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{{#Wiki_filter:}} | {{#Wiki_filter: Mr. John Eltnisky Senior Vice President April 26, 2016 Governance, Projects, and Engineering Duke Energy Carolinas, LLC P.O. Box 1006/EC07H Charlotte, NC 28201-1006 | ||
SUBJECT: CATAWBA NUCLEAR STATION, UNITS 1AND2; MCGUIRE NUCLEAR STATION, UNITS 1AND2; AND OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 -ISSUANCE OF AMENDMENTS REGARDING THE REVISION OF THE STEAM GENERATOR TECHNICAL SPECIFICATIONS TO REFLECT THE ADOPTION OFTSTF-510 (CAC NOS. MF6139, MF6140, MF6141, MF6142, MF6143, MF6144, AND MF6145) | |||
==Dear Mr. Eltnisky:== | |||
By letter dated April 16, 2015, Duke Energy Carolinas, LLC (Duke, the licensee), submitted a license amendment request (LAR) to revise the technical specifications {TSs of Catawba Nuclear Station, Units 1 and 2 (CNS); McGuire Nuclear Station, Units 1 and 2 (MNS); and Oconee Nuclear Station, Units 1, 2, and 3 (ONS). The LAR proposes to incorporate the guidance of Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." The guidance of TSTF-510 revises TS 3.4.20, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of the Improved Standard Technical Specification applicable to CNS and MNS. The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of the Improved Standard Technical Specification applicable to ONS. The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 280 to Renewed Facility Operating License NPF-35 and Amendment No. 276 to Renewed Facility Operating License NPF-52 for Catawba 1 and 2, respectively; Amendment No. 284 to Renewed Facility Operating License NPF-9 and Amendment No. 263 to Renewed Facility Operating License and NPF-17 for McGuire 1 and 2, respectively; and Amendment No. 396 to Renewed Facility Operating License DPR-38, Amendment No. 398 to Renewed Facility Operating Licenses DPR-47, and Amendment No. 397 to Renewed Facility Operating Licenses DPR-55 for Oconee 1, 2 and 3, respectively. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. If you have any questions, please contact me at 301-415-2481 or Ed.Miller@nrc.gov. Docket Nos. 50-413, 50-414, 50-369, 50-370, 50-269, 50-270, and 50-287 | |||
===Enclosures:=== | |||
1. Amendment No. 280 to NPF-35 2. Amendment No. 276 to NPF-52 3. Amendment No. 284 to NPF-9 4. Amendment No. 263 to NPF-17 5. Amendment No. 396 to DPR-38 6. Amendment No. 398 to DPR-47 7. Amendment No. 397 to DPR-55 8. Safety Evaluation cc w/encls: Distribution via Listserv / G. Edward Miller, Project anager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 280 Renewed License No. NPF-35 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-35, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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. Enclosure 1 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 280, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. NPF-35 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. NPF-52 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-52, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 1 O CFR Chapter I; 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. Enclosure 2 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. NPF-52 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERA TING LICENSE Amendment No. 284 Renewed License No. NPF-9 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-9, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 1 O CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 3 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-9 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. NPF-9 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2O1 6 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 Renewed License No. NPF-17 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-17, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 4 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-17 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. NPF-17 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 396 Renewed License No. DPR-38 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-38, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 5 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 396, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. DPR-38 and the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2O1 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 398 Renewed License No. DPR-47 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 6 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 398, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. DPR-47 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2o1 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 397 Renewed License No. DPR-55 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 7 | |||
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 397, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance. | |||
===Attachment:=== | |||
Changes to License No. DPR-55 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION .. .c.e;;-Michae1 T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Dateoflssuance: April 26, 2016 ATTACHMENT TO LICENSE AMENDMENT NO. 280 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses NPF-35, page 4 NPF-52, page 4 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-7a 5.5-8 5.5-9 5.6-6 Licenses NPF-35, page 4 NPF-52, page 4 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.5-14 5.5-15 5.5-16 5.5-17 5.5-18 5.5-19 5.6-6 | |||
-4-(2) Technical Specifications )he Technical Specifications contained in Appendix A, as revised through Amendment No. 280, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 1 O CFR t;i0.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. ( 4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. (5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* . Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision: The licensee may make changes to the approved.fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire . . *The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed. Renewed License No. NPF-35 Amendment No. 280 | |||
-4-(2) TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 1 O CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 1 O CFR 50.59 and otherwise complies with the requirements in that section (4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. (5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provisions: The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. *The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed. Renewed License No. NPF-52 Amendment No. 276 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity v . j LCO 3.4.18 SG tube integrity shall be maintained. SG Tube Integrity 3.4.18 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. ACTIONS Separate Condition entry is allowed for each SG tube. CONDITION A. One or more SG tubes A.1 satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator AND Program. A.2 Catawba Units 1 and 2 REQUIRED ACTION Verify tube integrity of the affected tube{s) is maintained until the next inspection. Plug the affected tube{s) in accordance with the Steam Generator Program. 3.4.18-1 COMPLETION TIME 7 days Prior to entering MODE 4 following the next refueling outage or SG tube inspection {continued) Amendment Nos. 280 and 276 ACTIONS (continued) CONDITION REQUIRED ACTION SG Tube Integrity 3.4.18 COMPLETION TIME 8. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A AND not met. 8.2 Be in MODE 5. 36 hours SG tube integrity not maintained. SURVEILLANCE REQUIREMENTS SR 3.4.18.1 SR 3.4.18.2 SURVEILLANCE FREQUENCY Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. Prior to entering MODE4 following a SG tube inspection Catawba Units 1 and 2 3.4.18-2 Amendment Nos. 280 and 276 5.5 Programs and Manuals (continued) 5.5.8 lnservice Testing Program Programs and Manuals 5.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. 5.5.9 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 (continued) Catawba Units 1 and 2 5.5-6 Amendment Nos. 280 and 276 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 condition of the tubing during a 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. 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 inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown), 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 150 gallons per day through each SG for a total of 600 gallons per day through all SGs. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. {continued) Catawba Units 1 and 2 5.5-7 Amendment Nos. 280 and 276 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 The following SG tube alternate plugging criteria shall be applied as an alternative to the 40% depth based criteria: 1. For Unit 2 only, tubes with service-induced flaws located greater than 14.01 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 14.01 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. For Unit 1, the number and portions of the tubes inspected and method of inspection shall be performed with the objective of detecting flaws of any type (for example, 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 tube-to-tubesheet weld is not part of the tube. For Unit 2, the number and portions of the tubes inspected and method of inspection shall be performed with the objective of detecting flaws of any type (for example, volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, and that may satisfy the applicable tube plugging criteria. In addition to meeting requirements d.1, d.2, d.3, and d.4 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. (continued) Catawba Units 1 and 2 5.5-8 Amendment Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation. 2. For Unit 1, after the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months (EFPM) or at least every third 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, c and d 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 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. Catawba Units 1 and 2 a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. (continued) 5.5-9 Amendment Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) 3. For Unit 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 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 4. For Unit 1, if crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 EFPM or one refueling outage {whichever results in more frequent inspections). For Unit 2, if crack indications are found in any SG tube from 14.01 inches below the top of the {continued) Catawba Units "1 and 2 5.5-10 Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 EFPM 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 crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary to secondary LEAKAGE. Catawba Units 1 and 2 5.5-11 ( continuerl \ Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 5.5.11 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables; b. Identification of the procedures used to measure the values of the critical variables; c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage; d. Procedures for the recording and management of data; e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. Ventilation Filter Testing Program (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with exceptions as noted in the UFSAR. a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below+/- 10%. (continued) Catawba Units 1 and 2 5.5-12 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued) ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Penetration and System Bypass < 1% < 0.05% < 1% < 1% < 1% Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm b. Demonstrate for each of the ESF systems that an inplace test of the carbon adsorber shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below+/- 10%. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Penetration and System Bypass < 1% < 0.05% < 1% < 1% < 1% Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of s 30°C and greater than or equal to the relative humidity specified below. ESF Ventilation System Penetration RH Annulus Ventilation < 4% 95% Control Room Area Ventilation < 0.95% 95% Aux. Bldg. Filtered Exhaust (Note 1) < 4% 95% Containment Purge (non-ESF) < 6% 95% Fuel Bldg. Ventilation < 4% 95% Note 1: The Auxiliary Building Filtered Exhaust System carbon adsorber samples shall be tested at a face velocity of 48 ft/min instead of the 40 ft/min specified in ASTM 03803-1989. 48 ft/min is the nominal limiting velocity the carbon adsorber may be exposed to under post accident conditions as a result of certain postulated failures. The results from this test shall then be corrected to a 2.27 inch bed in accordance with the guidance provided rn ASTM 03803-1989 pnor to comparing them to the Technical Specification criteria. 2.27 inches is the actual bed depth for the filter unit. (continued) Catawba Units 1 and 2 5.5-13 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 5.5.12 Ventilation Filter Testing Program {VFTP) (continued) d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the carbon adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below +/- 10%. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Delta P 8.0 in wg 8.0 in wg 8.0 in wg 8.0 in wg 8.0 in wg Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI N510-1980. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) Fuel Bldg. Ventilation Wattage @ 600 vac 45 +/-. 6.7 kW 25 +/-_2.5 kW 40 +/-. 4.0 kW 120 +/-. 12.0 kW 80 + 8/-17.3 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures". (continued) Catawba Units 1 and 2 5.5-14 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 5.5.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) The program shall include: a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank or connected gas storage tanks and fed into the offgas treatment system is less than the amount that would result in a Deep Dose Equivalent 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations exceeding the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following: a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has: 1. an API gravity or an absolute specific gravity within limits, 2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and (continued) Catawba Units 1 and 2 5.5-15 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 5.5.14 Diesel Fuel Oil Testing Program (continued) 3. a clear and bright appearance with proper color or a water and sediment content within limits; b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c. Total particulate concentration of the fuel oil is:::; 10 mg/I when tested every 31 days. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies. Technical Specifications (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 or Bases that requires NRC approval pursuant to 1 O 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 5.5.14.b.1 or 5.5.14.b.2 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), with approved exemptions. (continued) Catawba Units 1 and 2 5.5-16 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 Safety Function Determination Program (SFDP) This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following: a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and d. Other appropriate limitations and remedial or compensatory actions. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or b. A required system redundant to the system(s) in tum supported by the inoperable supported system is also inoperable; or c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered. Catawba Units 1 and 2 5.5-17 (continued) Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 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 Area Ventilation System (CRAVS), 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 inleakage 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 CRA VS, operating at a makeup flow rate of .:5. 4000 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. e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA 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 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively. Catawba Units 1 and 2 5.5-18 (continued) Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.17 Surveillance Frequency Control Program This Program provides controls for Surveillance Frequencies. The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met. a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. Catawba Units 1 and 2 5.5-19 Amendment Nos. 280 and 276 \ | |||
5.6 Reporting Requirements (continued) 5.6.8 Steam Generator (SG) Tube Inspection Report Repo.rting Requirements 5.6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of the inspection. The report shall include: a. The scope of inspections performed on each SG, b. Degradation mechanisms found, c. Non-destructive 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 in-situ testing, h. For Unit 2, the primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign 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. For Unit 2, the calculated accident leakage rate from the portion of the tubes below 14.01 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident leakage rate from the most limiting accident is less than 3.27 times the maximum primary to secondary LEAKAGE rate, the report shall describe how it was determined, and j. For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided. Catawba Units 1 and 2 5.6-6 Amendment Nos. 280 and 276 ATTACHMENT TO LICENSE AMENDMENT NO. 284 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND LICENSE AMENDMENT NO. 263 RENEWED FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses NPF-9, page 3 NPF-17, page 3 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.6-5 Licenses NPF-9, page 3 NPF-17, page 3 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.6-5 | |||
-3 -4) Pursuant to the Act and 10 CFR Parts 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; (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth iri 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; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. Renewed License No. NPF-9 Amendment No. 284 | |||
-3 -(4) Pursuant to the Act and 10 CFR Parts 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; (5) Pursuant to the Act and 10 CFR Parts, 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as my be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such by product material as may be produced by the Duke Training and Technology Center. C. This renewed operating 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 thereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.263,are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59, and otherwise complies with the requirements in that section. Renewed License No. NPF-17 Amendment No. 263 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained. SG Tube Integrity 3.4.18 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. ACTIONS ---------------------------NOTE-------------------------------Separate Condition entry is allowed for each SG tube. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube( s) is plugging criteria and not maintained until the next plugged in accordance refueling outage or SG with the Steam tube inspection. Generator Program. AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met. 8.2 Be in MODE 5. 36 hours OR SG tube integrity not maintained. McGuire Units 1 and 2 3.4.18-1 Amendment Nos. 284 I 263 SURVEILLANCE REQUIREMENTS SR 3.4.18.1 SR 3.4.18.2 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program. Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. McGuire Units 1 and 2 3.4.18-2 SG Tube Integrity 3.4.18 FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following a SG tube inspection Amendment Nos. 284 I 263 5.5 Programs and Manuals (continued} 5.5.8 lnservice Testing Program Programs and Manuals 5.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: a. Testing frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME Boiler OM Code shall be construed to supersede the requirements of any TS. 5.5.9 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 lnservice 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. (continued) McGuire Units 1 and 2 5.5-6 Amendment Nos. 284 I 263 5.5 Programs and Manuals (continued} 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 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 0.27 gallons per minute total. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. 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 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. ( McGuire Units 1 and 2 5.5-7 Amendment Nos. 284 / 263 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 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 72 effective full power months or at least every third 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, c, and d 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 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 144 effective full power months. This constitutes the first inspection period; b. During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspections period; c. During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d. During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. 3. If crack indications are found in any SG tube, 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. ( cortinut!d) McGuire Units 1 and 2 5.5-8 Amendment Nos. 284 I 263 5.6 Reporting Requirements 5.6.8 Steam Generator Tube Inspection Report Reporting Requirements 5.6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, 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, and g. The results of condition monitoring, including the results of tube pulls and in-situ testing. (continued) McGuire Units 1 and 2 5.6-5 Amendment Nos. 284 I 263 ATTACHMENT TO LICENSE AMENDMENT NO. 396 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND LICENSE AMENDMENT NO. 398 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND LICENSE AMENDMENT NO. 397 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses DPR-38, page 3 DPR-47, page 3* DPR-55, page 3 TSs 3.4.16-1 3.4.16-2 5.0-13 5.0-14 5.0-15 5.0-28 Licenses DPR-38, page 3 DPR-47, page 3 DPR-55, page 3 TSs 3.4.16-1 3.4.16-2 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-28 | |||
-3 -A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 396, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License f\Jo. DPR-38 Amendment No. 396 | |||
-3 -A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 398,are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another, The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-[1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of ariy of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No DPR-47 Amendment No. 398 | |||
-3 -A. . Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 397, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 397 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 Steam Generator (SG) Tube Integrity LCO 3.4.16 SG Tube integrity shall be maintained. SG Tube Integrity 3.4.16 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. ACTIONS Sepa rate Condition entry is allowed for each SG tube. CONDll"ION REQUIRED ACl"ION COMPLEl"ION l"IME A. One or more SG tubes A.1 Verify tube integrity of 7 days satisfying the tube the affected tube( s) is plugging criteria and maintained until the not plugged in next refueling outage accordance with the or SG tube inspection. Steam Generator Program. AND A.2 Plug the affected Prior to entering MODE 4 tube(s) in accordance following the next with the Steam refueling outage or SG Generator Program. tube inspection B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met. B.2 OR Be in MODE 5. 36 hours SG tube integrity not maintained. OCONEE UNITS 1, 2, & 3 3.4.16-1 Amendment Nos. 396, 398, & 397 SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.2 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program. Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. SG Tube Integrity 3.4.16 FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following an SG tube inspection OCONEE UNITS 1, 2, & 3 3.4.16-2 Amendment Nos. 396, 398, & 397 I Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 lnservice Testing Program {continued) 5.5.10 ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. 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. OCONEE UNITS 1, 2, & 3 5.0-13 Amendment Nos. 396, 398, & 397 I 5.5 Programs and Manuals 5.5.10 Steam Generator (SG) Program (continued} Programs and Manuals 5.5 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 150 gallons per day per SG. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. d. Provisions for SG tube inspections. Periodic 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 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 OCONEE UNITS 1, 2, & 3 5.0-14 Amendment Nos. 396, 398, & 397 I 5.5 Programs and Manuals 5.5.10 Steam Generator {SG) Program (continued) Programs and Manuals 5.5 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 72 effective full power months or at least every third 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, c and d 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 detemlination 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 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months This constitutes the fourth and subsequent inspection periods. OCONEE UNITS 1, 2, & 3 5.0-15 Amendment Nos. 396, 398, & 397 I Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Steam Generator (SG) Program (continued) 5.5.11 5.5.12 3. If crack indications are found in any SG tube, 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. Secondary Water Chemistry This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables; b. Identification of the procedures used to measure the values of the critical variabtes; c. Identification of process sampling points; d. Procedures for the recording and management of data; e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. Ventilation Filter Testing Program <VFTP} A program shall be established to implement the following required testing of filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, except that the testing specified at a frequency of 18 months is required at a frequency of 24 months. OCONEE UNITS 1, 2, & 3 5.0-16 Amendment Nos. 396, 398, & 397 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Ventilation Filter Testing Program (VFTP)(continued) 5.5.13 The VFTP is applicable to the Control Room Ventilation System (CRVS) Booster Fan Trains and the Spent Fuel Pool Ventilation System (SFPVS). a. Demonstrate, for the CRVS Booster Fan Trains, that a DOP test of the HEPA filters shows 99.5% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. b. Demonstrate, for the CRVS Booster Fan Trains, that a halogenated hydrocarbon test of the carbon adsorber shows ;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate +/- 10%. c. Demonstrate, for the CRVS Booster Fan Trains and SFPVS, that a laboratory test of a sample of the carbon adsorber shows ;;:: 97 .5% and 90% radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989 (30°C, 95% RH), respectively. d. Demonstrate, for the CRVS Booster Fan Trains, that the pressure drop across the pre-filter is 1 in. of water and the pressure drop across the HEPA filters is 2 in. of water at the system design flow rate +/- 10%. e. Demonstrate, for the SFPVS, that a dioctyl phthalate (DOP) test of the high efficiency particulate air (HEPA) filters shows;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. f. Demonstrate, for the SFPVS, that a halogenated hydrocarbon test of the carbon adsorber shows ;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup tanks and the quantity of radioactivity contained in waste gas holdup tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined. The liquid radwaste quantities shall be determined by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. OCONEE UNITS 1, 2, & 3 5.0-17 Amendment Nos. 396, 398, & 397 I 5.6 Reporting Requirements 5.6.7 Tendon Surveillance Report Reporting Requirements 5.6 Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. 5.6.8 Steam Generator Tube Inspection Report 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 5.5.10, 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, and, g. The results of condition monitoring, including the results of tube pulls and in-situ testing. OCONEE UNITS 1, 2, & 3 5.0-28 Amendment Nos. 396, 398, & 397 I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 280 TO RENEWED FACILITY OPERATING LICENSE NPF-35; AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NPF-52; AMENDMENT NO. 284 TO RENEWED FACILITY OPERATING LICENSE NPF-9; AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NPF-17; AMENDMENT NO. 396 TO RENEWED FACILITY OPERATING LICENSE DPR-38; AMENDMENT NO. 398 TO RENEWED FACILITY OPERATING LICENSE DPR-47; AND AMENDMENT NO. 397 TO RENEWED FACILITY OPERATING LICENSE DPR-55; DUKE ENERGY CAROLINAS. LLC CATAWBA NUCLEAR STATION. UNITS 1AND2 DOCKET NOS. 50-413 AND 50-414 MCGUIRE NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 OCONEE NUCLEAR STATION. UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270. AND 50-287 1.0 INTRODUCTION By letter dated April 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15119A224), Duke Energy Carolinas, LLC (the licensee) submitted a license amendment request (LAR) to revise the technical specifications (TSs) of 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 LAR proposes to incorporate the guidance of Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (ADAMS Accession No. ML 110610350). The guidance of TSTF-510 revises TS 3.4.20, "Steam Generator (SG) Enclosure 8 | |||
-2 -Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of NUREG-1431, Revision 4, "Standard Technical Specifications [STS] -Westinghouse Plants" (ADAMS Accession No. ML 12100A222), applicable to CNS and MNS. The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of NUREG-1430, Revision 4, "Standard Technical Specifications -Babcock and Wilcox Plants" (ADAMS Accession No. ML 12100A177) applicable to ONS. The specific changes concern SG inspection periods and address applicable administrative changes and clarifications. The licensee stated that the LAR is consistent with the Notice of Availability of TSTF-510, Revision 2, announced in the Federal Register on October 27, 2011 (76 FR 66763) as part of the consolidated line item improvement process. The current 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 performance-based, approach for ensuring the integrity of the SG tubes is maintained. The performance-based requirements were supplemented by prescriptive requirements relating to tube 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-510, Revision 2, reflect licensees' early implementation experience with respect to 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. | |||
-3 -Title 10 of the Code of Federal Regulations (10 CFR) establishes the requirements with respect to the integrity of SG tubing. Specifically, the General Design Criteria (GDC) in Appendix A to 10 CFR Part 50 state that the RCPB: *shall have "an extremely low probability of abnormal leakage ... and gross rupture" (GDC 14), *"shall be designed with sufficient margin" (GDC 15 and 31), | |||
* shall be of "the highest quality standards possible" (GDC 30), and *shall be designed to permit "periodic inspection and testing ... to assess ... structural and leak tight integrity" (GDC 32). The three ONS units were licensed prior to the 1971 publication of Appendix A, "General Design Criteria for Nuclear Power Plants,'' to 10 CFR Part 50. As such, ONS is not licensed to the GDC in Appendix A, but rather to the U.S. Atomic Energy Commission GDC that were contained in a proposed rulemaking published in the Federal Register of July 11, 1967 (ADAMS Accession No. ML043310029). Section 3.1 of the ONS Final Safety Analysis Report lists the GDC to which the plant was licensed. The ONS GDC addressing the RCPB are GDC 9, "Reactor Coolant Pressure Boundary"; GDC 33, "Reactor Coolant Pressure Boundary Capability"; GDC 34, "Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention"; and GDC 36, "Reactor Coolant Pressure Boundary Surveillance." These GDC are similar to GDC 14, 15, 31, and 32 in Appendix A of 10 CFR Part 50. The operating licenses for CNS and MNS were issued after 1971, and these units meet the GDC in Appendix A of 10 CFR Part 50. The regulations in 1 O CFR 50.55a specify that RCPB components must meet the requirements for Class 1 components in Section Ill of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Section 50.55a further requires, in part, that throughout the service life of a PWR facility, ASME Code Class 1 components meet the requirements, except design and access provisions and pre-service examination requirements, in Section XI, "Rules for lnservice Inspection (ISi) of Nuclear Power Plant Components," of the ASME Code, to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code. In the 1970s, ASME Code Section XI requirements pertaining to ISi of SG tubing were augmented by additional SG tube surveillance requirements (SRs) in the TSs. The regulation at 10 CFR 50.55a, paragraph (b)(2)(iii), states, "if the [TSs] ... include [SRs] for [SGs] different than those in Article IWB-2000, the [ISi] program for [SG] tubing is governed by the requirements in the [TSs]." As part of the plant's licensing basis, applicants for PWR licenses are required to analyze the consequences of postulated design-basis accidents such as an SG tube rupture and main steamline break. These analyses consider the primary-to-secondary leakage that may occur during these events and must show that the offsite radiological consequences do not exceed the applicable limits of the 10 CFR Part 100.11 guidelines for offsite doses (or 10 CFR 50.67, as appropriate), GDC-19 criteria for control room operator doses, or some fraction thereof as appropriate to the accident, or the NRG-approved licensing basis. | |||
-4 -The regulation at 10 CFR 50.36, "Technical specifications," establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. For CNS and MNS, the LCOs (and accompanying action statements) and the SRs in the STS that are relevant to SG tube integrity are in Specification 3.4.13, "Reactor Coolant System Operational Leakage," and Specification 3.4.20 (SR 3.4.20.2), "Steam Generator (SG) Tube Integrity." For ONS, the LCOs (and accompanying action statements) and the SRs in the STS that are relevant to SG tube integrity are in Specification 3.4.13, "Reactor Coolant System Operational Leakage," and Specification 3.4.17 (SR 3.4.17.2), "Steam Generator (SG) Tube Integrity." The SRs in the "Steam Generator (SG) Tube Integrity" specification reference the SG Program, which is defined in the STS administrative controls section. The regulation at 10 CFR 50.36(c)(5) defines 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 STS. The SG Program is defined in Specification 5.5.9 of the STS for CNS, MNS, and ONS, while the reporting requirements relating to implementation of the SG Program are in Specification 5.6.7 of the STS for CNS, MNS, and ONS. STS 5.5.9, "Steam Generator (SG) Program," for CNS, MNS, and ONS, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Tube integrity is maintained by meeting the performance criteria specified in STS 5.5.9 for structural and leakage integrity, consistent with the plant design and licensing bases. STS 5.5.9.a requires that a condition monitoring assessment be performed during each outage, during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met. STS 5.5.9.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 tube-to-tubesheet weld at the tube outlet, and (2) may satisfy the applicable tube repair criteria. The applicable tube repair criteria specified in STS 5.5.9.c are that tubes found during ISi to contain flaws with a depth equal to, or exceeding, 40 percent of the nominal wall thickness shall be plugged, unless the tubes are permitted to remain in service through application of the alternate repair criteria provided in STS 5.5.9.c.1. | |||
-5 -3.0 TECHNICAL EVALUATION The changes in TSTF-510, Revision 2, reflect licensees' early implementation experience with their current TSs. The changes in TSTF-510, Revision 2, are 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's licensing basis will be maintained between SG inspections. The NRC staff approved TSTF-510, Revision 2, for use with the consolidated line item process on October 19, 2011 (ADAMS Accession No. ML 112101604). Other than the variations or deviations discussed below, the licensee is not proposing any variations or deviations from the TS changes described in TSTF-510, Revision 2. As a result, the NRC staff's evaluation is focused on these differences, since the other changes were previously evaluated in the model safety evaluation (ADAMS Accession No. ML 112101513). Additionally, the model safety evaluation contains sections for the various SG tube materials. The sections pertaining to SGs with alloy 600 thermally treated tubes apply to CNS, Unit 2, while the section pertaining to SGs with alloy 690 thermally treated tubes apply to CNS, Unit 1; MNS, Units 1 and 2; and ONS Units 1, 2, and 3. 3.1 Administrative Changes and Variations | |||
* The CNS, MNS, and ONS TSs utilize different numbering than the STS on which TSTF-510, Revision 2, was based. For CNS and MNS, the "Steam Generator Tube Integrity" TS is numbered 3.4.18 rather than 3.4.20, and the "Steam Generator Tube Inspection Report" TS is numbered 5.6.8 rather than 5.6.7. For ONS, the "Steam Generator (SG) Program" TS is numbered 5.5.10 rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.16 rather than 3.4.17, and the "Steam Generator Tube Inspection Report" TS is numbered 5.6.8 rather than 5.6.7. | |||
* An NRC letter dated June 17, 2013 (ADAMS Accession No. ML 13120A541), clarified that if LARs proposing to implement TSTF-510, Revision 2, corrected an administrative inconsistency in paragraph 5.5.9.d.2 of the SG Program, they would not result in removal of submitted LARs from the consolidated line item improvement process. Because CNS, MNS, and ONS do not have any approved tube repair methods, this LAR fixes the administrative inconsistency in paragraphs 5.5.9.d.2 and 5.5.9.d.3 (for CNS), 5.5.9.d.2 (for MNS), and 5.5.1 O.d.2 (for ONS), by replacing "tube repair criteria" with "tube plugging criteria," per the NRC letter dated June 17, 2013. | |||
* The acronym EFPM (effective full power months) is not defined in TSTF-510, Revision 2, but is defined and used in various subsections of CNS TS 5.5.9.d.2 of the proposed LAR. | |||
-6 -The differences noted above are administrative and do not affect the applicability of TSTF-510, Revision 2, to the TSs of CNS, MNS, and ONS. As a result, the NRC staff finds the differences between what was approved for TSTF-510, Revision 2, and what is being proposed acceptable. Thus, the NRC concludes that the requirements of 10 CFR 50.36 are met and the proposed amendments are acceptable. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the North Carolina and South Carolina State officials were notified of the proposed issuance of the amendments. The State officials had no comments. 5.0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on June 23, 2015 (80 FR 35981). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments. 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: A. Johnson Date: April 26, 2016 If you have any questions, please contact me at 301-415-2481 or Ed.Miller@nrc.gov. Docket Nos. 50-413, 50-414, 50-369, 50-370, 50-269, 50-270, and 50-287 | |||
===Enclosures:=== | |||
1. Amendment No. 280 to NPF-35 2. Amendment No. 276 to NPF-52 3. Amendment No. 284 to NPF-9 4. Amendment No. 263 to NPF-17 5. Amendment No. 396 to DPR-38 6. Amendment No. 398 to DPR-47 7. Amendment No. 397 to DPR-55 8. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPL2-1 R/F RidsACRS_MailCTR Resource RidsNrrDeEsgb Resource RidsNrrDirsStsb Resource RidsNrrDorlDpr Resource RidsNrrDorllpl2-1 Resource RidsNrrPMCatawba Resource RidsNrrPMMcGuire Resource RidsNrrPMOconee Resource RidsNrrLALRonewicz Resource RidsRgn2MailCenter Resource AJohnson, NRR ADAMS Accession No.: ML16075A301 OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA NAME MOrenak LRonewicz DATE 03/23/16 04/13/16 OFFICE OGC-NLO DORL/LPL2-1 /BC NAME SF owler MMarkley DATE 04/8/16 04/26/16 | |||
Sincerely,/RA/ G. Edward Miller, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation *SE dated DE/ESGB/BC* DSS/STSB/BC GKulsea RElliott (SAnderson for) 03/04/16 03/30/16 DORL/LPL2-1 /PM GMiller 04/26/16 OFFICIAL RECORD COPY}} | |||
Revision as of 21:06, 19 March 2018
| ML16075A301 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Mcguire, Catawba, McGuire  |
| Issue date: | 04/26/2016 |
| From: | Miller G E Plant Licensing Branch II |
| To: | Eltnisky J Duke Energy Carolinas |
| Miller G E, NRR/DORL/LPLII-1 | |
| References | |
| CAC MF6139, CAC MF6140, CAC MF6141, CAC MF6142, CAC MF6143, CAC MF6144, CAC MF6145 | |
| Download: ML16075A301 (64) | |
Text
Mr. John Eltnisky Senior Vice President April 26, 2016 Governance, Projects, and Engineering Duke Energy Carolinas, LLC P.O. Box 1006/EC07H Charlotte, NC 28201-1006
SUBJECT: CATAWBA NUCLEAR STATION, UNITS 1AND2; MCGUIRE NUCLEAR STATION, UNITS 1AND2; AND OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 -ISSUANCE OF AMENDMENTS REGARDING THE REVISION OF THE STEAM GENERATOR TECHNICAL SPECIFICATIONS TO REFLECT THE ADOPTION OFTSTF-510 (CAC NOS. MF6139, MF6140, MF6141, MF6142, MF6143, MF6144, AND MF6145)
Dear Mr. Eltnisky:
By letter dated April 16, 2015, Duke Energy Carolinas, LLC (Duke, the licensee), submitted a license amendment request (LAR) to revise the technical specifications {TSs of Catawba Nuclear Station, Units 1 and 2 (CNS); McGuire Nuclear Station, Units 1 and 2 (MNS); and Oconee Nuclear Station, Units 1, 2, and 3 (ONS). The LAR proposes to incorporate the guidance of Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." The guidance of TSTF-510 revises TS 3.4.20, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of the Improved Standard Technical Specification applicable to CNS and MNS. The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of the Improved Standard Technical Specification applicable to ONS. The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 280 to Renewed Facility Operating License NPF-35 and Amendment No. 276 to Renewed Facility Operating License NPF-52 for Catawba 1 and 2, respectively; Amendment No. 284 to Renewed Facility Operating License NPF-9 and Amendment No. 263 to Renewed Facility Operating License and NPF-17 for McGuire 1 and 2, respectively; and Amendment No. 396 to Renewed Facility Operating License DPR-38, Amendment No. 398 to Renewed Facility Operating Licenses DPR-47, and Amendment No. 397 to Renewed Facility Operating Licenses DPR-55 for Oconee 1, 2 and 3, respectively. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. If you have any questions, please contact me at 301-415-2481 or Ed.Miller@nrc.gov. Docket Nos. 50-413, 50-414, 50-369, 50-370, 50-269, 50-270, and 50-287
Enclosures:
1. Amendment No. 280 to NPF-35 2. Amendment No. 276 to NPF-52 3. Amendment No. 284 to NPF-9 4. Amendment No. 263 to NPF-17 5. Amendment No. 396 to DPR-38 6. Amendment No. 398 to DPR-47 7. Amendment No. 397 to DPR-55 8. Safety Evaluation cc w/encls: Distribution via Listserv / G. Edward Miller, Project anager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 280 Renewed License No. NPF-35 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-35, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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. Enclosure 1
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 280, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. NPF-35 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. NPF-52 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-52, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 1 O CFR Chapter I; 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. Enclosure 2
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. NPF-52 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERA TING LICENSE Amendment No. 284 Renewed License No. NPF-9 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-9, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 1 O CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 3
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-9 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. NPF-9 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2O1 6 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 Renewed License No. NPF-17 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-17, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 4
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-17 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. NPF-17 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: April 26, 201 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 396 Renewed License No. DPR-38 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-38, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 5
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 396, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. DPR-38 and the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2O1 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 398 Renewed License No. DPR-47 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 6
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 398, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. DPR-47 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: Apr i 1 2 6, 2o1 6 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 397 Renewed License No. DPR-55 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), dated April 16, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 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 set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 7
-2 -2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 397, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
Attachment:
Changes to License No. DPR-55 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION .. .c.e;;-Michae1 T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Dateoflssuance: April 26, 2016 ATTACHMENT TO LICENSE AMENDMENT NO. 280 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses NPF-35, page 4 NPF-52, page 4 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-7a 5.5-8 5.5-9 5.6-6 Licenses NPF-35, page 4 NPF-52, page 4 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.5-14 5.5-15 5.5-16 5.5-17 5.5-18 5.5-19 5.6-6
-4-(2) Technical Specifications )he Technical Specifications contained in Appendix A, as revised through Amendment No. 280, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 1 O CFR t;i0.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. ( 4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. (5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* . Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision: The licensee may make changes to the approved.fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire . . *The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed. Renewed License No. NPF-35 Amendment No. 280
-4-(2) TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 1 O CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 1 O CFR 50.59 and otherwise complies with the requirements in that section (4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. (5) Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provisions: The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. *The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed. Renewed License No. NPF-52 Amendment No. 276 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity v . j LCO 3.4.18 SG tube integrity shall be maintained. SG Tube Integrity 3.4.18 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. ACTIONS Separate Condition entry is allowed for each SG tube. CONDITION A. One or more SG tubes A.1 satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator AND Program. A.2 Catawba Units 1 and 2 REQUIRED ACTION Verify tube integrity of the affected tube{s) is maintained until the next inspection. Plug the affected tube{s) in accordance with the Steam Generator Program. 3.4.18-1 COMPLETION TIME 7 days Prior to entering MODE 4 following the next refueling outage or SG tube inspection {continued) Amendment Nos. 280 and 276 ACTIONS (continued) CONDITION REQUIRED ACTION SG Tube Integrity 3.4.18 COMPLETION TIME 8. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met. 8.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SG tube integrity not maintained. SURVEILLANCE REQUIREMENTS SR 3.4.18.1 SR 3.4.18.2 SURVEILLANCE FREQUENCY Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. Prior to entering MODE4 following a SG tube inspection Catawba Units 1 and 2 3.4.18-2 Amendment Nos. 280 and 276 5.5 Programs and Manuals (continued) 5.5.8 lnservice Testing Program Programs and Manuals 5.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. 5.5.9 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 (continued) Catawba Units 1 and 2 5.5-6 Amendment Nos. 280 and 276 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 condition of the tubing during a 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. 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 inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown), 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 150 gallons per day through each SG for a total of 600 gallons per day through all SGs. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. {continued) Catawba Units 1 and 2 5.5-7 Amendment Nos. 280 and 276 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 The following SG tube alternate plugging criteria shall be applied as an alternative to the 40% depth based criteria: 1. For Unit 2 only, tubes with service-induced flaws located greater than 14.01 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 14.01 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. For Unit 1, the number and portions of the tubes inspected and method of inspection shall be performed with the objective of detecting flaws of any type (for example, 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 tube-to-tubesheet weld is not part of the tube. For Unit 2, the number and portions of the tubes inspected and method of inspection shall be performed with the objective of detecting flaws of any type (for example, volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, and that may satisfy the applicable tube plugging criteria. In addition to meeting requirements d.1, d.2, d.3, and d.4 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. (continued) Catawba Units 1 and 2 5.5-8 Amendment Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation. 2. For Unit 1, after the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months (EFPM) or at least every third 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, c and d 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 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. Catawba Units 1 and 2 a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. (continued) 5.5-9 Amendment Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) 3. For Unit 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 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 4. For Unit 1, if crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 EFPM or one refueling outage {whichever results in more frequent inspections). For Unit 2, if crack indications are found in any SG tube from 14.01 inches below the top of the {continued) Catawba Units "1 and 2 5.5-10 Nos. 280 and 276 5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Steam Generator (SG) Program (continued) tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 EFPM 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 crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary to secondary LEAKAGE. Catawba Units 1 and 2 5.5-11 ( continuerl \ Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 5.5.11 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables; b. Identification of the procedures used to measure the values of the critical variables; c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage; d. Procedures for the recording and management of data; e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. Ventilation Filter Testing Program (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with exceptions as noted in the UFSAR. a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below+/- 10%. (continued) Catawba Units 1 and 2 5.5-12 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued) ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Penetration and System Bypass < 1% < 0.05% < 1% < 1% < 1% Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm b. Demonstrate for each of the ESF systems that an inplace test of the carbon adsorber shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below+/- 10%. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Penetration and System Bypass < 1% < 0.05% < 1% < 1% < 1% Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of s 30°C and greater than or equal to the relative humidity specified below. ESF Ventilation System Penetration RH Annulus Ventilation < 4% 95% Control Room Area Ventilation < 0.95% 95% Aux. Bldg. Filtered Exhaust (Note 1) < 4% 95% Containment Purge (non-ESF) < 6% 95% Fuel Bldg. Ventilation < 4% 95% Note 1: The Auxiliary Building Filtered Exhaust System carbon adsorber samples shall be tested at a face velocity of 48 ft/min instead of the 40 ft/min specified in ASTM 03803-1989. 48 ft/min is the nominal limiting velocity the carbon adsorber may be exposed to under post accident conditions as a result of certain postulated failures. The results from this test shall then be corrected to a 2.27 inch bed in accordance with the guidance provided rn ASTM 03803-1989 pnor to comparing them to the Technical Specification criteria. 2.27 inches is the actual bed depth for the filter unit. (continued) Catawba Units 1 and 2 5.5-13 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 5.5.12 Ventilation Filter Testing Program {VFTP) (continued) d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the carbon adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the flowrate specified below +/- 10%. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) (2 fans) Fuel Bldg. Ventilation Delta P 8.0 in wg 8.0 in wg 8.0 in wg 8.0 in wg 8.0 in wg Flowrate 9000 cfm 6000 cfm 30,000 cfm 25,000 cfm 16,565 cfm e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI N510-1980. ESF Ventilation System Annulus Ventilation Control Room Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF) Fuel Bldg. Ventilation Wattage @ 600 vac 45 +/-. 6.7 kW 25 +/-_2.5 kW 40 +/-. 4.0 kW 120 +/-. 12.0 kW 80 + 8/-17.3 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures". (continued) Catawba Units 1 and 2 5.5-14 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 5.5.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) The program shall include: a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank or connected gas storage tanks and fed into the offgas treatment system is less than the amount that would result in a Deep Dose Equivalent 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations exceeding the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following: a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has: 1. an API gravity or an absolute specific gravity within limits, 2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and (continued) Catawba Units 1 and 2 5.5-15 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 5.5.14 Diesel Fuel Oil Testing Program (continued) 3. a clear and bright appearance with proper color or a water and sediment content within limits; b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c. Total particulate concentration of the fuel oil is:::; 10 mg/I when tested every 31 days. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies. Technical Specifications (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 or Bases that requires NRC approval pursuant to 1 O 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 5.5.14.b.1 or 5.5.14.b.2 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), with approved exemptions. (continued) Catawba Units 1 and 2 5.5-16 Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 Safety Function Determination Program (SFDP) This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following: a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and d. Other appropriate limitations and remedial or compensatory actions. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or b. A required system redundant to the system(s) in tum supported by the inoperable supported system is also inoperable; or c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered. Catawba Units 1 and 2 5.5-17 (continued) Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 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 Area Ventilation System (CRAVS), 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 inleakage 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 CRA VS, operating at a makeup flow rate of .:5. 4000 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. e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA 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 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively. Catawba Units 1 and 2 5.5-18 (continued) Amendment Nos. 280 and 276 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.17 Surveillance Frequency Control Program This Program provides controls for Surveillance Frequencies. The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met. a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. Catawba Units 1 and 2 5.5-19 Amendment Nos. 280 and 276 \
5.6 Reporting Requirements (continued) 5.6.8 Steam Generator (SG) Tube Inspection Report Repo.rting Requirements 5.6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of the inspection. The report shall include: a. The scope of inspections performed on each SG, b. Degradation mechanisms found, c. Non-destructive 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 in-situ testing, h. For Unit 2, the primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign 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. For Unit 2, the calculated accident leakage rate from the portion of the tubes below 14.01 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident leakage rate from the most limiting accident is less than 3.27 times the maximum primary to secondary LEAKAGE rate, the report shall describe how it was determined, and j. For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided. Catawba Units 1 and 2 5.6-6 Amendment Nos. 280 and 276 ATTACHMENT TO LICENSE AMENDMENT NO. 284 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND LICENSE AMENDMENT NO. 263 RENEWED FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses NPF-9, page 3 NPF-17, page 3 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.6-5 Licenses NPF-9, page 3 NPF-17, page 3 TSs 3.4.18-1 3.4.18-2 5.5-6 5.5-7 5.5-8 5.6-5
-3 -4) Pursuant to the Act and 10 CFR Parts 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; (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth iri 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; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. Renewed License No. NPF-9 Amendment No. 284
-3 -(4) Pursuant to the Act and 10 CFR Parts 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; (5) Pursuant to the Act and 10 CFR Parts, 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as my be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such by product material as may be produced by the Duke Training and Technology Center. C. This renewed operating 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 thereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.263,are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59, and otherwise complies with the requirements in that section. Renewed License No. NPF-17 Amendment No. 263 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained. SG Tube Integrity 3.4.18 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. ACTIONS ---------------------------NOTE-------------------------------Separate Condition entry is allowed for each SG tube. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube( s) is plugging criteria and not maintained until the next plugged in accordance refueling outage or SG with the Steam tube inspection. Generator Program. AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met. 8.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained. McGuire Units 1 and 2 3.4.18-1 Amendment Nos. 284 I 263 SURVEILLANCE REQUIREMENTS SR 3.4.18.1 SR 3.4.18.2 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program. Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. McGuire Units 1 and 2 3.4.18-2 SG Tube Integrity 3.4.18 FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following a SG tube inspection Amendment Nos. 284 I 263 5.5 Programs and Manuals (continued} 5.5.8 lnservice Testing Program Programs and Manuals 5.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following: a. Testing frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME Boiler OM Code shall be construed to supersede the requirements of any TS. 5.5.9 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 lnservice 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. (continued) McGuire Units 1 and 2 5.5-6 Amendment Nos. 284 I 263 5.5 Programs and Manuals (continued} 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 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 0.27 gallons per minute total. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. 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 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. ( McGuire Units 1 and 2 5.5-7 Amendment Nos. 284 / 263 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program (continued) Programs and Manuals 5.5 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 72 effective full power months or at least every third 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, c, and d 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 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 144 effective full power months. This constitutes the first inspection period; b. During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspections period; c. During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d. During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. 3. If crack indications are found in any SG tube, 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. ( cortinut!d) McGuire Units 1 and 2 5.5-8 Amendment Nos. 284 I 263 5.6 Reporting Requirements 5.6.8 Steam Generator Tube Inspection Report Reporting Requirements 5.6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, 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, and g. The results of condition monitoring, including the results of tube pulls and in-situ testing. (continued) McGuire Units 1 and 2 5.6-5 Amendment Nos. 284 I 263 ATTACHMENT TO LICENSE AMENDMENT NO. 396 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND LICENSE AMENDMENT NO. 398 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND LICENSE AMENDMENT NO. 397 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses DPR-38, page 3 DPR-47, page 3* DPR-55, page 3 TSs 3.4.16-1 3.4.16-2 5.0-13 5.0-14 5.0-15 5.0-28 Licenses DPR-38, page 3 DPR-47, page 3 DPR-55, page 3 TSs 3.4.16-1 3.4.16-2 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-28
-3 -A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 396, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License f\Jo. DPR-38 Amendment No. 396
-3 -A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 398,are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another, The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-[1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of ariy of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No DPR-47 Amendment No. 398
-3 -A. . Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 397, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 397 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 Steam Generator (SG) Tube Integrity LCO 3.4.16 SG Tube integrity shall be maintained. SG Tube Integrity 3.4.16 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. ACTIONS Sepa rate Condition entry is allowed for each SG tube. CONDll"ION REQUIRED ACl"ION COMPLEl"ION l"IME A. One or more SG tubes A.1 Verify tube integrity of 7 days satisfying the tube the affected tube( s) is plugging criteria and maintained until the not plugged in next refueling outage accordance with the or SG tube inspection. Steam Generator Program. AND A.2 Plug the affected Prior to entering MODE 4 tube(s) in accordance following the next with the Steam refueling outage or SG Generator Program. tube inspection B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met. B.2 OR Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SG tube integrity not maintained. OCONEE UNITS 1, 2, & 3 3.4.16-1 Amendment Nos. 396, 398, & 397 SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.2 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program. Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program. SG Tube Integrity 3.4.16 FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following an SG tube inspection OCONEE UNITS 1, 2, & 3 3.4.16-2 Amendment Nos. 396, 398, & 397 I Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 lnservice Testing Program {continued) 5.5.10 ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities; c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. 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. OCONEE UNITS 1, 2, & 3 5.0-13 Amendment Nos. 396, 398, & 397 I 5.5 Programs and Manuals 5.5.10 Steam Generator (SG) Program (continued} Programs and Manuals 5.5 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 150 gallons per day per SG. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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. d. Provisions for SG tube inspections. Periodic 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 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 OCONEE UNITS 1, 2, & 3 5.0-14 Amendment Nos. 396, 398, & 397 I 5.5 Programs and Manuals 5.5.10 Steam Generator {SG) Program (continued) Programs and Manuals 5.5 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 72 effective full power months or at least every third 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, c and d 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 detemlination 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 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months This constitutes the fourth and subsequent inspection periods. OCONEE UNITS 1, 2, & 3 5.0-15 Amendment Nos. 396, 398, & 397 I Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Steam Generator (SG) Program (continued) 5.5.11 5.5.12 3. If crack indications are found in any SG tube, 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. Secondary Water Chemistry This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables; b. Identification of the procedures used to measure the values of the critical variabtes; c. Identification of process sampling points; d. Procedures for the recording and management of data; e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. Ventilation Filter Testing Program <VFTP} A program shall be established to implement the following required testing of filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, except that the testing specified at a frequency of 18 months is required at a frequency of 24 months. OCONEE UNITS 1, 2, & 3 5.0-16 Amendment Nos. 396, 398, & 397 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Ventilation Filter Testing Program (VFTP)(continued) 5.5.13 The VFTP is applicable to the Control Room Ventilation System (CRVS) Booster Fan Trains and the Spent Fuel Pool Ventilation System (SFPVS). a. Demonstrate, for the CRVS Booster Fan Trains, that a DOP test of the HEPA filters shows 99.5% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. b. Demonstrate, for the CRVS Booster Fan Trains, that a halogenated hydrocarbon test of the carbon adsorber shows ;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate +/- 10%. c. Demonstrate, for the CRVS Booster Fan Trains and SFPVS, that a laboratory test of a sample of the carbon adsorber shows ;;:: 97 .5% and 90% radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989 (30°C, 95% RH), respectively. d. Demonstrate, for the CRVS Booster Fan Trains, that the pressure drop across the pre-filter is 1 in. of water and the pressure drop across the HEPA filters is 2 in. of water at the system design flow rate +/- 10%. e. Demonstrate, for the SFPVS, that a dioctyl phthalate (DOP) test of the high efficiency particulate air (HEPA) filters shows;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. f. Demonstrate, for the SFPVS, that a halogenated hydrocarbon test of the carbon adsorber shows ;;:: 99% removal when tested in accordance with ANSI N510-1975 at the system design flow rate+/- 10%. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup tanks and the quantity of radioactivity contained in waste gas holdup tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined. The liquid radwaste quantities shall be determined by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. OCONEE UNITS 1, 2, & 3 5.0-17 Amendment Nos. 396, 398, & 397 I 5.6 Reporting Requirements 5.6.7 Tendon Surveillance Report Reporting Requirements 5.6 Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. 5.6.8 Steam Generator Tube Inspection Report 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 5.5.10, 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, and, g. The results of condition monitoring, including the results of tube pulls and in-situ testing. OCONEE UNITS 1, 2, & 3 5.0-28 Amendment Nos. 396, 398, & 397 I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 280 TO RENEWED FACILITY OPERATING LICENSE NPF-35; AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NPF-52; AMENDMENT NO. 284 TO RENEWED FACILITY OPERATING LICENSE NPF-9; AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NPF-17; AMENDMENT NO. 396 TO RENEWED FACILITY OPERATING LICENSE DPR-38; AMENDMENT NO. 398 TO RENEWED FACILITY OPERATING LICENSE DPR-47; AND AMENDMENT NO. 397 TO RENEWED FACILITY OPERATING LICENSE DPR-55; DUKE ENERGY CAROLINAS. LLC CATAWBA NUCLEAR STATION. UNITS 1AND2 DOCKET NOS. 50-413 AND 50-414 MCGUIRE NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 OCONEE NUCLEAR STATION. UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270. AND 50-287 1.0 INTRODUCTION By letter dated April 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15119A224), Duke Energy Carolinas, LLC (the licensee) submitted a license amendment request (LAR) to revise the technical specifications (TSs) of 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 LAR proposes to incorporate the guidance of Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (ADAMS Accession No. ML 110610350). The guidance of TSTF-510 revises TS 3.4.20, "Steam Generator (SG) Enclosure 8
-2 -Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of NUREG-1431, Revision 4, "Standard Technical Specifications [STS] -Westinghouse Plants" (ADAMS Accession No. ML 12100A222), applicable to CNS and MNS. The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG) Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report," of NUREG-1430, Revision 4, "Standard Technical Specifications -Babcock and Wilcox Plants" (ADAMS Accession No. ML 12100A177) applicable to ONS. The specific changes concern SG inspection periods and address applicable administrative changes and clarifications. The licensee stated that the LAR is consistent with the Notice of Availability of TSTF-510, Revision 2, announced in the Federal Register on October 27, 2011 (76 FR 66763) as part of the consolidated line item improvement process. The current 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 performance-based, approach for ensuring the integrity of the SG tubes is maintained. The performance-based requirements were supplemented by prescriptive requirements relating to tube 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-510, Revision 2, reflect licensees' early implementation experience with respect to 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.
-3 -Title 10 of the Code of Federal Regulations (10 CFR) establishes the requirements with respect to the integrity of SG tubing. Specifically, the General Design Criteria (GDC) in Appendix A to 10 CFR Part 50 state that the RCPB: *shall have "an extremely low probability of abnormal leakage ... and gross rupture" (GDC 14), *"shall be designed with sufficient margin" (GDC 15 and 31),
- shall be of "the highest quality standards possible" (GDC 30), and *shall be designed to permit "periodic inspection and testing ... to assess ... structural and leak tight integrity" (GDC 32). The three ONS units were licensed prior to the 1971 publication of Appendix A, "General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. As such, ONS is not licensed to the GDC in Appendix A, but rather to the U.S. Atomic Energy Commission GDC that were contained in a proposed rulemaking published in the Federal Register of July 11, 1967 (ADAMS Accession No. ML043310029). Section 3.1 of the ONS Final Safety Analysis Report lists the GDC to which the plant was licensed. The ONS GDC addressing the RCPB are GDC 9, "Reactor Coolant Pressure Boundary"; GDC 33, "Reactor Coolant Pressure Boundary Capability"; GDC 34, "Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention"; and GDC 36, "Reactor Coolant Pressure Boundary Surveillance." These GDC are similar to GDC 14, 15, 31, and 32 in Appendix A of 10 CFR Part 50. The operating licenses for CNS and MNS were issued after 1971, and these units meet the GDC in Appendix A of 10 CFR Part 50. The regulations in 1 O CFR 50.55a specify that RCPB components must meet the requirements for Class 1 components in Section Ill of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Section 50.55a further requires, in part, that throughout the service life of a PWR facility, ASME Code Class 1 components meet the requirements, except design and access provisions and pre-service examination requirements, in Section XI, "Rules for lnservice Inspection (ISi) of Nuclear Power Plant Components," of the ASME Code, to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code. In the 1970s, ASME Code Section XI requirements pertaining to ISi of SG tubing were augmented by additional SG tube surveillance requirements (SRs) in the TSs. The regulation at 10 CFR 50.55a, paragraph (b)(2)(iii), states, "if the [TSs] ... include [SRs] for [SGs] different than those in Article IWB-2000, the [ISi] program for [SG] tubing is governed by the requirements in the [TSs]." As part of the plant's licensing basis, applicants for PWR licenses are required to analyze the consequences of postulated design-basis accidents such as an SG tube rupture and main steamline break. These analyses consider the primary-to-secondary leakage that may occur during these events and must show that the offsite radiological consequences do not exceed the applicable limits of the 10 CFR Part 100.11 guidelines for offsite doses (or 10 CFR 50.67, as appropriate), GDC-19 criteria for control room operator doses, or some fraction thereof as appropriate to the accident, or the NRG-approved licensing basis.
-4 -The regulation at 10 CFR 50.36, "Technical specifications," establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. For CNS and MNS, the LCOs (and accompanying action statements) and the SRs in the STS that are relevant to SG tube integrity are in Specification 3.4.13, "Reactor Coolant System Operational Leakage," and Specification 3.4.20 (SR 3.4.20.2), "Steam Generator (SG) Tube Integrity." For ONS, the LCOs (and accompanying action statements) and the SRs in the STS that are relevant to SG tube integrity are in Specification 3.4.13, "Reactor Coolant System Operational Leakage," and Specification 3.4.17 (SR 3.4.17.2), "Steam Generator (SG) Tube Integrity." The SRs in the "Steam Generator (SG) Tube Integrity" specification reference the SG Program, which is defined in the STS administrative controls section. The regulation at 10 CFR 50.36(c)(5) defines 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 STS. The SG Program is defined in Specification 5.5.9 of the STS for CNS, MNS, and ONS, while the reporting requirements relating to implementation of the SG Program are in Specification 5.6.7 of the STS for CNS, MNS, and ONS. STS 5.5.9, "Steam Generator (SG) Program," for CNS, MNS, and ONS, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Tube integrity is maintained by meeting the performance criteria specified in STS 5.5.9 for structural and leakage integrity, consistent with the plant design and licensing bases. STS 5.5.9.a requires that a condition monitoring assessment be performed during each outage, during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met. STS 5.5.9.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 tube-to-tubesheet weld at the tube outlet, and (2) may satisfy the applicable tube repair criteria. The applicable tube repair criteria specified in STS 5.5.9.c are that tubes found during ISi to contain flaws with a depth equal to, or exceeding, 40 percent of the nominal wall thickness shall be plugged, unless the tubes are permitted to remain in service through application of the alternate repair criteria provided in STS 5.5.9.c.1.
-5 -3.0 TECHNICAL EVALUATION The changes in TSTF-510, Revision 2, reflect licensees' early implementation experience with their current TSs. The changes in TSTF-510, Revision 2, are 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's licensing basis will be maintained between SG inspections. The NRC staff approved TSTF-510, Revision 2, for use with the consolidated line item process on October 19, 2011 (ADAMS Accession No. ML 112101604). Other than the variations or deviations discussed below, the licensee is not proposing any variations or deviations from the TS changes described in TSTF-510, Revision 2. As a result, the NRC staff's evaluation is focused on these differences, since the other changes were previously evaluated in the model safety evaluation (ADAMS Accession No. ML 112101513). Additionally, the model safety evaluation contains sections for the various SG tube materials. The sections pertaining to SGs with alloy 600 thermally treated tubes apply to CNS, Unit 2, while the section pertaining to SGs with alloy 690 thermally treated tubes apply to CNS, Unit 1; MNS, Units 1 and 2; and ONS Units 1, 2, and 3. 3.1 Administrative Changes and Variations
- The CNS, MNS, and ONS TSs utilize different numbering than the STS on which TSTF-510, Revision 2, was based. For CNS and MNS, the "Steam Generator Tube Integrity" TS is numbered 3.4.18 rather than 3.4.20, and the "Steam Generator Tube Inspection Report" TS is numbered 5.6.8 rather than 5.6.7. For ONS, the "Steam Generator (SG) Program" TS is numbered 5.5.10 rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.16 rather than 3.4.17, and the "Steam Generator Tube Inspection Report" TS is numbered 5.6.8 rather than 5.6.7.
- An NRC letter dated June 17, 2013 (ADAMS Accession No. ML 13120A541), clarified that if LARs proposing to implement TSTF-510, Revision 2, corrected an administrative inconsistency in paragraph 5.5.9.d.2 of the SG Program, they would not result in removal of submitted LARs from the consolidated line item improvement process. Because CNS, MNS, and ONS do not have any approved tube repair methods, this LAR fixes the administrative inconsistency in paragraphs 5.5.9.d.2 and 5.5.9.d.3 (for CNS), 5.5.9.d.2 (for MNS), and 5.5.1 O.d.2 (for ONS), by replacing "tube repair criteria" with "tube plugging criteria," per the NRC letter dated June 17, 2013.
- The acronym EFPM (effective full power months) is not defined in TSTF-510, Revision 2, but is defined and used in various subsections of CNS TS 5.5.9.d.2 of the proposed LAR.
-6 -The differences noted above are administrative and do not affect the applicability of TSTF-510, Revision 2, to the TSs of CNS, MNS, and ONS. As a result, the NRC staff finds the differences between what was approved for TSTF-510, Revision 2, and what is being proposed acceptable. Thus, the NRC concludes that the requirements of 10 CFR 50.36 are met and the proposed amendments are acceptable. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the North Carolina and South Carolina State officials were notified of the proposed issuance of the amendments. The State officials had no comments. 5.0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on June 23, 2015 (80 FR 35981). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments. 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: A. Johnson Date: April 26, 2016 If you have any questions, please contact me at 301-415-2481 or Ed.Miller@nrc.gov. Docket Nos. 50-413, 50-414, 50-369, 50-370, 50-269, 50-270, and 50-287
Enclosures:
1. Amendment No. 280 to NPF-35 2. Amendment No. 276 to NPF-52 3. Amendment No. 284 to NPF-9 4. Amendment No. 263 to NPF-17 5. Amendment No. 396 to DPR-38 6. Amendment No. 398 to DPR-47 7. Amendment No. 397 to DPR-55 8. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPL2-1 R/F RidsACRS_MailCTR Resource RidsNrrDeEsgb Resource RidsNrrDirsStsb Resource RidsNrrDorlDpr Resource RidsNrrDorllpl2-1 Resource RidsNrrPMCatawba Resource RidsNrrPMMcGuire Resource RidsNrrPMOconee Resource RidsNrrLALRonewicz Resource RidsRgn2MailCenter Resource AJohnson, NRR ADAMS Accession No.: ML16075A301 OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA NAME MOrenak LRonewicz DATE 03/23/16 04/13/16 OFFICE OGC-NLO DORL/LPL2-1 /BC NAME SF owler MMarkley DATE 04/8/16 04/26/16
Sincerely,/RA/ G. Edward Miller, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation *SE dated DE/ESGB/BC* DSS/STSB/BC GKulsea RElliott (SAnderson for) 03/04/16 03/30/16 DORL/LPL2-1 /PM GMiller 04/26/16 OFFICIAL RECORD COPY