Source: https://www.federalregister.gov/documents/2015/06/09/2015-12783/applications-and-amendments-to-facility-operating-licenses-and-combined-licenses-involving-proposed
Timestamp: 2017-09-24 18:22:02
Document Index: 726265851

Matched Legal Cases: ['§\u20092', 'art 2', 'art] 50', 'art] 50', 'art] 50', 'art 50']

Comments must be filed by July 9, 2015. A request for a hearing must be filed by August 10, 2015. Any potential party as defined in Sec. 2.4 of Title 10 of the Code of Federal Regulations (10 CFR), who believes access to SUNSI is necessary to respond to this notice must request document access by June 19, 2015.
32616-32622 (7 pages)
NRC-2015-0123
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating, Unit 3, Westchester County, New York
Florida Power and Light Company (FPL), Docket No. 50-389, St. Lucie Plant, Unit 2 (SL-2), St. Lucie County, Florida
https://www.federalregister.gov/d/2015-12783 https://www.federalregister.gov/d/2015-12783
The U.S. Nuclear Regulatory Commission (NRC) received and is considering approval of two amendment Start Printed Page 32617requests. The amendment requests are for Indian Point Nuclear Generating, Unit 3; and St. Lucie Plant, Unit 2. The NRC proposes to determine that each amendment request involves no significant hazards consideration. In addition, each amendment request contains sensitive unclassified non-safeguards information (SUNSI).
Comments must be filed by July 9, 2015. A request for a hearing must be filed by August 10, 2015. Any potential party as defined in § 2.4 of Title 10 of the Code of Federal Regulations (10 CFR), who believes access to SUNSI is necessary to respond to this notice must request document access by June 19, 2015.
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0123. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
Please refer to Docket ID NRC-2015-0123 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-2015-0123.
Please include Docket ID NRC-2015-0123, facility name, unit number(s), application date, and subject in your comment submission.
Within 60 days after the date of publication of this notice, any person(s) Start Printed Page 32618whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.
Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions Start Printed Page 32619should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.
Date of amendment request: February 12, 2015. A publicly-available version is in ADAMS under Accession No. ML15061A275.
Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would revise the Reactor Coolant System (RCS) heatup and cooldown limitations in the Unit 3 Technical Specification (TS) 3.4.3, and the Low Temperature Overpressure Protection System requirements in Unit 3 TS 3.4.12 in order to compensate for an increased service life. The existing RCS pressure and temperature limits are valid for a lifetime burnup of 27.2 Effective Full Power Years (EFPY), which is estimated to be reached by September 2015, and the revised limits are for a lifetime burnup of 37 EFPY, which are not anticipated to be reached until December 2023.
Entergy has determined that this proposed TS change does not involve a significant hazards consideration as defined by 10 CFR 50.92(c).
The proposed TS [technical specification] changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Except for a setpoint change for automatic PORV [power-operated operated relief valve] actuation, there are no physical changes to the plant being introduced by the proposed changes to the heatup and cooldown limitation curves. The proposed changes do not modify the RCS [reactor coolant system] pressure boundary. That is, there are no changes in operating pressure, materials, or seismic loading. The proposed changes do not adversely affect the integrity of the RCS pressure boundary such that its function in the control of radiological consequences is affected. The proposed heatup and cooldown limitation curves were generated in accordance with the fracture toughness requirements of 10 CFR [Part] 50, Appendix G, and ASME B&PV code [American Society of Mechanical Engineers Boiler & Pressure Vessel code], Section XI, Appendix G, to the 1998 edition through the 2000 Addenda. The proposed heatup and cooldown limitation curves were established in compliance with the methodology used to Start Printed Page 32620calculate and predict effects of radiation on embrittlement of RPV [reactor pressure vessel] beltline materials. Use of this methodology provides compliance with 10 CFR [Part] 50 Appendix G and provides margins of safety that ensure non-ductile failure of the RPV and the other RCS carbon and low alloy steel components will not occur. The proposed heatup and cooldown limitation curves prohibit operation in regions where it is possible for non-ductile failure of carbon and low alloy RCS materials to occur. Hence, the primary coolant pressure boundary integrity will be maintained throughout the limit of applicability of the curves, 37 EFPY [effective full-power years].
Operation within the proposed Low Temperature Overpressure Protection System (LTOP) limits ensures that overpressurization of the RCS at low temperatures will not result in component stresses in excess of those allowed by the ASME B&PV Code Section Xl Appendix G.
The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Further, the proposed changes to the heatup and cooldown limitation curves and the LTOP limits do not affect any activities or equipment other than the RCS pressure boundary and do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Consequently, the proposed changes do not create the possibility of a new or different kind of accident, from any accident previously evaluated.
The proposed TS changes do not involve a significant reduction in the margin of safety. The revised heatup and cooldown limitation curves and LTOP limits are established in accordance with current regulations and the ASME B&PV Code 1998 edition through the 2000 Addenda, Appendix G. These proposed changes are acceptable because the ASME B&PV Code maintains the margin of safety required by 10 CFR [Part] 50, Appendix G. Because operation will be within these limits, the RCS materials will continue to behave in a non-brittle manner consistent with the original design bases.
Attorney for licensee: Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, New York 10601.
Date of amendment request: December 30, 2014, as supplemented by letter dated March 23, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15002A091 and ML15084A011, respectively.
Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would revise the Technical Specifications (TSs) to allow for the use of AREVA fuel at SL-2. Additionally, pursuant to 10 CFR 50.12, FPL requests an exemption from the provisions of 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors,” and appendix K to 10 CFR part 50, “ECCS Evaluation Models,” to allow for the use of M5® fuel rod cladding in future core reload applications for SL-2.
The proposed changes for St. Lucie Units 2 revise the Technical Specification (TS) 5.3.1 to include M5® cladding, delete the linear heat rate surveillance requirement with W(z) in TS 4.2.1.3 and include previously approved AREVA Topical Reports in the list of COLR [core operating limits report] methodologies in TS 6.9.1.11. [Another] change is in TS License Condition 3.N, which is related to future analysis of the current fuel and is considered an administrative change, all as a result of changing the fuel supplier.
The fuel assembly design is not an initiator to any accident previously evaluated. Therefore, there is no significant increase in the probability of any accident previously evaluated. However, the fuel design parameters and the correlations used in the analyses supporting the operation of St. Lucie Unit 2 with the new proposed AREVA fuel are dependent on the fuel assembly design. All the analyses, potentially impacted by the fuel design, have been re-analyzed using the correlations and the methodology applicable to the proposed fuel design and previously approved by the NRC for similar applications. There are no changes to any limits specified in the Technical Specifications. M5® cladding to be used in the proposed AREVA fuel design has been previously approved by the NRC for PWR [pressurized-water reactor] applications, including St. Lucie Unit 1. The core design peaking factors remain unchanged from the current analyses values, except for the large break LOCA [loss-of-coolant accident] which is shown to meet all the 10 CFR 50.46 criteria with the increased peak linear heat rate limit.
No new or different accidents result from utilizing the proposed AREVA CE [combustion engineering] 16x16 fuel design. Other than the fuel design change, the proposed license amendment does not involve a physical alteration of the plant or plant systems (i.e., no new or different type of equipment will be installed which would create a new or different kind of accident). The change to the linear heat rate surveillance requirement, when operating on excore detector monitoring system, and the use of M5® cladding do not affect or create any accident initiator. There is no change to the methods governing normal plant operation and the changes do not impose any new or different operating requirements. The core monitoring system remains unchanged.
The changes proposed in this license amendment request are related to the fuel design with M5® cladding and the methodology supporting the analysis of accidents impacted by the fuel design change. The analysis methods used are previously approved by the NRC for similar applications. The change to the surveillance requirement for the linear heat rate does not change any accident analysis requirements. The fuel design limits related to the DNBR [departure from nuclear boiling ratio] and fuel centerline melt remain consistent with the limits previously approved for the proposed fuel design change. The overpressure limits for the reactor coolant system integrity and the containment integrity remain unchanged. All the analyses performed to support the fuel design change meet all applicable acceptance criteria. The LOCA analyses, with the peak linear heat rate limit increase, continue to meet all the applicable 10 CFR 50.46 acceptance criteria, Start Printed Page 32621and thus the proposed changes do not affect margin to safety for any accidents previously evaluated.
Based on the previous discussion of the amendment request, it is determined that the proposed amendment does not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any previously evaluated; nor (3) involve a significant reduction in a margin of safety. [Therefore,] the amendment does not involve a significant hazards consideration.
If challenges to the NRC staff determinations are filed, these procedures give way to the normal process for litigating disputes concerning access to information. The availability of interlocutory review by the Commission of orders ruling on such NRC staff determinations (whether Start Printed Page 32622granting or denying access) is governed by 10 CFR 2.311.[3]
Dated at Rockville, Maryland, this 20th day of May, 2015.
[FR Doc. 2015-12783 Filed 6-8-15; 8:45 am]