML15233A494. Description of amendment request: The amendment would make administrative changes to update personnel and committee titles in the Technical Specifications (TSs), delete outdated or completed additional actions contained in Appendix B of the license, and relocate the definition of Process Control Program from the TSs to the Updated Safety Analysis Report (USAR). The changes are proposed by the licensee to use consistent terminology with Exelon Generation Company as part of their Operating Services Agreement. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes are administrative in nature, involving changes to personnel and committee titles, deletion and or re- location of requirements redundant to regulations, and deletion of conditions controlling the first performance of testing that has since been completed. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because: (1) the proposed amendment does not represent a change to the system design, (2) the proposed amendment does not alter, degrade, or prevent action described or assumed in any accident in the USAR from being performed, (3) the proposed amendment does not alter any assumptions previously made in evaluating radiological consequences, and [(4)] the proposed amendment does not affect the integrity of any fission product barrier. No other safety related equipment is affected by the proposed change. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed changes do not introduce any new accident initiators, nor do these changes reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function. 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 amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not alter the manner in which safety limits or limiting safety system settings are determined. The safety analysis acceptance criteria are not affected by these proposed changes. Further, the proposed changes do not change the design function of any equipment assumed to operate in the event of an accident. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817. NRC Branch Chief: Michael T. Markley. Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, San Luis Obispo County, California Date of amendment request: June 17, 2015, as supplemented by letter dated August 31, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15176A539 and ML15243A363, respectively. Description of amendment request: The amendments would revise the licensing bases to adopt the alternative source term (AST) as allowed by 10 CFR 50.67, ``Accident source term.'' The AST methodology, as established in NRC Regulatory Guide (RG) 1.183, ``Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession No. ML003716792), is used to calculate the offsite and control room radiological consequences of postulated accidents for DCPP, Unit Nos. 1 and 2. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This license amendment does not physically impact any system, structure, or component (SSC) that is a potential initiator of an accident. Therefore, implementation of AST, the AST assumptions and inputs, the proposed [Technical Specification (TS)] changes, and new [chi]/Q values have no impact on the probability for initiation of any design basis accident. Once the occurrence of an accident has been postulated, the new accident source term and [atmospheric dispersion factors ([chi]/Q)] values are inputs to analyses that evaluate the radiological consequences of the postulated events. Reactor coolant specific activity, testing criteria of charcoal filters, and the accident induced primary-to-secondary system leakage performance criterion are not initiators for any accident previously evaluated. The proposed change to require the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 is not an accident initiator for any [[Page 61487]] accident previously evaluated. The change in the classifications of a portion of the 40-inch Containment Penetration Area Ventilation line and a portion of the 2-inch gaseous radwaste system line is also not an accident initiator for any accident previously evaluated. Thus, the proposed TS changes and AST implementation will not increase the probability of an accident. The change to the decay time prior to fuel movement is not an accident initiator. Decay time is used to determine the source term for the dose consequence calculation following a potential [fuel handling accident (FHA)] and has no effect on the probability of the accident. Likewise, the change to the Control Room radiation monitors setpoint cannot cause an accident and the operation of containment spray during the recirculation phase is used for mitigation of a [loss-of-coolant accident (LOCA)], and thus not an accident initiator. As a result, there are no proposed changes to the parameters or conditions that could contribute to the initiation of an accident previously evaluated in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). As such, the AST cannot affect the probability of an accident previously evaluated. Regarding accident consequences, equipment and components affected by the proposed changes are mitigative in nature and relied upon once the accident has been postulated. The license amendment implements a new calculation methodology for determining accident consequences and does not adversely affect any plant component or system that is credited to mitigate fuel damage. Subsequently, no conditions have been created that could significantly increase the consequences of any accidents previously evaluated. Requiring that the 48-inch containment purge supply and exhaust valves be sealed closed during operating MODES 1, 2, 3, and 4 eliminates a potential path for radiological release following events that result in radioactive material releases to the containment, thus reducing potential consequences of the event. The steam generator tube inspection testing criterion for accident induced leakage is being changed, resulting in lower leakage rates, and thus less potential releases due to primary-to-secondary leakage. The auxiliary building ventilation system allowable methyl iodide penetration limit is being changed, which results in more stringent testing requirements, and thus higher filter efficiencies for reducing potential releases. Changes to the operation of the containment spray system to require operation during the recirculation mode are also mitigative in nature. While the plant design basis has always included the ability to implement containment spray during recirculation, this license amendment now requires operation of containment spray in the recirculation mode for dose mitigation. DCPP is designed and licensed to operate using containment spray in the recirculation mode. As such, operation of containment spray in the recirculation mode has already been analyzed, evaluated, and is currently controlled by Emergency Operating Procedures. Usage of recirculation spray reduces the consequence of the postulated event. Likewise, the additional shielding to the Control Room and the addition of a [high-efficiency particulate air (HEPA)] filter to the [Technical Support Center (TSC)] ventilation system reduces the consequences of the postulated event to the Control Room and TSC personnel. Lowering the limit for [Dose Equivalent XE-133 (DEX)] lowers potential releases. By reclassifying a portion of the 40-inch Containment Penetration Area Ventilation line and a portion of the 2-inch gaseous radwaste system line to PG&E Design Class I, these lines will be seismically qualified, thus assuring that post-LOCA release points are the same as those used for determining [chi]/Q values. The change to the decay time from 100 hours to 72 hours prior to fuel movement is an input to the FHA. Although less decay will result in higher released activity, the results of the FHA dose consequence analysis remain within the dose acceptance criteria of the event. Also, the radiation levels to an operator from a raised fuel assembly may increase due to a lower decay time, however, any exposure will continue to be maintained under 10 CFR 20 limits by the plant Radiation Protection Program. Plant-specific radiological analyses have been performed using the AST methodology, assumption and inputs, as well as new [chi]/Q values. The results of the dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. Implementing the AST involves no facility equipment, procedure, or process changes that could significantly affect the radioactive material actually released during an event. Subsequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated. Based on the above discussion, 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 accident from any accident previously evaluated? Response: No. This license amendment does not alter or place any SSC in a configuration outside its design or analysis limits and does not create any new accident scenarios. The AST methodology is not an accident initiator, as it is a method used to estimate resulting postulated design basis accident doses. The proposed TS changes reflect the plant configuration that supports implementation of the new methodology and supports reduction in dose consequences. DCPP is designed and licensed to operate using containment spray in the recirculation mode. This change will not affect any operational aspect of the system or any other system, thus no new modes of operation are introduced by the proposed change. The function of the radiation monitors has not changed; only the setpoint has changed as a result of an assessment of all potential release pathways. The continued operation of containment spray and the radiation monitor setpoint change do not create any new failure modes, alter the nature of events postulated in the UFSAR, nor introduce any unique precursor mechanism. Requiring the 48-inch containment purge valves to be sealed closed during operating MODES 1, 2, 3, and 4 does not introduce any new accident precursor. This change only eliminates a potential release path for radionuclides following a LOCA. The proposed TS testing criteria for the auxiliary building ventilation system charcoal filters and the proposed performance criteria for steam generator tube integrity also cannot create an accident, but results in requiring more efficient filtration of potentially released iodine and less allowable primary-to-secondary leakage. The proposed changes to the DEX activity limit, the TS terminology, and the decay time of the fuel before movement are also unrelated to accident initiators. The only physical changes to the plant being made in support of AST is the addition of Control Room shielding in an area previously modified, the addition of a HEPA filter at the intake of the TSC normal ventilation system, and the upgrade to the damper actuators, pressure switches, and damper solenoid valves to support reclassifying a portion of the Containment Penetration Area Ventilation line to PG&E Design Class I. Both Control Room shielding and HEPA filtration are mitigative in nature and do not have any impact on plant operation or system response following an accident. The Control Room modification for adding the shielding will meet applicable loading limits, so the addition of the shielding cannot initiate a failure. Upgrading damper actuators, pressure switches, and damper solenoid valves involve replacing existing components with components that are PG&E Design Class I. Therefore, the addition of shielding, a HEPA filter, and upgrading components cannot create a new or different kind of accident. Since the function of the SSCs has not changed for AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents. Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Implementing the AST is relevant only to calculated dose consequences of potential design basis accidents evaluated in Chapter 15 of the UFSAR. The changes proposed in this license amendment involve the use of a new analysis methodology and related regulatory acceptance criteria. New atmospheric dispersion factors, which are based on site specific meteorological data, were calculated in accordance with regulatory guidelines. The proposed TS, TS Bases, and UFSAR changes reflect the plant configuration that will support implementation of the new methodology and result in operation in accordance with regulatory guidelines that support the revisions to the radiological analyses of the limiting design basis accidents. Conservative [[Page 61488]] methodologies, per the guidance of RG 1.183, have been used in performing the accident analyses. The radiological consequences of these accidents are all within the regulatory acceptance criteria associated with the use of AST methodology. The change to the minimum decay time prior to fuel movement results in higher fission product releases after a FHA. However, the results of the FHA dose consequence analysis remain within the dose acceptance criteria of the event. The proposed changes continue to ensure that the dose consequences of design basis accidents at the exclusion area, low population zone boundaries, in the TSC, and in the Control Room are within the corresponding acceptance criteria presented in RG 1.183 and 10 CFR 50.67. The margin of safety for the radiological consequences of these accidents is provided by meeting the applicable regulatory limits, which are set at or below the 10 CFR 50.67 limits. An acceptable margin of safety is inherent in these limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120. NRC Branch Chief: Michael T. Markley. South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Units 2 and 3, Fairfield County, South Carolina Date of amendment request: June 30, 2015. A publicly-available version is in ADAMS under Accession No. ML15181A470. Description of amendment request: The amendment request proposes changes to the Main Control Room Emergency Habitability System (VES) configuration and equipment safety designation. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The design functions of the VES for the main control room (MCR) are to provide breathable air, maintain positive pressurization relative to the outside, provide cooling of MCR equipment and facilities, and provide passive air filtration within the MCR boundary. The VES is designed to satisfy these functions for up to 72 hours following a design basis accident. The proposed changes to the ASME [American Society of Mechanical Engineers] safety classification of components, equipment orientation and configuration, addition and deletion of components, and correction to the number of emergency air storage tanks would not adversely affect any design function. The proposed changes maintain the design function of the VES with safety-related equipment and system configuration consistent with the descriptions in UFSAR [Updated Final Safety Analysis Report] Subsection 6.4.2. The proposed changes do not affect the support or operation of mechanical and fluid systems. There is no change to the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes described create any new accident precursors. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes to revise the VES design related to the ASME safety classification, equipment orientation and configuration, addition and deletion of components, and correction to the number of emergency air storage tanks maintains consistency with the design function information in the USFAR. The proposed changes do not create a new fault or sequence of events that could result in a radioactive release. The proposed changes would not affect any safety-related accident mitigating function. 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 amendment involve a significant reduction in a margin of safety? Response: No. The proposed changes do not affect the ability of the VES to maintain the safety-related functions to the MCR. The VES continues to meet the requirements for which it was designed and continues to meet the regulations. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514. NRC Branch Chief: Lawrence J. Burkhart. 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. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of amendment request: June 17, 2015, as supplemented by letters dated July 14, August 28, and September 3, 2015. Publicly- available versions are in ADAMS under Accession Nos. ML15170A474, ML15197A357, ML15243A044, and ML15246A638, respectively. Brief description of amendment request: The amendment would modify the technical specifications to define support systems needed in the first 48 hours after a unit shutdown when steam generators are not available for heat removal. The amendment would also make changes consistent with Technical Specification Task Force Traveler-273-A, Revision 2, to provide clarifications related to the requirements of the Safety Function Determination Program. Date of publication of individual notice in Federal Register: September 15, 2015 (80 FR 55383). [[Page 61489]] Expiration date of individual notice: October 15, 2015 (public comments); November 16, 2015 (hearing requests). Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of amendment request: August 13, 2015. A publicly-available version is in ADAMS under Accession No. ML15225A344. Brief description of amendment request: To revise a current License Condition (Section 2.F) regarding the Fire Protection Program and propose a new License Condition regarding a fire protection requirement. Date of publication of individual notice in Federal Register: September 4, 2015 (80 FR 53581). Expiration date of individual notice: October 5, 2015 (public comments); November 3, 2015 (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. 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 application of amendments: June 30, 2014, as supplemented by letter dated June 8, 2015. Brief description of amendments: The amendments revised the Technical Specifications related to Technical Specification 3.5.2 by reducing the allowed maximum Rated Thermal Power at which each unit can operate when select High Pressure Injection system equipment is inoperable. Date of Issuance: September 24, 2015. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: 395, 397 and 396. A publicly-available version is in ADAMS under Accession No. ML15166A387; documents related to these amendments are listed in the Safety Evaluation enclosure with the amendments. Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and the technical specifications. Date of initial notice in Federal Register: September 16, 2014 (79 FR 55510). The supplement dated June 8, 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. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 24, 2015. No significant hazards consideration comments received: No. Duke Energy Progress, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Hartsville, South Carolina Date of amendment request: February 10, 2014, as supplemented by letters dated April 4, 2014, August 28, 2014, and September 4, 2015. Brief description of amendment: The amendment revised Technical Specification (TS) 3.3.1 for the Reactor Protection System Instrumentation Turbine Trip function on Low Auto Stop Oil Pressure to a Turbine Trip function on Low Electro-Hydraulic (EH) Fluid Oil Pressure. The amendment revised the Allowable Value and Nominal Trip Setpoint and revised the TS by applying additional testing requirements listed in Technical Specification Task Force (TSTF) Traveler TSTF-493- A, Revision 4, ``Clarify Application of Setpoint Methodologies for Limiting Safety System Setting Functions,'' for Low EH Fluid Oil Pressure trip. Date of issuance: September 22, 2015. Effective date: As of the date of issuance and shall be implemented within 120 days of completion of the modification during Refueling Outage 31 in fall of 2018. Amendment No.: 243. A publicly-available version is in ADAMS under Accession No. ML15040A073; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-23: 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 August 28, 2014, and September 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 September 22, 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: September 2, 2014, as supplemented by letters dated April 23 and August 20, 2015. Brief description of amendment: The amendment revised the Surveillance Requirements (SRs) related to gas accumulation for the emergency core cooling system and reactor core isolation cooling system. The amendment also adds new SRs related to gas accumulation for the residual heat removal and shutdown cooling systems. The NRC staff has concluded that the Technical Specification (TS) changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas [[Page 61490]] Accumulation,'' dated February 21, 2013, as part of the consolidated line item improvement process. The TS Bases associated with these SRs were also changed. Date of issuance: September 21, 2015. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 188. A publicly-available version is in ADAMS under Accession No. ML15195A061; 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: January 6, 2015 (80 FR 522). The supplements dated April 23 and August 20, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 6, 2015 (80 FR 522). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2015. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: November 3, 2014, as supplemented by letter dated April 14, 2015. Brief description of amendments: The amendments added new Limiting Conditions for Operation (LCOs) 3.0.5 and 3.0.6 to the Applicability section of the Technical Specifications (TSs). LCO 3.0.5 establishes an allowance for restoring equipment to service under administrative controls when the equipment has been removed from service or declared inoperable to comply with TS Action requirements. LCO 3.0.6 provides actions to be taken when the inoperability of a support system results in the inoperability of the related supported systems. In addition, the amendments added the Safety Function Determination Program to the Administrative Controls section of the TSs. This program is intended to ensure that a loss of safety function is detected and appropriate actions are taken when LCO 3.0.6 is entered. Date of issuance: September 15, 2015. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 219 (Unit 1) and 181 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML15218A501; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-39 and NPF-85: Amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: December 23, 2014 (79 FR 77046). The supplemental letter dated April 14, 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 September 15, 2015. No significant hazards consideration comments received: No. National Institute of Standards and Technology (NIST), Docket No. 50- 184, Center for Neutron Research, National Bureau of Standards Test Reactor (NBSR), Montgomery County, Maryland Date of amendment request: June 23, 2014, as supplemented on August 20, 2014, February 26, 2015, and June 12, 2015. Brief description of amendment: The amendment revised the NIST NBSR's Technical Specifications Section 3.6 and Surveillance Requirement 4.6, pertaining to the NIST reactor emergency power system, which adds specifications and testing requirements for the new valve- regulated lead acid batteries of the new uninterruptable power supplies. Date of issuance: September 10, 2015. Effective date: As of the date of issuance. Amendment No.: 10. A publicly-available version is in ADAMS under Accession No. ML15237A146; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. TR-5: Amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 7, 2015 (80 FR 38760). The supplemental letters dated February 26, 2015, and June 12, 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 September 10, 2015. No significant hazards consideration comments received: No. STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas Date of amendment request: April 29, 2015. Brief description of amendment: The amendments revised the Updated Final Safety Analysis Report (UFSAR) Table 15.6-17 to correct errors introduced in UFSAR Revisions 16 and 17. Date of issuance: September 22, 2015. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: Unit 1-207; Unit 2-195. A publicly-available version is in ADAMS under Accession No. ML15209A641; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Facility Operating Licenses. Date of initial notice in Federal Register: July 21, 2015 (80 FR 43130). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 22, 2015. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of amendment request: April 6, 2015, as supplemented by letter dated July 15, 2015. Brief description of amendment: The amendment revised the Technical Specifications by modifying the acceptance criteria for the emergency diesel generator steady-state frequency range in associated surveillance requirements. Date of issuance: September 17, 2015. Effective date: As of the date of issuance and shall be implemented after the issuance of the Facility Operating License for Unit 2. Amendment No.: 102. A publicly-available version is in ADAMS under Accession No. ML15230A155; [[Page 61491]] documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NFP-90: Amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: May 26, 2015 (80 FR 30103). The supplemental letter dated July 15, 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 September 17, 2015. No significant hazards consideration determination comments received: No. V. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances) 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 for the amendment 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. Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual notice of consideration of issuance of amendment, proposed no significant hazards consideration determination, and opportunity for a hearing. For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments. In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible. Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved. The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 51.12(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 application for amendment, (2) the amendment to Facility Operating License or Combined License, as applicable, 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. A. Opportunity To Request a Hearing and Petition for Leave To Intervene The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC's Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or by email to [email protected]. 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 [[Page 61492]] 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. 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 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect. Arizona Public Service Company, Docket No. 50-529, Palo Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona Date of amendment request: September 4, 2015, as supplemented by letter dated September 15, 2015. Description of amendment request: The amendment added a Note to Technical Specification Surveillance Requirement (SR) 3.1.5.3, Control Element Assembly (CEA) freedom of movement surveillance, such that Unit 2, CEA 88 may be excluded from the remaining quarterly performance of the SR in Unit 2, Cycle 19 due to a degraded upper gripper coil. The amendment allows the licensee to delay exercising CEA 88 until after repairs can be made during the upcoming fall 2015 outage. Date of issuance: September 25, 2015. Effective date: This license amendment is effective as of the date of issuance and shall be implemented prior to the SR 3.1.5.3 performance due date for CEA 88 in Unit 2, Cycle 19. Amendment No.: 196. A publicly-available version is in ADAMS under Accession No. ML15266A005; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-51: Amendment revised the Operating License and Technical Specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Public notice of the proposed amendment was published in the Arizona Republic, located in Phoenix, Arizona, from September 21 through September 22, 2015. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments were received. The supplemental letter dated September 15, 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 NSHC determination as published in the Arizona Republic. The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated September 25, 2015. Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817. NRC Branch Chief: Michael T. Markley. Dated at Rockville, Maryland, this 1st day of October 2015. For the Nuclear Regulatory Commission. Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2015-25860 Filed 10-9-15; 8:45 am] BILLING CODE 7590-01-P
| Category | Regulatory Information | |
| Collection | Federal Register | |
| sudoc Class | AE 2.7: GS 4.107: AE 2.106: | |
| Publisher | Office of the Federal Register, National Archives and Records Administration | |
| Section | Notices | |
| Action | Biweekly notice. | |
| Dates | Comments must be filed by November 12, 2015. A request for a hearing must be filed by December 14, 2015. | |
| Contact | Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1384, email: [email protected] | |
| FR Citation | 80 FR 61476 |