ML13107A016

Seabrook Station, Unit 1, Issuance of Amendment Regarding the License Amendment Request 13-02, Application to Revise Technical Specifications to Adopt TSTSF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Sel

ML13107A016
Person / Time
Site:	Seabrook
Issue date:	10/25/2013
From:	Lamb J G Plant Licensing Branch 1
To:	Walsh K NextEra Energy Seabrook
Lamb J G
References
TAC MF1372
Download: ML13107A016 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 25, 2013 SUBJECT: SEABROOK STATION, UNIT NO. 1 -ISSUANCE OF AMENDMENT REGARDING LICENSE AMENDMENT REQUEST 13-02, APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-510, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NO. MF1372) Dear Mr. Walsh: The Commission has issued the enclosed Amendment No. 138 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1. This amendment consists of changes to the facility technical specifications (TSs) in response to your application dated March 27, 2013, as supplemented by letter dated June 25, 2013. The amendment modifies TS requirements regarding steam generator tube inspections and reporting as described in TS Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," using the Consolidated Line Item Improvement Process. A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-443 Enclosures: 1. Amendment No. 138 to NPF-86 2. Safety Evaluation cc w/encls: Distribution via Listserv G. Lamb, Senior Project Manager t Licensing Branch 1-2 Di sion of Operating Reactor Licensing 0 ice of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY SEABROOK. LLC. ET AL.* DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 138 License No. NPF-86 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment filed by NextEra Energy Seabrook, LLC, et al., (the licensee) dated March 27, 2013, as supplemented June 25, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 1 0 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. *NextEra Energy Seabrook, LLC is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Light Plant and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

-2 -2. Accordingly, the license is amended by changes to paragraphs 2.C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 138, and the Environmental Protection Plan contained in Appendix Bare incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days. Attachment: Changes to the License and Technical Specifications Date of Issuance: October 25, 2013 FOR THE NUCLEAR REGULATORY COMMISSION )?12 D. Veronica Rodriguez, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 138 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. Remove 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the area of change. Remove Insert 3/4 4-13 3/4 4-13 6-11 6-11 6-12 6-12 6-13 6-13 6-14 6-14 6-14a 6-14a 6-14b 6-14b 6-14c 6-21 6-21 6-21a 6-21a

-3 -(4) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein; and (7) DELETED C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 138*, and the Environmental Protection Plan contained in t Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) License Transfer to FPL Energy Seabrook, LLC**

• Implemented a. On the closing date(s) of the transfer of any ownership interests in Seabrook Station covered by the Order approving the transfer, FPL Energy Seabrook, LLC**, shall obtain from each respective transferring owner all of the accumulated decommissioning trust funds for the facility, and ensure the deposit of such funds and additional funds, if necessary, into a decommissioning trust or trusts for Seabrook Station established by FPL Energy Seabrook, LLC**, such that the amount of such funds deposited meets or exceeds the amount required under 10 CFR 50.75 with respect to the interest in Seabrook Station FPL Energy Seabrook, LLC**, acquires on such dates(s). **On April 16, 2009, the name "FPL Energy Seabrook, LLC" was changed to "NextEra Energy Seabrook, LLC". AMENDMENT NO. 138 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERA TORS LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: --------------------------------------NOTE-----------------------------------------Separate action entry is allowed for each steam generator tube. a. With one or more steam generator tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program, 1. Within 7 days, verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and 2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or steam generator tube inspection. b. With steam generator tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program. 4.4.5.2 Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection. SEABROOK UNIT-1 3/4 4-13 Amendment 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications. a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews. b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following: 1. A change in the TS incorporated in the license or 2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59. c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR. d. Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). k. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following: a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met. SEABROOK -UNIT 1 6-11 Amendment No. 34, 55, 67, 88, 104, -1-1-a, 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE. 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm total or 500 gpd through any one SG. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage." c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. SEABROOK-UNIT 1 6-12 Amendment No. 34, 39, 104, 109, 115, 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria: Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection. d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to..: tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The portion of the tube below 15.21 inches from the top of the tubesheet is excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation. 2. After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection SEABROOK-UNIT 1 6-13 Amendment No. 34, 104, 1 OQ, 115, 123, 131,138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the remaining life of the SGs, inspect 100 % of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods. 3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary to secondary leakage. SEABROOK-UNIT 1 6-14 Amendment No. 34, 78, 104, 115, 119, 123, 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) I. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Makeup Air and Filtration System (CREMAFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements: a. The definition of the CRE and the CRE boundary. b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance. c. Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREMAFS, operating at a flow rate of less than or equal to 600 CFM at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary. SEABROOK-UNIT 1 6-14a Amendment No. 119, 126, 134, 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) e. The quantitative limits on unfiltered air in-leakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph c. The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis. f. The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered in-leakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively. m. Reactor Coolant Pump Flywheel Inspection Program In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975 at least once every 20 years. In lieu of Position C.4.b(1) and C.4.b(2), this inspection shall be by either of the following examinations: a. An in-place examination, utilizing ultrasonic testing, over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or b. A surface examination, utilizing magnetic particle testing and/or penetrant testing, of the exposed surfaces of the disassembled flywheel. SEABROOK-UNIT 1 6-14b Amendment No. 449, 138 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.8 REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1 In addition to the reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. STARTUP REPORT 6.8.1.1 A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station. The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed. SEABROOK-UNIT 1 6-14c Amendment No. 138 ADMINISTRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector. STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, Steam Generator (SG) Program. The report shall include: a. The scope of inspections performed on each SG, b. Degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each degradation mechanism, f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, g. The results of condition monitoring, including the results of tube pulls and situ testing, h. The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, i. The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and SEABROOK-UNIT 1 6-21 Amendment No. 22, 66, 88, 104, 107, 115, 123, 131,138 ADMINISTRATIVE CONTROLS 6.8.1.7 (Continued) j. The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided. SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report. 6.9 (THIS SPECIFICATION NUMBER IS NOT USED) SEABROOK-UNIT 1 6-21a Amendment No. 123, 131, 138 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 138 TO FACILITY OPERATING LICENSE NO. NPF-86 1.0 INTRODUCTION NEXTERA ENERGY SEABROOK. LLC SEABROOK STATION. UNIT NO. 1 DOCKET NO. 50-443 By application dated March 27, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13093A 151 ), as supplemented by letter dated June 25, 2013 (ADAMS Accession No. ML 13183A 1 08), Next Era Energy Seabrook, LLC (NextEra, the licensee) requested changes to the technical specifications (TSs) for Seabrook Station, Unit 1 (Seabrook). The amendment modifies TS requirements regarding steam generator (SG) tube inspections and reporting, as described in TS Task Force (TSTF)-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," using the Consolidated Line Item Improvement Process (CLIIP). The supplement dated June 25, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on April 30, 2013 (78 FR 25316). Specifically, per the application, the proposed change revises Specifications 3.4.5, "Steam Generators;" 6.7.6.k, "Steam Generator (SG) Program;" and 6.8.1.7, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications. The proposed amendment is consistent with TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." The licensee stated that the license amendment request (LAR) is consistent with the Notice of Availability of TSTF-51 0, Revision 2, announced in the Federal Register on October 27, 2011 (76 FR 66763), as part of the CLIIP. The current Standard TS (STS) requirements in the above specifications were established in May 2005 with the U.S. Nuclear Regulatory Commission (NRC) staff's approval of TSTF-449, Revision 4, "Steam Generator Tube Integrity" (NRC Federal Register Notice of Availability [70 FR 24126]). The TSTF-449 changes to the STS incorporated a new, largely based approach for ensuring the integrity of the SG tubes is maintained. The based requirements were supplemented by prescriptive requirements relating to tube

-2 -inspections and tube repair limits to ensure that conditions adverse to quality are detected and corrected on a timely basis. As of September 2007, the TSTF-449, Revision 4, changes were adopted in the plant TS for all pressurized-water reactors (PWRs). The proposed changes in TSTF-51 0, Revision 2, reflect licensees' early implementation experience with respect to the TSTF-449, Revision 4. TSTF-510 characterizes the changes as editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with implementing industry documents, and usability without changing the intent of the requirements. The proposed changes are an improvement to the existing SG inspection requirements and continue to provide assurance that the plant licensing basis will be maintained between SG inspections. 2.0 REGULATORY EVALUATION The SG tubes in PWRs have a number of important safety functions. These tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied upon to maintain primary system pressure and inventory. As part of the RCPB, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system and are relied upon to isolate the radioactive fission products in the primary coolant from the secondary system. In addition, the SG tubes are relied upon to maintain their integrity to be consistent with the containment objectives of preventing uncontrolled fission product release under conditions resulting from core damage during severe accidents. Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.36, "Technical Specifications," establishes the requirements related to the content of the TS. Pursuant to 1 0 CFR 50.36, TSs are required to include items in the following categories related to this issue: limiting conditions for operations (LCOs), surveillance requirements (SRs), and administrative controls. LCOs, SRs and administrative controls in the Seabrook TS relevant to SG tube integrity are in 3.4.5, "Steam Generators," 6.7.6.k, "Steam Generator (SG) Program," and 6.8.1.7, "Steam Generator Tube Inspection Report." The regulations in 10 CFR 50.36(c)(5) define administrative controls as "the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner." Programs established by the licensee to operate the facility in a safe manner, including the SG Program, are listed in the administrative controls section of the TS. Seabrook's SG Program is defined in TS sections as stated above. Specification 6.7.6.k requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. SG tube integrity is maintained by meeting the performance criteria specified in TS 6.7.6.k.b for structural and leakage integrity, consistent with the plant design and licensing basis. Specification 6.7.6.k.a requires that a condition monitoring assessment be performed during each outage in which the SG tubes are inspected, to confirm that the performance criteria are being met. Specification 6.7.6.k.d includes provisions regarding the scope, frequency, and methods of SG tube inspections. These provisions require that the inspections be performed with the objective of detecting flaws of any type that: (1) may be present along the length of a tube, from the tube-to-tubesheet weld at the tube inlet to the

-3-tube-to-tubesheet weld at the tube outlet; and (2) may satisfy the applicable tube plugging criteria. The applicable tube plugging criteria, specified in TS 6.7.6.k.c, are that tubes found during inservice inspections to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness shall be plugged, unless the tubes are permitted to remain in service through application of the alternate plugging criteria provided in TS 6.7.6.k.c. 3.0 TECHNICAL EVALUATION According to the application, NextEra is proposing the following administrative variations from the TS changes described in TSTF-510, Revision 2, or the applicable parts of the NRC staff's model safety evaluation, dated October 27, 2011:

• The Seabrook TSs utilize different numbering and titles than the Standard TSs on which TSTF-51 0 was based. Specifically, Seabrook TS 3.4.5, "Steam Generators," is equivalent toTS 3.4.20, "Steam Generator (SG) Tube Integrity," in TSTF-51 0. Seabrook TS 6.7.6.k contains the Steam Generator (SG) Program, which is designated TS 5.5.9 in TSTF-51 0; and Seabrook TS 6.8.1.7, "Steam Generator Tube Inspection Report," is equivalent toTS 5.6.7 in TSTF-51 0. These differences are administrative and do not affect the applicability of TSTF-510 to the Seabrook TSs.

• The proposed change corrects an administrative inconsistency in TSTF-510, Paragraph d.2 of the SG Tube Inspection Program. In Section 2.0, "Proposed Change," TSTF-51 0 states that references to "tube repair criteria" in Paragraph dare revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted, "If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated." (Emphasis added). NextEra does not have an approved tube repair criteria; therefore, the sentence is revised to state "tube plugging" criteria. Each proposed change, along with the licensee's applicable variations to the TS, is described individually below, followed by the NRC staff's assessment of the change. 3.1 TS 6.7.6.k, "Steam Generator (SG) Program" Proposed Change: The last sentence of the introductory paragraph currently states: "In addition, the Steam Generator Program shall include the following provisions: ... " The change would delete the word "provisions" such that the sentence would state: "In addition, the Steam Generator Program shall include the following: ... " The basis for this change is that subsequent paragraphs in TS 6.7.6.k start with "Provisions for ... " and the word "provisions" in the introductory paragraph is duplicative.

-4-Assessment: The NRC staff has reviewed TS 6.7.6.k and agrees that the word, "provisions" in the introductory paragraph is duplicative. The NRC staff agrees that the change is administrative in nature, and therefore is acceptable. 3.2 Paragraph 6. 7.6.k.b.1. "Structural integrity performance criterion" The first sentence currently states: Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. Proposed Change: Revise the sentence as follows: Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design-basis accidents. The basis for the change is that this sentence inappropriately includes anticipated transients in the description of normal operating conditions. Assessment: The NRC staff agrees the current wording is incorrect and that anticipated transients should be differentiated from normal operating conditions. Therefore, the NRC staff finds the change acceptable. 3.3. Paragraph 6.7.6.k.c, "Provisions for SG tube repair criteria," Paragraph 6.7.6.k.d, "Provisions for SG tube inspections," and LCO 3.4.5, "Steam Generators" Proposed Change: Change all references from "tube repair criteria" to "tube plugging criteria." This change is intended to be consistent with the treatment of SG tube repair throughout Specification 6.7.6.k. Assessment: The NRC staff finds that the proposed change provides a more accurate label of the criteria and, therefore, adds clarity to the specification. This is because one of two actions must be taken when the criteria are exceeded. One action is to remove the tube from service by plugging the tube at both tube ends. Per the application, Seabrook has no repair techniques included in their TS. As a result, the licensee can only plug defective tubes. Accordingly, the licensee has elected to adopt the change to "tube plugging" in accordance with TSTF-510, Revision 2, in all applicable places in the TS. The proposed change provides a more accurate description of the response to tube degradation, adding clarity to the TS. Therefore, the NRC staff finds the change acceptable.

-5 -3.4 Paragraph 6.7.6.k.d, "Provisions for SG tube inspections" Proposed Change: Change the term "assessment of degradation" to "degradation assessment" to be consistent with the terminology used in industry program documents. Assessment: The NRC staff agrees that the terminology should be consistent and finds the change acceptable. 3.5 Paragraph 6.7.6.k.d.1 Proposed change: The paragraph currently states: "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement." The change would replace "SG replacement" with "SG installation." The basis for the change is that it will allow the SG Program to apply to both existing plants and new plants. Assessment: The NRC staff finds that the change involves no change in the inspection requirement, will be more generic terminology from plant to plant and is acceptable. 3.6 Paragraph 6.7.6.k.d.2 The paragraph currently states: Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected. Proposed Change: Revise Paragraph 6.7.6.k.d.2 as follows: After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period

-6 -defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. a) After the first refueling outage following SG installation, inspect 1 00% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the remaining life of the SGs, inspect 1 00% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods. Assessment: Paragraph 6.7.6.k.d.2 in its current form and with the proposed changes is similar for each of the tube alloy types, but with differences that reflect the improved resistance of alloy 600 Thermally Treated (TT) to stress-corrosion cracking relative to alloy 600 Mill Annealed (MA) and the improved resistance of alloy 690 TT relative to both alloy 600 MA and alloy 600 TT. These differences include progressively larger maximum inspection interval requirements and sequential inspection periods (during which 100% of the tubes must be inspected) for alloy 600 MA, 600 TT, and 690 TT tubes, respectively. In addition, because of the longer maximum inspection intervals allowed for alloy 600 TT and 690 TT tubes, Paragraph 6.7.6.k.d.2 includes a restriction on the distribution of sampling over each sequential inspection period for alloy 600 TT. For SGs with alloy 600 TT tubing, the licensee proposes to move the last sentence of current Paragraph 6.7.6.k.d.2 to the beginning of the paragraph and make editorial changes to improve clarity. The NRC staff finds these changes to be of a clarifying nature, not changing the current intent of the sentence. However, the LAR also includes two changes to when inspections are performed as follows:

• The second inspection period would be revised from 90 to 96 effective full-power months (EFPM).

• The third and subsequent inspection periods would be revised from 60 to 72 EFPM. The licensee characterizes these changes as marginal increases for consistency with typical fuel cycle lengths that better accommodate the scheduling of inspections. The NRC staff notes that plants with alloy 600 TT SG tubes typically inspect at 18-or 36-month intervals (one or two fuel cycles, respectively) depending on whether stress-corrosion crack activity was observed during the most recent inspection. With these intervals, the last scheduled inspection during the first inspection period would occur at 108 months after the first refueling outage following SG installation. This is 12 months before the end of the first 120 EFPM inspection period. However, with the proposed changes to the length of the second and subsequent inspection periods, the NRC staff finds that the last scheduled inspections in the second and subsequent inspection periods will coincide exactly with the end of these periods.

-7-The proposed changes would generally increase the number of inspections in each of the second and subsequent inspection periods by up to one additional inspection. This could reduce the required average minimum sample size during these periods. However, inspection sample sizes will continue to be subject to Paragraph 6.7.6.k.d, which states that in addition to meeting the requirements of Paragraphs 6.7.6.k.d.1, d.2, and d.3, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure SG tube integrity is maintained until the next scheduled inspection. Therefore, the NRC staff concludes that with the proposed changes to the length of the second and subsequent inspection periods, compliance with the SG program requirements in TS 6.7.6 will continue to ensure both adequate inspection scopes and tube integrity. For each inspection period, Paragraph 6.7.6.k.d.2 currently requires that at least 50 percent of the tubes be inspected by the refueling outage nearest to the mid-point of the inspection period and the remaining 50 percent by the refueling outage nearest the end of the inspection period. The NRC staff notes that if there are not an equal number of inspections in the first half and second half of the inspection period, the average minimum sampling requirement may be markedly different for inspections in the first half of the inspection period compared to those in the second half, even when there are uniform intervals between each inspection. For example, a plant in the first (120 EFPM) inspection period with a scheduled 36-month interval (two fuel cycles) between each inspection would currently be required to inspect 50 percent of the tubes by the refueling outage nearest the midpoint of the inspection, which would be the third refueling outage in the period, 6-months before the mid-point. However, since no inspection is scheduled for that outage, then the full 50 percent sample must be performed during the inspection scheduled for the second refueling outage in the period. Two inspections would be scheduled to occur in the second half of the inspection period, at 72 and 1 08 months into the inspection period. Thus, the current sampling requirement could be satisfied by performing a 25 percent sample during each of these inspections or other combinations of sampling (e.g., 10 percent during one and 40 percent in the other) totaling 50 percent. Based on operational evidence, the NRC staff finds there is no basis to require the minimum initial sample size to vary so much from inspection to inspection. The licensee proposes to revise this requirement such that the minimum sample size for a given inspection in a given inspection period is 100 percent, divided by the number of scheduled inspections during that inspection period. For the above example, the proposed change would result in a uniform initial minimum sample size of 33.3 percent for each of the three scheduled inspections during the inspection period. The NRC staff concludes this proposed revision to be an improvement to the existing requirement since it provides a more consistent minimum initial sampling requirement. The proposed changes to Paragraph 6.7.6.k.d.2 include two new sentences addressing the prorating of required tube sample sizes if a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria. For example, new information from another similar plant becomes available indicating the potential for circumferential cracking at a specific location on the tube. Previous degradation assessments had not identified the potential for this type of degradation at this location. Thus, previous inspections of this location had not been performed with a technique capable of detecting circumferential cracks. However, now that the potential for circumferential cracking has been identified at this location, Paragraph 6.7.6.d requires a method of inspection

-8 -to be performed with the objective of detecting circumferential cracks which may be present at this location and that may satisfy the applicable tube plugging criteria. Furthermore, suppose this inspection is performed for the first time during the third of four SG inspections scheduled for one of the inspection periods. Paragraph 6.7.6.d.2 currently does not specify whether this location needs to be 1 00 percent inspected by the end of the inspection period, or whether a prorated approach may be taken. The NRC staff addressed this question in Issue 1 of NRC Regulatory Information Summary (RIS) 2009-04, "Steam Generator Tube Inspection Requirements," dated April 3, 2009 (ADAMS Accession No. ML083470557), as follows: Issue 1: A licensee may identify a new potential degradation mechanism after the first inspection in a sequential period. If this occurs, what are the expectations concerning the scope of examinations for this new potential degradation mechanism for the remainder of the period (e.g., do 100 percent of the tubes have to be inspected by the end of the period or can the sample be prorated for the remaining part of the period)? [NRC Staff Position:] The TS contain requirements that are a mixture of prescriptive and performance-based elements. Paragraph "d" of these requirements indicates that the inspection scope, inspection methods, and inspection intervals shall be sufficient to ensure that SG tube integrity is maintained until the next SG inspection. Paragraph "d" is a performance-based element because it describes the goal of the inspections but does not specify how to achieve the goal. However, paragraph "d.2" is a prescriptive element because it specifies that the licensee must inspect 100 percent of the tubes at specified periods. If an assessment of degradation performed after the first inspection in a sequential period results in a licensee concluding that a new degradation mechanism (not anticipated during the prior inspections in that period) may potentially occur, the scope of inspections in the remaining portion of the period should be sufficient to ensure SG tube integrity for the period between inspections. In addition, to satisfy the prescriptive requirements of paragraph "d.2" that the licensee must inspect 1 00 percent of the tubes within a specified period, a prorated sample for the remaining portion of the period is appropriate for this potentially new degradation mechanism. This prorated sample should be such that if the licensee had implemented it at the beginning of the period, the TS requirement for the 1 00 percent inspection in the entire period (for this degradation mechanism) would have been met. A prorated sample is appropriate because (1) the licensee would have performed the prior inspections in this sequential period consistently with the requirements, and (2) the scope of inspections must be sufficient to ensure that the licensee maintains SG tube integrity for the period between inspections.

-9 -The NRC staff finds that relocation of information in sentences 3 and 4 as described above clarifies the existing requirement consistently with the NRC staff's position from RIS 2009-04 quoted above and is, therefore, acceptable. The proposed fifth sentence in Paragraph 6.7.6.d.2 states, "Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage." Allowing extension of the inspection periods by up to an additional 3 EFPMs potentially impacts the average tube inspection sample size to be implemented during a given inspection in that period. For example, if three SG inspections are scheduled to occur within the nominal 120-EFPM period, then the minimum sample size for each of the four inspections could average as little as 25 percent of the tube population. If a fourth inspection can be included within the period by extending the period by 3 EFPM, then the minimum sample size for each of the five inspections could average as little as 20 percent of the tube population. Since the subsequent period begins at the end of the included SG inspection outage, the proposed change does not impact the required frequency of SG inspection. Required tube inspection sample sizes are also subject to the performance-based requirement in Paragraph 6.7.6.d, which states, in part, that in addition to meeting the requirements of Paragraph 6.7.6.d.1, d.2, and d.3, "the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next scheduled SG inspection." This requirement remains unchanged under the proposal. The NRC staff concludes the proposed fifth sentence involves only a relatively minor relaxation to the existing sampling requirements in Paragraph 6.7.6.d.2. However, the performance-based requirements in 6.7.6.d ensure that adequate inspection sampling will be performed to ensure tube integrity is maintained. The NRC staff concludes that the proposed change is acceptable. Finally, the first sentence of the proposed revision to Paragraph 6.7.6.k.d.2 replaces the last sentence of the current Paragraph 6.7.6.k.d.2. This sentence establishes the minimum allowable SG inspection frequency as at least every 48 EFPM or at least every other refueling outage (whichever results in more frequent inspections). This minimum inspection frequency is unchanged from the current sentence. The NRC staff finds that the wording changes in the sentence are of an editorial and clarifying nature and are not material, such that the current intent of the requirement is unchanged. Thus, the NRC staff concludes the first sentence of proposed paragraph 6.7.6.k.d.2 is acceptable. 3.7 Paragraph 6.7.6.k.d.3 The first sentence of paragraph 6.7.6.k.d.3 currently states: If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). Proposed Change: Revise this sentence as follows: If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each affected and potentially affected SG for the

-10-degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). The proposed change is replacing the words for each SG" with the words for each affected and potentially affected SG." The licensee states that the existing wording can be misinterpreted. The licensee further states that the intention is that those SGs that are affected and those SGs that are potentially affected must be inspected for the degradation mechanism that caused the crack indication. However, some licensees have questioned whether the current reference to "each SG" requires only the SGs that are affected to be inspected for the degradation mechanism. The proposed revision is intended to clarify the intent of the requirement. Assessment: Paragraph 6.7.6.k.d.2 permits SG inspection intervals to extend over multiple fuel cycles for SGs with alloy 600 TT, assuming that such intervals can be implemented while ensuring tube integrity is maintained in accordance with paragraph 6.7.6.k.d. However, corrosion cracks may not become detectable by inspection until the crack depth approaches the tube repair limit. In addition, stress-corrosion cracks may exhibit high growth rates. For these reasons, once cracks have been found in any SG tube, paragraph 6.7.6.k.d.3 restricts the allowable interval to the next scheduled inspection to 24 EFPM or one refueling outage (whichever results in more frequent inspections). The intent of this requirement is that it applies to the affected SG and to any other SG which may be potentially affected by the degradation mechanism that caused the known crack(s). For example, a root cause analysis in response to the initial finding of one or more cracks might reveal that the crack(s) are associated with a manufacturing anomaly which causes locally high residual stress which in turn caused the early initiation of cracks at the affected locations. If it can be established that the extent of condition of the manufacturing anomaly applies only to one SG and not the others, then the NRC staff agrees that only the affected SG needs to be inspected within 24 EFPM or one refueling cycle in accordance with paragraph 6.7.6.k.d.2. The next scheduled inspections of the other SGs will continue to be subject to all other provisions of paragraph 6.7.6.k.d. The NRC staff finds the proposed change to paragraph 6.7.6.k.d.3 acceptable, because it clarifies the intent of the paragraph. 3.8 Specification 6.8.1.7. "Steam Generator Tube Inspection Report" This specification lists items a. through k. to be included in a report which shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.7.6.k, "Steam Generator (SG) Program." Proposed Change: Item b. currently reads: "Active degradation mechanisms found ... " to be revised to read: "Degradation mechanisms found ... " Item e. currently reads: "Number of tubes plugged during the inspection outage for each active degradation mechanism ... " to be revised to read: "Number of tubes plugged during the inspection outage for each degradation mechanism ... "

-11 -Item f. currently reads: "Total number and percentage of tubes plugged to date ... " to be revised to read: "The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator ... " Item h. currently reads: "The effective plugging percentage for all plugging in each SG" and will be deleted. Items i., j., and k., will be re-lettered to correct the sequence of lettering due to the deletion of Item h. Assessment: This proposal would delete the word "Active" in items b. and e. above. Thus, all degradation mechanisms found, whether deemed to be active or not, would now be reportable. The NRC staff finds the proposed change acceptable. The proposal to combine items f. and h. is an editorial change that does not materially change the reporting requirements. The NRC staff finds these changes acceptable. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the New Hampshire and Massachusetts State officials were notified of the proposed issuance of the amendment. The State officials provided no comments. 5.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (78 FR 25316, April 30, 2013). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of the amendment. 6.0 CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) There is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor: R. Grover and J. Lamb Date: October 25, 2013 Mr. Kevin Walsh, Site Vice President c/o Michael Ossing Seabrook Station NextEra Energy Seabrook, LLC P.O. Box 300 Seabrook, NH 0387 4 October 25, 2013 SUBJECT: SEABROOK STATION, UNIT NO.1 -ISSUANCE OF AMENDMENT REGARDING LICENSE AMENDMENT REQUEST 13-02, APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTSF-510, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NO. MF1372) Dear Mr. Walsh: The Commission has issued the enclosed Amendment No. 138 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1. This amendment consists of changes to the facility technical specifications (TSs) in response to your application dated March 27, 2013, as supplemented by letter dated June 25, 2013. The amendment modifies TS requirements regarding steam generator tube inspections and reporting as described in TS Task Force (TSTF)-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," using the Consolidated Line Item Improvement Process. A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-443 Enclosures: 1. Amendment No. 138 to NPF-86 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPLI-2 R/F Sincerely, Ira/ John G. Lamb, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDoriDpr Resource RidsNsirDspCsirb Resource RidsNrrDorllp11-2 Resource RidsNrrPMSeabrook Resource RidsAcrsAcnw_MaiiCTR Resource RidsRgn1 MaiiCenter Resource RidsNrrLAABaxter Resource ADAMS Accession No.: ML13107A016 *via email OFFICE LPL1-2/PM LPL 1-2/LA DSS/STSB/BC OGC-NLO LPL 1-2/BC(A) VRodriguez NAME Jlamb A Baxter* REIIiott BHarris (JHuQhev for) DATE 10/21/2013 10/23/2013 09/25/2013 10/02/2013 10/25/2013 .. Off1c1al Record Copy LPL1-2/PM Jlamb 10/25/2013