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| | number = ML16055A359 | | | number = ML16055A359 |
| | issue date = 03/09/2016 | | | issue date = 03/09/2016 |
| | title = Diablo Canyon, Units 1 and 2 - Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (Colr) Paragon and Nexus Methodologies (CAC MF6120 a | | | title = Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies |
| | author name = Lingam S P | | | author name = Lingam S |
| | author affiliation = NRC/NRR/DORL/LPLIV-1 | | | author affiliation = NRC/NRR/DORL/LPLIV-1 |
| | addressee name = Halpin E D | | | addressee name = Halpin E |
| | addressee affiliation = Pacific Gas & Electric Co | | | addressee affiliation = Pacific Gas & Electric Co |
| | docket = 05000275, 05000323 | | | docket = 05000275, 05000323 |
| | license number = DPR-080, DPR-082 | | | license number = DPR-080, DPR-082 |
| | contact person = Lingam S P | | | contact person = Lingam S |
| | case reference number = CAC MF6120, CAC MF6121 | | | case reference number = CAC MF6120, CAC MF6121 |
| | document type = Letter, License-Operating (New/Renewal/Amendments) DKT 50, Safety Evaluation, Technical Specifications | | | document type = Letter, License-Operating (New/Renewal/Amendments) DKT 50, Safety Evaluation, Technical Specifications |
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| =Text= | | =Text= |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424 March 9, 2016 | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2016 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING REVISION TO UPDATED FINAL SAFETY ANALYSIS REPORT FOR BEST ESTIMATE ANALYZER FOR THE CORE OPERATIONS-NUCLEAR POWER DISTRIBUTION MONITORING SYSTEM METHODOLOGY AND TECHNICAL SPECIFICATION 5.6.5 (CAC NOS. MF6120 AND MF6121) | | DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REVISION TO UPDATED FINAL SAFETY ANALYSIS REPORT FOR BEST ESTIMATE ANALYZER FOR THE CORE OPERATIONS-NUCLEAR POWER DISTRIBUTION MONITORING SYSTEM METHODOLOGY AND TECHNICAL SPECIFICATION 5.6.5 (CAC NOS. MF6120 AND MF6121) |
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| ==Dear Mr. Halpin:== | | ==Dear Mr. Halpin:== |
| The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 224 to Facility Operating License No. DPR-80 and Amendment No. 226 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TSs) in response to your application dated April 16, 2015. The amendments revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the DCPP UFSAR Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System." The amendments also revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with NRG-approved proprietary TR WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." | | |
| E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Docket Nos. 50-275 and 50-323 | | The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 224 to Facility Operating License No. DPR-80 and Amendment No. 226 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TSs) in response to your application dated April 16, 2015. |
| | The amendments revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the DCPP UFSAR Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System." The amendments also revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with NRG-approved proprietary TR WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." |
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| | E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. |
| | Sincerely, S.vv-o._ cy ' "~ |
| | Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323 |
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| ==Enclosures:== | | ==Enclosures:== |
| 1. Amendment No. 224 to DPR-80 2. Amendment No. 226 to DPR-82 3. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, S.vv-o._ cy ' Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. DPR-80 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 21. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015. | | : 1. Amendment No. 224 to DPR-80 |
| | : 2. Amendment No. 226 to DPR-82 |
| | : 3. Safety Evaluation cc w/encls: Distribution via Listserv |
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| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. DPR-80 |
| | : 1. The Nuclear Regulatory Commission (the Commission) has found that: |
| | A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows: |
| | (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. |
| | : 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 21. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015. |
| | FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation |
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| ==Attachment:== | | ==Attachment:== |
| Changes to the Facility Operating License No. DPR-80 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 9 , 2O1 6 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 226 License No. DPR-82 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows: (2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 20. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015. | | |
| | Changes to the Facility Operating License No. DPR-80 and Technical Specifications Date of Issuance: March 9 , 2O1 6 |
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| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 226 License No. DPR-82 |
| | : 1. The Nuclear Regulatory Commission (the Commission) has found that: |
| | A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows: |
| | (2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. |
| | : 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 20. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015. |
| | FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation |
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| ==Attachment:== | | ==Attachment:== |
| Changes to the Facility Operating License No. DPR-82 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 9 , 2o1 6 ATTACHMENT TO LICENSE AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of the Facility Operating License Nos. DPR-80 and DPR-82, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE REMOVE 5.0-20 5.0-21 INSERT Technical Specifications INSERT 5.0-20 5.0-21 (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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein. (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224 are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. (3) Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as: a. Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential; Amendment No. 224 (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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein. (2) Technical Specifications (SSER 32. Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. (3) Initial Test Program (SSER 31. Section 4.4.1) Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change. *The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed. Amendment No. 226 5.6 Reporting Requirements Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: 1. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control F0 Surveillance Technical Specification, (Westinghouse Proprietary), 2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (Westinghouse Proprietary), 3. WCAP-8385, Power Distribution Control and Load Following Procedures, (Westinghouse Proprietary), 4. WCAP-10054-P-A, Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse Proprietary), 5. WCAP-10054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model, July 1997 (Westinghouse Proprietary), 6. WCAP-12945-P-A, Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis, June 1996 (Westinghouse Proprietary), 7. WCAP-12945-P-A, Addendum 1-A, Revision 0, "Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best Estimate LOCA Evaluation Models," December 2004. (Westinghouse Proprietary) (Unit 1 Only), 8. WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005. (Westinghouse Proprietary) (Unit 2 Only), 9. WCAP-8567-P-A, "Improved Thermal Design Procedure," 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." DIABLO CANYON -UNITS 1 & 2 (continued) 5.0-20 Unit 1 -Amendment No. 224 Unit 2 -Amendment No. 226 5.6 Reporting Requirements Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met. d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Protection (LTOP) arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: 1. Specification 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and 2. Specification 3.4.12, "Low Temperature Overpressure Protection (L TOP) System." DIABLO CANYON -UNITS 1 & 2 5.0-21 (continued) Unit 1 -Amendment No. 498, 224 Unit 2 -Amendment No. * .+w, 226 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323
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| ==1.0 INTRODUCTION==
| | Changes to the Facility Operating License No. DPR-82 and Technical Specifications Date of Issuance: March 9 , 2o1 6 |
| By letter dated April 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15107A333), Pacific Gas and Electric Company (the licensee) submitted a license amendment request (LAR) to revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System," September 2012 (ADAMS Accession No. ML 12270A386 for non-proprietary version) for Diablo Canyon Power Plant (DCPP), Units 1 and 2. This LAR also proposes to revise Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988, with U.S. Nuclear Regulatory Commission (NRC)-approved proprietary TR WCAP-16045-P-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 (ADAMS Accession No. ML042250322 for non-proprietary version), and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," August 2007 (ADAMS Accession No. ML072570329 for non-proprietary version).
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| ==2.0 REGULATORY EVALUATION==
| | ATTACHMENT TO LICENSE AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of the Facility Operating License Nos. DPR-80 and DPR-82, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. |
| The regulatory requirements and criteria specifically applicable to the proposed license amendment are summarized and addressed below: NRC Generic Letter 88-16 The guidance in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988 (ADAMS Accession No. ML031130447), indicates that it is acceptable for licensees to control reactor physics parameter limits by Enclosure 3 specifying the calculation methodology. The generic letter indicates that such parameter limits may be removed from TSs and placed in a cycle-specific COLR. The COLR is defined in the TSs, and the Reporting Requirements in the TSs require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised. Generic Letter 88-16 also recommends that licensees include, in the TSs, a list of references for the NRG-approved methodologies, which are used to generate the cycle-specific parameter operating limits. 10 CFR 50.34 Section 50.34, "Contents of applications; technical information," of Title 10 to the Code of Federal Regulations ( 10 CFR) Part 50 requires that safety analysis reports include analysis and evaluation of the design and performance of structures, systems, and components for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload design process, licensees (or vendors) perform reload safety evaluations to ensure that their safety analyses remain bounding for the design cycle. To confirm that the analyses remain bounding, licensees confirm that key inputs to the safety analyses (such as the critical power ratio) are conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a reanalysis or reevaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied. 10 CFR 50.36 In Section 50.36, "Technical specifications," of 10 CFR Part 50, the Commission established its regulatory requirements related to the content of the TSs. Consistent with 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: 1) safety limits, limiting safety system settings (LSSS), and limiting control settings; 2) limiting conditions for operations (LCOs); 3) surveillance requirements; 4) design features; and 5) administrative controls.
| | Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 5.0-20 5.0-20 5.0-21 5.0-21 |
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| ==3.0 TECHNICAL EVALUATION==
| | (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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. |
| 3.1 Proposed TS Changes Current TS 5.6.5 Section b states, in part: 10. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores." Revised TS 5.6.5 Section b would state, in part: 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."
| | C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: |
| 3.2 Technical Evaluation of the Application of the Revised BEACON Methodology The NRG-approved BEACON methodology described in Westinghouse TR WCAP-12472-P-A Addendum 4, "BEACON Core Monitoring and Operation Support System, Addendum 4", has been reviewed for application to DCPP, Units 1 and 2, as summarized below. The BEACON core monitoring system currently in use at DCPP, Units 1 and 2, WCAP-12472-NP-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," January 2000 (ADAMS Accession No. ML003678155), employs a predictive three-dimensional nodal model of the reactor core calibrated to an in-core flux map, which is periodically obtained utilizing the movable in-core detector system {MIDS). The resultant nodal power distribution is frequently updated as part of the core follow calculation, and is adjusted for the axial offset measured by the ex-core detectors; the radial distribution of power is further corrected using a calorimetric process based on core exit thermocouple measurements. Updating of the power distribution based on ex-core detector and core exit thermocouple readings is performed on a nearly continuous basis. A reasonably accurate representation of the core power distribution is determined through use of the MIDS. The flux traces obtained with MIDS are used to calibrate the BEACON predictive model, and also to calibrate both the ex-core detector response, and the core exit thermal couple readings. Between MIDS in-core flux scans, the ex-core and core exit thermocouple calibration factors are utilized in the BEACON power distribution calculation. A key factor in the BEACON methodology is the uncertainties applied to the BEACON measured power distributions. One component of the measurement uncertainty is the variability of the thermocouple calibration factors (also referred to as mixing factors). The mixing factor uncertainty (standard deviation) is shown to exhibit a power level dependency due to the separate effects of decreased temperature difference (lff RPL) at lower power levels and changes in cross-flow patterns at reduced power levels. In the current methodology, the mixing factor uncertainty is determined as a function of power level by accounting for the effects of decreased LH RPL and changes in cross-flow patterns separately. Additionally, an average thermocouple standard deviation at hot full-power conditions is determined from past performance of thermocouples from previous cycles and utilized to account for measurement uncertainty. The updated methodology described in WCAP-12472-P-A, Addendum 4, utilizes core exit thermocouple data collected during initial power ascension following a refueling outage. This process simultaneously considers the effects of both LlT RPL and cross-flow patterns on mixing factor uncertainty, and eliminates the reliance on previous cycle thermocouple data, thus resulting in a more accurate and reliable evaluation of thermocouple uncertainties. Additionally, the use of current cycle data allows for the inclusion of any changes to the thermocouples since the previous cycle. These changes include the failure or replacement of individual thermocouples, relocation of a thermocouple or instrumentation system changes. The NRC Safety Evaluation Report (SER) dated August 9, 2012, approving WCAP-12472-P-A, Addendum 4 (ADAMS Accession No. ML 12150A 14 7), states that the updated BEACON thermocouple methodology is acceptable. The NRC staff's review of the applicant's LAR shows that the licensee has considered the specific application of the NRG-approved methodology to DCPP, Units 1 and 2, including the plant-specific configuration of the core exit thermocouple design and treatment of thermocouple failures. In the NRC staff's audit report dated February 23, 2016 (ADAMS Accession No. ML 16053A323), the staff noted that the licensee stated and confirmed that WCAP-12472-P-A, Addendum 4 may replace WCAP-12472-P-A, Addendum 1-A without loss of applicable licensing basis. The proposed change in this LAR does not impact DCPP's compliance with the NRC SER conditions associated with the approval of the original BEACON WCAP-12472 and WCAP-12472-P-A, Addendum 1-A. Therefore, the three licensing conditions cited in the NRC SER documenting the staff's approval of the original BEACON topical report WCAP-12472-P-A, Addendum 1-A remains applicable with the implementation of Addendum 4 of WCAP-12472-P-A. Based on the above, the NRC staff finds that the methodology described in WCAP-12472-NP-A, Addendum 4 may be implemented at DCPP, Units 1 and 2 as proposed by the licensee. 3.3 Technical Evaluation of the Application of WCAP-16045-P-A and addendum 3.3.1 PARAGON and NEXUS Code Descriptions The PARAGON code, described in WCAP-16045-P-A, is a neutron transport code, intended to replace PHOENIX-P for use in providing nuclear input data for the Westinghouse proprietary Advanced Nodal Code (ANC), which is a core simulator code. NEXUS (WCAP-16045-NP-A, Addendum 1 A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," August 2007) is an improvement to the PARAGON code system. WCAP-16045-NP-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 and WCAP-16045-NP-A, Addendum 1-A are NRG-approved, based on extensive validation and verification. As described in the topical report and addendum, the data set used for validation and verification includes numerous criticality experiments and plant data. The plant data reflect a widely varied set of operating conditions that includes numerous fuel designs, lattice geometries, and burnable absorber loadings in use in the Westinghouse-and Combustion Engineering-designed nuclear steam supply systems. The verification was expanded to include comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology. The results showed excellent agreement between NEXUS predictions and available data. 3.3.2 Applicability to DCPP, Units 1 and 2 The NRC staff reviewed information contained in the UFSAR to verify that the database discussed in Section 3.3.1 of this SE includes fuel with similar characteristics to that used at DCPP. Chapter 4, "Reactor," of the UFSAR indicates that DCPP, Units 1 and 2, use 17x17 Westinghouse optimized fuel. This design feature is represented in the database described above; therefore, the NRC staff determined that the NEXUS code system is acceptable for use at DCPP. The only condition or limitation established in the SERs approving PARAGON and NEXUS precludes their use for mixed-oxide (MOX) cores. The NRC staff verified and documented in its audit report dated February 23, 2016, that DCPP does not use MOX fuel in the core reload design. The NRC staff determined, therefore, that a separate licensing action would be required in order to use MOX fuel at DCPP, so the MOX limitation in the SERs approving PARAGON and NEXUS is satisfied. 3.4 Additional Review Topics The addendum to WCAP-16045-P-A brings about a change to the way boron letdown curves are calculated and provide input into the overall nuclear design method. Because of this, in previous reviews of PARAGON-NEXUS/ANG applications, the NRC staff has found it necessary to verify that no changes were made to the analysis methods for post loss-of-coolant accident (LOCA) subcriticality and boric acid precipitation behavior. The PARAGON code, as described in WCAP-16045-P-A, is a neutron transport code intended to replace PHOENIX-P for use in providing cross-section data to the Westinghouse proprietary ANC, a core simulator code. NEXUS, described in WCAP-16045-P-A, Addendum 1-A, is an improvement to the PARAGON/ANG code system that changes the method of communicating the nuclear data output from PARAGON to ANC. Rather than using boron letdown curves, NEXUS accounts for variations in the neutron spectrum by parameterizing the PARAGON cross-section output and reconstructing it within ANC. WCAP-16045-P-A and addendum are NRG-approved, and the NRC staff has previously determined that they have been extensively validated and verified using a variety of experiments and plant data that account for a wide range of differences in operating conditions. In Addendum 1-A, the verification was expanded to include the comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology, showing excellent agreement between NEXUS predictions and available data. The use of WCAP-16045-NP-A, Addendum 1, for DCPP does not affect the inputs or method(s) for ensuring core subcriticality, both short-and long-term post-LOCA, thereby precluding the potential for return to power following a large break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the use of Addendum 1, the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection will continue to remain valid. Core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs and refueling boron concentration, will continue to be calculated using NRC approved methods. The NRC staff concludes this is acceptable, because it confirms that the implementation of PARAGON and NEXUS at DCPP will not affect the calculation of post-LOCA boron requirements or emergency procedures to mitigate post-LOCA boric acid precipitation. Based on the above, the staff concludes that the existing post-LOCA analyses for subcriticality and long-term cooling remain applicable.
| | (1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein. |
| 3.5 Summary The NRC staff has determined that the proposed changes in the UFSAR to the BEACON core monitoring system methodology described in WCAP-12472-P-A, Addendum 4 and the TS changes required to replace the PHOENIX-P code system with the PARAGON/NEXUS code system are acceptable for DCPP. This determination is based on 1) that the original BEACON TR WCAP-12472-P-A remains applicable, 2) that PARAGON and NEXUS are NRG-approved methods, which have been determined to be applicable to DCPP, and 3) that the DCPP post-LOCA subcriticality and boric acid precipitation analyses will be unaffected by the proposed changes. Based on the above, the staff concludes that the proposed TS revision is consistent with the guidance provided in GL 88-16, and that the generic qualification and validation of the PARAGON-NEXUS/ANG system satisfies the requirements of 10 CFR 50.34 and 10 CFR 50.36.
| | (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224 are hereby incorporated in the license. |
| | Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. |
| | (3) Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as: |
| | : a. Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential; Amendment No. 224 |
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|
| ==4.0 STATE CONSULTATION==
| | (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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. |
| In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
| | C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: |
| | (1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein. |
| | (2) Technical Specifications (SSER 32. Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license. |
| | Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. |
| | (3) Initial Test Program (SSER 31. Section 4.4.1) |
| | Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change. |
| | *The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed. |
| | Amendment No. 226 |
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| |
|
| ==5.0 ENVIRONMENTAL CONSIDERATION==
| | Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) |
| The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 published in the Federal Register on June 9, 2015 (80 FR 32628). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments. | | : b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: |
| | : 1. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control F0 Surveillance Technical Specification, (Westinghouse Proprietary), |
| | : 2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (Westinghouse Proprietary), |
| | : 3. WCAP-8385, Power Distribution Control and Load Following Procedures, (Westinghouse Proprietary), |
| | : 4. WCAP-10054-P-A, Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse Proprietary), |
| | : 5. WCAP-10054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model, July 1997 (Westinghouse Proprietary), |
| | : 6. WCAP-12945-P-A, Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis, June 1996 (Westinghouse Proprietary), |
| | : 7. WCAP-12945-P-A, Addendum 1-A, Revision 0, "Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best Estimate LOCA Evaluation Models," December 2004. (Westinghouse Proprietary) (Unit 1 Only), |
| | : 8. WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005. (Westinghouse Proprietary) |
| | (Unit 2 Only), |
| | : 9. WCAP-8567-P-A, "Improved Thermal Design Procedure," |
| | : 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and |
| | : 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." |
| | (continued) |
| | DIABLO CANYON - UNITS 1 & 2 5.0-20 Unit 1 - Amendment No. ~.~,+94-,4-%,+98, 224 Unit 2 - Amendment No. ~.~.~.4-96,+99, 226 |
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|
| ==6.0 CONCLUSION==
| | Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) |
| 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: F. Forsaty, NRR Date: March 9, 2016 E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Docket Nos. 50-275 and 50-323 | | : c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met. |
| | : d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. |
| | 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) |
| | : a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Protection (LTOP) arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: |
| | : 1. Specification 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and |
| | : 2. Specification 3.4.12, "Low Temperature Overpressure Protection (LTOP) |
| | System." |
| | (continued) |
| | DIABLO CANYON - UNITS 1 & 2 5.0-21 Unit 1 - Amendment No. ~. 498, 224 Unit 2 - Amendment No. ~ * .+w, 226 |
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|
| ==Enclosures:== | | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323 |
| 1. Amendment No. 224 to DPR-80 2. Amendment No. 226 to DPR-82 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPL4-1 Reading RidsACRS_MailCTR Resource RidsNrrDorlDpr Resource RidsNrrDorllpl4-1 Resource RidsNrrDssSnpb Resource RidsNrrDssSrxb Resource ADAMS Accession No. ML 16055A359 Sincerely, IRA/ Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDssStsb Resource RidsNrrLAJBurkhardt Resource RidsNrrPMDiabloCanyon Resource RidsRgn4Mai1Center Resource KWest, NRR/DSS/STSB FForsaty, NRR/DSS/SRXB *SE via memo OFFICE NRR/DORULPL4-1 /PM NRR/DORULPL4-1 /LA NRR/DSS/SRXB/BC (A)* NRR/DSS/SNPB/BC NAME SLingam JBurkhardt CJackson (EOesterle for) JDean DATE 2/26/16 2/25/16 2/18/16 3/7/16 OFFICE NRR/DSS/STSB/BC OGC -NLO w/comments NRR/DORULPL4-1/BC N RR/DORULPL4-1 /PM NAME RElliott JLindell RPascarelli (FLyon for) SLingam DATE 3/1/16 3/4/16 3/9/16 3/9/16 OFFICIAL RECORD COPY | | |
| }} | | ==1.0 INTRODUCTION== |
| | |
| | By letter dated April 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15107A333), Pacific Gas and Electric Company (the licensee) submitted a license amendment request (LAR) to revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System," September 2012 (ADAMS Accession No. ML12270A386 for non-proprietary version) for Diablo Canyon Power Plant (DCPP), Units 1 and 2. This LAR also proposes to revise Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," |
| | Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988, with U.S. Nuclear Regulatory Commission (NRC)-approved proprietary TR WCAP-16045-P-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 (ADAMS Accession No. ML042250322 for non-proprietary version), and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," August 2007 (ADAMS Accession No. ML072570329 for non-proprietary version). |
| | |
| | ==2.0 REGULATORY EVALUATION== |
| | |
| | The regulatory requirements and criteria specifically applicable to the proposed license amendment are summarized and addressed below: |
| | NRC Generic Letter 88-16 The guidance in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988 (ADAMS Accession No. ML031130447), |
| | indicates that it is acceptable for licensees to control reactor physics parameter limits by Enclosure 3 |
| | |
| | specifying the calculation methodology. The generic letter indicates that such parameter limits may be removed from TSs and placed in a cycle-specific COLR. The COLR is defined in the TSs, and the Reporting Requirements in the TSs require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised. Generic Letter 88-16 also recommends that licensees include, in the TSs, a list of references for the NRG-approved methodologies, which are used to generate the cycle-specific parameter operating limits. |
| | 10 CFR 50.34 Section 50.34, "Contents of applications; technical information," of Title 10 to the Code of Federal Regulations ( 10 CFR) Part 50 requires that safety analysis reports include analysis and evaluation of the design and performance of structures, systems, and components for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload design process, licensees (or vendors) perform reload safety evaluations to ensure that their safety analyses remain bounding for the design cycle. |
| | To confirm that the analyses remain bounding, licensees confirm that key inputs to the safety analyses (such as the critical power ratio) are conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a reanalysis or reevaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied. |
| | 10 CFR 50.36 In Section 50.36, "Technical specifications," of 10 CFR Part 50, the Commission established its regulatory requirements related to the content of the TSs. Consistent with 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: |
| | : 1) safety limits, limiting safety system settings (LSSS), and limiting control settings; 2) limiting conditions for operations (LCOs); 3) surveillance requirements; 4) design features; and |
| | : 5) administrative controls. |
| | |
| | ==3.0 TECHNICAL EVALUATION== |
| | |
| | 3.1 Proposed TS Changes Current TS 5.6.5 Section b states, in part: |
| | : 10. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores." |
| | Revised TS 5.6.5 Section b would state, in part: |
| | : 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and |
| | : 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." |
| | |
| | 3.2 Technical Evaluation of the Application of the Revised BEACON Methodology The NRG-approved BEACON methodology described in Westinghouse TR WCAP-12472-P-A Addendum 4, "BEACON Core Monitoring and Operation Support System, Addendum 4", has been reviewed for application to DCPP, Units 1 and 2, as summarized below. |
| | The BEACON core monitoring system currently in use at DCPP, Units 1 and 2, WCAP-12472-NP-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," January 2000 (ADAMS Accession No. ML003678155), employs a predictive three-dimensional nodal model of the reactor core calibrated to an in-core flux map, which is periodically obtained utilizing the movable in-core detector system {MIDS). The resultant nodal power distribution is frequently updated as part of the core follow calculation, and is adjusted for the axial offset measured by the ex-core detectors; the radial distribution of power is further corrected using a calorimetric process based on core exit thermocouple measurements. |
| | Updating of the power distribution based on ex-core detector and core exit thermocouple readings is performed on a nearly continuous basis. A reasonably accurate representation of the core power distribution is determined through use of the MIDS. The flux traces obtained with MIDS are used to calibrate the BEACON predictive model, and also to calibrate both the ex-core detector response, and the core exit thermal couple readings. Between MIDS in-core flux scans, the ex-core and core exit thermocouple calibration factors are utilized in the BEACON power distribution calculation. |
| | A key factor in the BEACON methodology is the uncertainties applied to the BEACON measured power distributions. One component of the measurement uncertainty is the variability of the thermocouple calibration factors (also referred to as mixing factors). The mixing factor uncertainty (standard deviation) is shown to exhibit a power level dependency due to the separate effects of decreased temperature difference (lff RPL) at lower power levels and changes in cross-flow patterns at reduced power levels. |
| | In the current methodology, the mixing factor uncertainty is determined as a function of power level by accounting for the effects of decreased LHRPL and changes in cross-flow patterns separately. Additionally, an average thermocouple standard deviation at hot full-power conditions is determined from past performance of thermocouples from previous cycles and utilized to account for measurement uncertainty. |
| | The updated methodology described in WCAP-12472-P-A, Addendum 4, utilizes core exit thermocouple data collected during initial power ascension following a refueling outage. This process simultaneously considers the effects of both LlT RPL and cross-flow patterns on mixing factor uncertainty, and eliminates the reliance on previous cycle thermocouple data, thus resulting in a more accurate and reliable evaluation of thermocouple uncertainties. Additionally, the use of current cycle data allows for the inclusion of any changes to the thermocouples since the previous cycle. These changes include the failure or replacement of individual thermocouples, relocation of a thermocouple or instrumentation system changes. |
| | The NRC Safety Evaluation Report (SER) dated August 9, 2012, approving WCAP-12472-P-A, Addendum 4 (ADAMS Accession No. ML12150A147), states that the updated BEACON thermocouple methodology is acceptable. The NRC staff's review of the applicant's LAR shows that the licensee has considered the specific application of the NRG-approved methodology to |
| | |
| | DCPP, Units 1 and 2, including the plant-specific configuration of the core exit thermocouple design and treatment of thermocouple failures. |
| | In the NRC staff's audit report dated February 23, 2016 (ADAMS Accession No. ML16053A323), the staff noted that the licensee stated and confirmed that WCAP-12472-P-A, Addendum 4 may replace WCAP-12472-P-A, Addendum 1-A without loss of applicable licensing basis. |
| | The proposed change in this LAR does not impact DCPP's compliance with the NRC SER conditions associated with the approval of the original BEACON WCAP-12472 and WCAP-12472-P-A, Addendum 1-A. Therefore, the three licensing conditions cited in the NRC SER documenting the staff's approval of the original BEACON topical report WCAP-12472-P-A, Addendum 1-A remains applicable with the implementation of Addendum 4 of WCAP-12472-P-A. |
| | Based on the above, the NRC staff finds that the methodology described in WCAP-12472-NP-A, Addendum 4 may be implemented at DCPP, Units 1 and 2 as proposed by the licensee. |
| | 3.3 Technical Evaluation of the Application of WCAP-16045-P-A and addendum 3.3.1 PARAGON and NEXUS Code Descriptions The PARAGON code, described in WCAP-16045-P-A, is a neutron transport code, intended to replace PHOENIX-P for use in providing nuclear input data for the Westinghouse proprietary Advanced Nodal Code (ANC), which is a core simulator code. NEXUS (WCAP-16045-NP-A, Addendum 1A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," |
| | August 2007) is an improvement to the PARAGON code system. |
| | WCAP-16045-NP-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 and WCAP-16045-NP-A, Addendum 1-A are NRG-approved, based on extensive validation and verification. As described in the topical report and addendum, the data set used for validation and verification includes numerous criticality experiments and plant data. The plant data reflect a widely varied set of operating conditions that includes numerous fuel designs, lattice geometries, and burnable absorber loadings in use in the Westinghouse- and Combustion Engineering-designed nuclear steam supply systems. The verification was expanded to include comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology. The results showed excellent agreement between NEXUS predictions and available data. |
| | 3.3.2 Applicability to DCPP, Units 1 and 2 The NRC staff reviewed information contained in the UFSAR to verify that the database discussed in Section 3.3.1 of this SE includes fuel with similar characteristics to that used at DCPP. Chapter 4, "Reactor," of the UFSAR indicates that DCPP, Units 1 and 2, use 17x17 Westinghouse optimized fuel. This design feature is represented in the database described above; therefore, the NRC staff determined that the NEXUS code system is acceptable for use at DCPP. The only condition or limitation established in the SERs approving PARAGON and NEXUS precludes their use for mixed-oxide (MOX) cores. The NRC staff verified and |
| | |
| | documented in its audit report dated February 23, 2016, that DCPP does not use MOX fuel in the core reload design. |
| | The NRC staff determined, therefore, that a separate licensing action would be required in order to use MOX fuel at DCPP, so the MOX limitation in the SERs approving PARAGON and NEXUS is satisfied. |
| | 3.4 Additional Review Topics The addendum to WCAP-16045-P-A brings about a change to the way boron letdown curves are calculated and provide input into the overall nuclear design method. Because of this, in previous reviews of PARAGON-NEXUS/ANG applications, the NRC staff has found it necessary to verify that no changes were made to the analysis methods for post loss-of-coolant accident (LOCA) subcriticality and boric acid precipitation behavior. |
| | The PARAGON code, as described in WCAP-16045-P-A, is a neutron transport code intended to replace PHOENIX-P for use in providing cross-section data to the Westinghouse proprietary ANC, a core simulator code. NEXUS, described in WCAP-16045-P-A, Addendum 1-A, is an improvement to the PARAGON/ANG code system that changes the method of communicating the nuclear data output from PARAGON to ANC. Rather than using boron letdown curves, NEXUS accounts for variations in the neutron spectrum by parameterizing the PARAGON cross-section output and reconstructing it within ANC. |
| | WCAP-16045-P-A and addendum are NRG-approved, and the NRC staff has previously determined that they have been extensively validated and verified using a variety of experiments and plant data that account for a wide range of differences in operating conditions. |
| | In Addendum 1-A, the verification was expanded to include the comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology, showing excellent agreement between NEXUS predictions and available data. |
| | The use of WCAP-16045-NP-A, Addendum 1, for DCPP does not affect the inputs or method(s) for ensuring core subcriticality, both short- and long-term post-LOCA, thereby precluding the potential for return to power following a large break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the use of Addendum 1, the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection will continue to remain valid. Core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs and refueling boron concentration, will continue to be calculated using NRC approved methods. |
| | The NRC staff concludes this is acceptable, because it confirms that the implementation of PARAGON and NEXUS at DCPP will not affect the calculation of post-LOCA boron requirements or emergency procedures to mitigate post-LOCA boric acid precipitation. Based on the above, the staff concludes that the existing post-LOCA analyses for subcriticality and long-term cooling remain applicable. |
| | |
| | 3.5 Summary The NRC staff has determined that the proposed changes in the UFSAR to the BEACON core monitoring system methodology described in WCAP-12472-P-A, Addendum 4 and the TS changes required to replace the PHOENIX-P code system with the PARAGON/NEXUS code system are acceptable for DCPP. This determination is based on 1) that the original BEACON TR WCAP-12472-P-A remains applicable, 2) that PARAGON and NEXUS are NRG-approved methods, which have been determined to be applicable to DCPP, and 3) that the DCPP post-LOCA subcriticality and boric acid precipitation analyses will be unaffected by the proposed changes. Based on the above, the staff concludes that the proposed TS revision is consistent with the guidance provided in GL 88-16, and that the generic qualification and validation of the PARAGON-NEXUS/ANG system satisfies the requirements of 10 CFR 50.34 and 10 CFR 50.36. |
| | |
| | ==4.0 STATE CONSULTATION== |
| | |
| | In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments. |
| | |
| | ==5.0 ENVIRONMENTAL CONSIDERATION== |
| | |
| | The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 published in the Federal Register on June 9, 2015 (80 FR 32628). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). |
| | Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need 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: F. Forsaty, NRR Date: March 9, 2016 |
| | |
| | ML16055A359 *SE via memo OFFICE NRR/DORULPL4-1 /PM NRR/DORULPL4-1 /LA NRR/DSS/SRXB/BC (A)* NRR/DSS/SNPB/BC NAME SLingam JBurkhardt CJackson (EOesterle for) JDean DATE 2/26/16 2/25/16 2/18/16 3/7/16 OFFICE NRR/DSS/STSB/BC OGC - NLO w/comments NRR/DORULPL4-1/BC N RR/DORULPL4-1 /PM NAME RElliott JLindell RPascarelli (FLyon for) SLingam DATE 3/1/16 3/4/16 3/9/16 3/9/16}} |
Letter Sequence Approval |
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MONTHYEARML15355A1572015-12-24024 December 2015 Regulatory Audit Plans for 01/12-14/2016 Audit at Westinghouse Facility to Support Review of License Amendment Requests for Adoption of Alternative Source Term and Revision of the UFSAR for Beacon Methodology Project stage: Other ML15343A1272015-12-24024 December 2015 Summary of Meeting with Pacific Gas and Electric to Discuss Diablo Canyon, Units 1 and 2 Amendment Request to Revise UFSAR for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies Project stage: Meeting ML16053A3232016-02-23023 February 2016 Audit Report for 01/12-14/2016 Regulatory Audit at Westinghouse Facility to Support Review of License Amendment Request for Revision of the UFSAR for Beacon Methodology Project stage: Approval ML16055A3592016-03-0909 March 2016 Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies Project stage: Approval DCL-16-071, Core Operating Limits Reports for Cycle 202016-07-11011 July 2016 Core Operating Limits Reports for Cycle 20 Project stage: Other 2016-02-23
[Table View] |
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Category:Letter
MONTHYEARIR 05000275/20230042024-02-0909 February 2024 Integrated Inspection Report 05000275/2023004 and 05000323/2023004 ML24002B1802024-01-29029 January 2024 LRA Audit Plan DCL-24-010, Nuclear Material Transaction Report for New Fuel2024-01-29029 January 2024 Nuclear Material Transaction Report for New Fuel ML24018A0152024-01-29029 January 2024 License Renewal Application Review Schedule Letter ML24017A2492024-01-24024 January 2024 Letter to Neil Peyron, Chairman, Tule River Tribe, Re. Diablo Canyon ML24024A1752024-01-24024 January 2024 Letter to Tribal Council San Luis Obispo County Chumash Indians on Section 106 Consultation and Scoping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24003A8902024-01-24024 January 2024 Letter to P. Gerfen - Diablo Canyon Notice of Intent to Conduct Scoping and Prepare an Environmental Impact Statement ML24012A1582024-01-24024 January 2024 Letter to Hon. Violet Sage Walker, Chairwoman Northern Chumash Tribal Council on Section 106 Consultation and Scoping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24012A0062024-01-24024 January 2024 Achp Scoping Letter for Diablo Canyon License Renewal ML24012A0552024-01-24024 January 2024 Letter to J. Polanco, Shpo, on Request to Initiate Section 106 Consultation and Scooping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24024A1562024-01-24024 January 2024 Letter to Hon. Gabe Frausto, Coastal Band of Chumash Indians on Section 106 Consultation and Scoping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24024A1652024-01-24024 January 2024 Letter to Hon. Mona Olivas Tucker, Yak Tityu Tityu Yak Tilhini Northern Chumas Indians on Section 106 Consultation and Scoping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24024A1612024-01-24024 January 2024 Letter Hon. Gary Pierce, Salian Tribe of Monterey and San Luis Obispo Counties on Section 106 Consultation and Scoping Process for the Environmental Review of Diablo Canyon Nuclear Plant, Unit 1 and 2 License Renewal Application ML24012A0362024-01-24024 January 2024 Request to Initiate Section 106 Consultation and Scoping Process for the Environmental Review License Renewal Application DCL-24-009, Nuclear Material Transaction Report for New Fuel2024-01-17017 January 2024 Nuclear Material Transaction Report for New Fuel DCL-24-008, Schedule Considerations for Review of the DCPP License Renewal Application2024-01-17017 January 2024 Schedule Considerations for Review of the DCPP License Renewal Application DCL-24-004, Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2024-01-15015 January 2024 Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b IR 05000275/20230112024-01-12012 January 2024 NRC License Renewal Phase 1 Inspection Report 05000275/2023011 DCL-23-129, Nuclear Material Transaction Report for New Fuel2023-12-27027 December 2023 Nuclear Material Transaction Report for New Fuel ML23326A0122023-12-21021 December 2023 12-21-23 Letter to the Honorable Byron Donalds from Chair Hanson Responds to Letter Regarding 2.206 Petition to Close Diablo Canyon Nuclear Power Plant, Unit 1 ML23341A0042023-12-19019 December 2023 LRA Acceptance Letter ML23352A2342023-12-18018 December 2023 Notification of Age-Related Degradation Inspection (05000275/2024014 and 05000323/2024014) and Request for Information DCL-23-122, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-14014 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation DCL-23-128, Emergency Plan Update2023-12-13013 December 2023 Emergency Plan Update ML23355A0952023-12-0808 December 2023 License Renewal Application Online Reference Portal DCL-23-125, Core Operating Limits Report for Unit 1 Cycle 252023-12-0606 December 2023 Core Operating Limits Report for Unit 1 Cycle 25 ML23291A2702023-11-28028 November 2023 Letter to Nakia Zavalla, Shpo, the Santa Ynez Band of Chumash Indians; Re., Diablo Canyon ISFSI Initiating Section 106 Consultation ML23320A2442023-11-28028 November 2023 Letter to Kerri Vera, Director of Department of Environmental Protection, Tule River Tribe; Re., Diablo Canyon ISFSI Initiating Section 106 Consultation ML23325A1382023-11-27027 November 2023 ISFSI Tribal Letter to San Luis Obispo County Chumash Indians ML23325A1322023-11-27027 November 2023 ISFSI Tribal Letter to Salian Tribe of Montgomery, San Luis Obispo ML23325A1332023-11-27027 November 2023 ISFSI Tribal Letter to Yak Tityu Tityu Northern Chumash Indians ML23307A0062023-11-27027 November 2023 ISFSI Tribal Letter to Northern Chumash Tribal Council ML24003A7242023-11-27027 November 2023 Independent Safety Committee; Diab Lo Canyon Nuclear Power Plant Operations, Thirty-Third Annual Report on the Safety ML23325A1292023-11-27027 November 2023 ISFSI Tribal Letter to Coastal Band of Chumash ML23320A1502023-11-17017 November 2023 Individual Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, & Opportunity for Hearing (Exigent Circumstances) (EPID L-2023-LLA-0155) - LTR DCL-23-121, Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System2023-11-16016 November 2023 Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System ML23296A0982023-11-15015 November 2023 Notification and Request for Consultation Regarding Pacific Gas and Electric Diablo Canyon Independent Spent Fuel Storage Installation Material License Renewal Request (Docket Number: 72-26) DCL-23-120, License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System2023-11-14014 November 2023 License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System ML23293A1052023-11-14014 November 2023 Receipt and Availability of License Renewal Application IR 05000275/20230032023-11-13013 November 2023 Integrated Inspection Report 05000275/2023003 and 05000323/2023003 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV DCL-23-118, License Renewal Application2023-11-0707 November 2023 License Renewal Application ML23318A2102023-10-31031 October 2023 Independent Safety Committee; Thirty-Third Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations DCL-2023-520, Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP)2023-10-19019 October 2023 Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP) DCL-23-103, Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System2023-10-13013 October 2023 Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System IR 05000275/20240152023-10-10010 October 2023 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection (050002752024015 and 050003232024015) 2024-02-09
[Table view] Category:License-Operating (New/Renewal/Amendments) DKT 50
MONTHYEARML23012A2172023-02-0909 February 2023 Issuance of Amendment Nos. 244 and 245 Revision to Technical Specifications to Adopt TSTF-569, Revision 2, Revise Response Time Testing Definition ML21242A2162022-12-21021 December 2022 Issuance of Amendment Nos. 243 and 244 Technical Specifications and Revised License Conditions for the Permanently Defueled Condition (EPID: L-2020-LLA-0261 ML22326A1632022-11-21021 November 2022 Licensee Comment Email on Post-Shutdown Emergency Plan Amendment ML22187A0252022-11-16016 November 2022 Issuance of Amendment Nos. 242 and 243 Revision to Emergency Plan for Post-Shutdown and Permanently Defueled Condition ML22221A1682022-09-0606 September 2022 Issuance of Amendment Nos. 241 and 242 Revision to Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspection ML21160A1742021-09-0202 September 2021 Issuance of Amendment Nos. 239 and 240 Revising Technical Specification 3.2.1, Heat Flux Hot Channel Factor (Fq(Z)), to Implement Methodology from WCAP-17661, Revision 1 ML21188A3452021-07-0808 July 2021 Issuance of Amendment No. 238 Revision to TS 3.7.8, Auxiliary Saltwater (Asw) System (Emergency Circumstances) ML20218A2762020-09-11011 September 2020 Issuance of Amendment Nos. 237 and 239 to Relocate Technical Specification 5.3, Unit Staff Qualifications, to Updated Final Safety Analysis Report ML20235R6352020-08-31031 August 2020 Issuance of Amendment Nos. 236 and 238 Revision to Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Exigent Circumstances ML20232C6482020-08-17017 August 2020 Lochbaum Public Comment 1 on Exigent LAR to Modify TS 3.7.5, Auxiliary Feedwater System ML19196A3092019-08-21021 August 2019 Issuance of Amendment Nos. 233 and 235 Regarding Revision to Emergency Plan to Change Staffing and Extend Staff Augmentation Times for Emergency Response Organization Positions ML17212A3792017-09-25025 September 2017 Diablo Canyon Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 231 and 233 Re: Emergency Action Level Scheme Change (CAC Nos. MF8528 and MF8529) ML17012A2462017-04-27027 April 2017 Issuance of Amendments Revise Licensing Bases to Adopt Alternative Source Term ML17018A3412017-02-23023 February 2017 Issuance of Amendments Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection (LTOP) System, ML16330A6722017-01-0505 January 2017 Issuance of Amendment Nos. 228 and 230, Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation ML16139A0082016-12-21021 December 2016 Issuance of Amendment Nos. 227 and 229, Digital Replacement of the Process Protection System Portion of the Reactor Protection System and Engineered Safety Features Actuation System DCL-16-061, Supplement to License Amendment Request 15-03, Application of Alternative Source Term.2016-06-0909 June 2016 Supplement to License Amendment Request 15-03, Application of Alternative Source Term. DCL-16-061, Diablo Canyon, Units 1 and 2 - Supplement to License Amendment Request 15-03, Application of Alternative Source Term.2016-06-0909 June 2016 Diablo Canyon, Units 1 and 2 - Supplement to License Amendment Request 15-03, Application of Alternative Source Term. ML16117A2522016-05-19019 May 2016 Issuance of Amendment Nos. 226 and 228, Revise Minimum Flow Value Specified in Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ML16035A4412016-04-14014 April 2016 Issuance of Amendment Nos. 225 and 227, Adopt National Fire Protection Association NFPA 805, Performance-Based Standard for Fire Protection for LWR Generating Plants (2001 Edition) ML16055A3592016-03-0909 March 2016 Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies ML16032A0062016-02-19019 February 2016 Redacted, Issuance of Amendment Nos. 223 and 225, Revise UFSAR Section 15.4.2.2 to Incorporate Pressurizer Filling Analysis for Major Rupture of a Main Feedwater Pipe Accident Into Licensing Basis ML15029A1732016-01-0505 January 2016 ISFSI - Issuance of Amendment Nos. 222, 224, and 4, Authorize Use of Preemption Authority Granted Under Provisions of Section 161A of the Atomic Energy Act of 1954, as Amended (CAC MF2809, MF2810, and L24991) ML15281A1642015-12-0303 December 2015 Issuance of Amendment Nos. 221 and 223, Revise Updated Final Safety Analysis Report and Approval to Perform Fuel Assembly Structural Analyses That Considers Application of Leak Before Break ML15245A5422015-09-30030 September 2015 Issuance of Amendment Nos. 220 and 222, Request to Revise Operating License Condition Related to Cyber Security Plan Milestone H Full Implementation Date ML15204A2222015-08-27027 August 2015 Issuance of Amendment Nos. 219 and 221, Adopt Technical Specifications Task Force Traveler TSTF-432, Revision 1, Change in Technical Specification End States (WCAP-16294) ML15162A8822015-07-0101 July 2015 Issuance of Amendment Nos. 218 and 220, Revise Surveillance Requirements in TS 3.8.1, AC Sources, Related to Diesel Generator Testing ML14042A1932014-02-27027 February 2014 Issuance of Amendment Nos. 217 and 219, Editorial and Clarifying Changes to Technical Specification (TS) 3.7.10, Control Room Ventilation System (CRVS) and TS 5.6.5, Core Operating Limits Report (COLR) (TAC MF1959-MF1960) ML13178A0052013-09-0505 September 2013 Issuance of Amendment Nos. 216 and 218, Adoption of TSTF-421-A, Revision to RCP Flywheel Inspection Program (WCAP-15666), Using Consolidated Line Item Improvement Process ML12352A0672013-01-31031 January 2013 Issuance of Amendment Nos. 215 and 217, Revise Technical Specifications Related to Auxilary Feedwater System to Adopt TSTF-245, Rev. 1, TSTF-340, Rev. 3, TSTF-412, Rev. 3, and TSTF-439, Rev. 2 ML12345A3792013-01-0909 January 2013 Issuance of Amendment Nos. 214 and 216, Revise Final Safety Analysis Report Update Section 4.3.2.2 to Allow Use of Beacon Power Distribution Monitoring System ML12338A0202012-12-0404 December 2012 Issuance of Amendment Nos. 213 and 215, Revise Technical Specification 3.7.10, Control Room Ventilation System (Crvs), Required Action A.1 Completion Time (Emergency Circumstances) ML1207903382012-03-29029 March 2012 Issuance of Amendment Nos. 211 and 213, Revise Technical Specifications to Adopt TSTF-163, Rev. 2, Minimum Vs. Steady State Voltage and Frequency, and Final Safety Analysis Report Update ML1116402742011-07-15015 July 2011 Issuance of Amendment Nos. 210 and 212, License Amendment Request to Revise License Condition and Approve Cyber Security Plan and Associated Implementation Schedule ML1020100842011-01-24024 January 2011 Issuance of Amendment Nos. 209 and 211, Revise Technical Specification 3.4.15, RCS Leakage Detection Instrumentation, Operability Requirements and Actions ML1025304432010-09-29029 September 2010 Issuance of Amendment Nos. 208 and 210, Revise FSAR Update to Include Damping Values for Seismic Design/Analysis - Integrated Head Assembly ML1018901172010-07-30030 July 2010 Issuance of Amendment Nos. 207 and 209, Revise FSAR Update to Include Critical Damping Value for Structural Dynamic Qualification of Control Rod Drive Mechanism Pressure Housings ML0925405442009-09-17017 September 2009 Issuance of Amendment No. 208, Revise Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15 (Exigent Circumstances) ML0921203992009-08-19019 August 2009 Issuance of Amendment Nos. 206 and 207, Modify Technical Specifications to Adopt TSTF-511, Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26 ML0907701812009-04-29029 April 2009 License Amendment, Issuance of Amendment Nos. 205 and 206, Revise TS 3.3.1, Reactor Trip System (RTS) Instrumentation, to Delete High-Negative Rate Trip Function ML0906308082009-03-30030 March 2009 Issuance of Amendment Nos. 204 and 205, Revise TS 3.7.5, Auxiliary Feedwater (AFW) System, and TS 3.7.6, Condensate Storage Tank (CST) and Fire Water Storage Tank (Fwst) ML0900803872009-01-15015 January 2009 Issuance of Amendment Nos. 203 and 204, Changes to Technical Specification 5.5.16, Containment Leakage Rate Testing Program ML0834601732008-12-31031 December 2008 Issuance of Amendment No. 202 and 203 Increase the Completion Times for Required Actions Related to Technical Specifications 3.5.2, the ECCS, and 3.6.6, CS & CS (Tac MD7512 & MD7513) ML0832906092008-12-23023 December 2008 Issuance of Amendment Nos. 201 and 202, Adoption of Changes to Standard Technical Specifications Under Technical Specifications Task Force (TSTF)-448 ML0824102622008-10-30030 October 2008 Issuance of Amendment Nos. 200 and 201, Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk-Informed Initiative 5b) ML0732400062008-01-0808 January 2008 Units, 1 and 2, Issuance of Amendment Nos. 198 and 199, Revise Technical Specification (TS) 3.3.2, TS 5.5.9, and TS 5.6.10 to Support Replacement of Steam Generators ML0719102142007-07-11011 July 2007 License Amendments, Incorporated the Mitigation Strategies Required by Section B.5.b. of Commission Order EA-02-026 ML0713707312007-06-26026 June 2007 Issuance of License Amendment Nos. 197 and 198 TS 5.5.16, Containment Leakage Rate Testing Program, for Consistency with 10 CFR 50.55a(g)(4) ML0703500512007-04-17017 April 2007 Issuance of Amendment Nos. 195 and 196 TS 3.4.1 Reactor Coolant System Pressure, Temperature, & Flow Departure from Nucleate Boiling Limits & TS 5.6.5 Core Operating Limits Report. ML0704405032007-03-19019 March 2007 License Amendment, Deletion of License Condition Section 2.G in Facility Operating License (CLIIP Item) 2023-02-09
[Table view] Category:Safety Evaluation
MONTHYEARML23199A3122023-07-20020 July 2023 Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule ML22340A6752023-03-0606 March 2023 Approval of Certified Fuel Handler Training and Retraining Program ML23012A2172023-02-0909 February 2023 Issuance of Amendment Nos. 244 and 245 Revision to Technical Specifications to Adopt TSTF-569, Revision 2, Revise Response Time Testing Definition ML21242A2162022-12-21021 December 2022 Issuance of Amendment Nos. 243 and 244 Technical Specifications and Revised License Conditions for the Permanently Defueled Condition (EPID: L-2020-LLA-0261 ML22334A0822022-12-0101 December 2022 Correction to Issuance of Amendment Nos. 242 and 243 Revision to Emergency Plan for Post-Shutdown and Permanently Defueled Condition ML22187A0252022-11-16016 November 2022 Issuance of Amendment Nos. 242 and 243 Revision to Emergency Plan for Post-Shutdown and Permanently Defueled Condition ML22221A1682022-09-0606 September 2022 Issuance of Amendment Nos. 241 and 242 Revision to Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspection ML21160A1742021-09-0202 September 2021 Issuance of Amendment Nos. 239 and 240 Revising Technical Specification 3.2.1, Heat Flux Hot Channel Factor (Fq(Z)), to Implement Methodology from WCAP-17661, Revision 1 ML21188A3452021-07-0808 July 2021 Issuance of Amendment No. 238 Revision to TS 3.7.8, Auxiliary Saltwater (Asw) System (Emergency Circumstances) ML20249C8502020-10-22022 October 2020 Proposed Alternative NDE-RCS-SE-2R22 to the Requirements of the ASME Code ML20218A2762020-09-11011 September 2020 Issuance of Amendment Nos. 237 and 239 to Relocate Technical Specification 5.3, Unit Staff Qualifications, to Updated Final Safety Analysis Report ML20235R6352020-08-31031 August 2020 Issuance of Amendment Nos. 236 and 238 Revision to Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Exigent Circumstances ML19274B0822019-10-0707 October 2019 Correction to Issuance of Amendment Nos. 233 and 235 to Revise the Emergency Plan to Change Staffing and Extend Staff Augmentation Times for ERO Positions ML19196A3092019-08-21021 August 2019 Issuance of Amendment Nos. 233 and 235 Regarding Revision to Emergency Plan to Change Staffing and Extend Staff Augmentation Times for Emergency Response Organization Positions ML19176A0082019-06-28028 June 2019 Request for Relief from the Inservice Inspection Program Requirements of ASME Code, Section XI, for Steam Generator and Pressurizer Nozzle Inside Radius Sections ML19023A0262019-01-25025 January 2019 Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4 ML17263A0592017-10-0606 October 2017 Diablo Canyon Power Plant, Unit 2 - Relief Requests to Allow Use of Proposed Alternatives to Implement Requirements of ASME Code, for Reactor Pressure Vessel Welds for the Third ISI Interval (CAC Nos. MF9386, MF9387, MF9388, MF9389, and MF9390) ML17212A3792017-09-25025 September 2017 Diablo Canyon Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 231 and 233 Re: Emergency Action Level Scheme Change (CAC Nos. MF8528 and MF8529) ML17132A2312017-05-16016 May 2017 Diablo Canyon Power Plant, Units 1 and 2 - Correction Letter Regarding Amendment Nos. 230 and 232 to Adopt Alternative Source Term (CAC Nos. MF6399 and MF6400) ML17012A2462017-04-27027 April 2017 Issuance of Amendments Revise Licensing Bases to Adopt Alternative Source Term ML17037D2672017-03-0909 March 2017 Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternate Depth Sizing Criteria for the Fourth Inservice Inspection ML17018A3412017-02-23023 February 2017 Issuance of Amendments Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection (LTOP) System, ML16349A3862016-12-28028 December 2016 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML16139A0082016-12-21021 December 2016 Issuance of Amendment Nos. 227 and 229, Digital Replacement of the Process Protection System Portion of the Reactor Protection System and Engineered Safety Features Actuation System ML16207A3972016-07-27027 July 2016 Relief Request NDE-SIF-U1, Allow Use of Alternate Examination Volume Coverage Requirements, Class 2 Seal Injection Welds, for the Third 10-Year Inservice Inspection Interval ML16117A2522016-05-19019 May 2016 Issuance of Amendment Nos. 226 and 228, Revise Minimum Flow Value Specified in Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ML16035A4412016-04-14014 April 2016 Issuance of Amendment Nos. 225 and 227, Adopt National Fire Protection Association NFPA 805, Performance-Based Standard for Fire Protection for LWR Generating Plants (2001 Edition) ML16055A3592016-03-0909 March 2016 Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies ML16056A5132016-03-0707 March 2016 Relief Request FLIG-U1, Request to Extend Third 10-Year Inservice Inspection Interval for Reactor Vessel Stud Hole Ligament Examinations ML16055A2812016-02-26026 February 2016 Relief Request NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements, Third 10-Year Inservice Inspection Interval ML16032A0062016-02-19019 February 2016 Redacted, Issuance of Amendment Nos. 223 and 225, Revise UFSAR Section 15.4.2.2 to Incorporate Pressurizer Filling Analysis for Major Rupture of a Main Feedwater Pipe Accident Into Licensing Basis ML16013A3402016-01-27027 January 2016 Relief Request NDE-PNS-U2A, ASME Code Section XI Requirements to Allow Use of Alternate Exam Volume Coverage Requirements, for the Third 10-year Inservice Inspection Interval ML15037A0062016-01-0505 January 2016 ISFSI - Safety Evaluation, Encl. 4, Amendment Nos. 222, 224, and 4 - Authorize Use of Preemption Authority Granted Under Provisions of Section 161A of the Atomic Energy Act of 1954, as Amended (CAC MF2809, MF2810, and L24991) ML15029A1732016-01-0505 January 2016 ISFSI - Issuance of Amendment Nos. 222, 224, and 4, Authorize Use of Preemption Authority Granted Under Provisions of Section 161A of the Atomic Energy Act of 1954, as Amended (CAC MF2809, MF2810, and L24991) ML15337A0212015-12-0808 December 2015 Relief Request NDE-SG-MS-IR - Relief from Exam Requirements for Main Steam Nozzles in Replacement Steam Generators, Third 10-Year Inservice Inspection Interval ML15281A1642015-12-0303 December 2015 Issuance of Amendment Nos. 221 and 223, Revise Updated Final Safety Analysis Report and Approval to Perform Fuel Assembly Structural Analyses That Considers Application of Leak Before Break ML15299A0342015-11-0404 November 2015 Relief Request NDE-RCS-SE-2R19, Use of Alternate Sizing Qualification Criteria Through a Protective Clad Layer, Third 10-Year Inservice Inspection Interval ML15245A5422015-09-30030 September 2015 Issuance of Amendment Nos. 220 and 222, Request to Revise Operating License Condition Related to Cyber Security Plan Milestone H Full Implementation Date ML15261A0072015-09-18018 September 2015 Request for Relief PRS-3, Proposed Alternative to Requirements for System Leakage Testing Class 1 Piping Vent, Drain, and Test Isolation Devices, for the Third 10-year Inservice Inspection Interval (TAC MF6520-MF6521) ML15204A2222015-08-27027 August 2015 Issuance of Amendment Nos. 219 and 221, Adopt Technical Specifications Task Force Traveler TSTF-432, Revision 1, Change in Technical Specification End States (WCAP-16294) ML15187A0352015-07-15015 July 2015 Relief Request REP-SI, Revision 2, Proposed Alternative for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments, Remaining Useful Life +20 Years ML15162A8822015-07-0101 July 2015 Issuance of Amendment Nos. 218 and 220, Revise Surveillance Requirements in TS 3.8.1, AC Sources, Related to Diesel Generator Testing ML15168A0242015-06-19019 June 2015 Request for Alternative RPV-U1 - Extension to Allow Use of Alternate Reactor Inspection Interval Requirements for the Third 10-Year Inservice Inspection Interval ML14251A4682014-09-16016 September 2014 Relief Request NDE-SFW-13, ASME Code Section XI, Request for Alternative to Allow Use of Alternate Examination Requirements, Third 10-Year Inservice Inspection Interval ML14196A0952014-08-0404 August 2014 Request for Relief from ASME Code Requirements for Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary ML14104B6132014-05-14014 May 2014 Relief Request NDE-PNS-U2, Requirements for Examination Coverage of ASME Class 1 Pressurizer Nozzle-to-Head Welds, Third 10-Year Inservice Inspection Interval ML14042A1932014-02-27027 February 2014 Issuance of Amendment Nos. 217 and 219, Editorial and Clarifying Changes to Technical Specification (TS) 3.7.10, Control Room Ventilation System (CRVS) and TS 5.6.5, Core Operating Limits Report (COLR) (TAC MF1959-MF1960) ML13192A3542013-09-12012 September 2013 Relief Request RVFLNG-INT3 - Alternative to ASME Code Pressure Test Requirements for Class 1 Reactor Vessel Flange Leakoff Lines, Third 10-Year Inservice Inspection Interval (TAC MF0408-MF0409) ML13178A0052013-09-0505 September 2013 Issuance of Amendment Nos. 216 and 218, Adoption of TSTF-421-A, Revision to RCP Flywheel Inspection Program (WCAP-15666), Using Consolidated Line Item Improvement Process ML13232A3082013-08-28028 August 2013 Relief Request REP-1 U2, Revision 2, Alternative to Code Requirements for Pressurizer Structural Weld Overlays, for the Third 10-Year Inservice Inspection Interval 2023-07-20
[Table view] Category:Technical Specifications
MONTHYEARML23012A2172023-02-0909 February 2023 Issuance of Amendment Nos. 244 and 245 Revision to Technical Specifications to Adopt TSTF-569, Revision 2, Revise Response Time Testing Definition ML21242A2162022-12-21021 December 2022 Issuance of Amendment Nos. 243 and 244 Technical Specifications and Revised License Conditions for the Permanently Defueled Condition (EPID: L-2020-LLA-0261 DCL-22-059, License Amendment Request 22-03 Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions2022-07-26026 July 2022 License Amendment Request 22-03 Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions ML21188A3452021-07-0808 July 2021 Issuance of Amendment No. 238 Revision to TS 3.7.8, Auxiliary Saltwater (Asw) System (Emergency Circumstances) ML21188A0372021-07-0707 July 2021 Attachment 1 TS_3_7_8 (Marked Up Pages) ML21188A0352021-07-0707 July 2021 Attachment 2 TS_3_7_8 (Retype Pages) ML21188A0342021-07-0707 July 2021 Attachment 3-TS Bases Section 3_7_8 (Marked-up Pages) DCL-19-094, License Amendment Request 19-02 to Relocate Technical Specification 5.3, Unit Staff Qualifications, to the Updated Final Safety Analysis Report2019-12-0505 December 2019 License Amendment Request 19-02 to Relocate Technical Specification 5.3, Unit Staff Qualifications, to the Updated Final Safety Analysis Report DCL-18-100, License Amendment Request 18-02 - License Amendment Request to Revise Technical Specification 5.6.5b1 Core Operating Limits Report (COLR) for Full Spectrum Loss-of-Coolant Accident Methodology2018-12-26026 December 2018 License Amendment Request 18-02 - License Amendment Request to Revise Technical Specification 5.6.5b1 Core Operating Limits Report (COLR) for Full Spectrum Loss-of-Coolant Accident Methodology ML16330A6722017-01-0505 January 2017 Issuance of Amendment Nos. 228 and 230, Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation ML16139A0082016-12-21021 December 2016 Issuance of Amendment Nos. 227 and 229, Digital Replacement of the Process Protection System Portion of the Reactor Protection System and Engineered Safety Features Actuation System ML16035A4412016-04-14014 April 2016 Issuance of Amendment Nos. 225 and 227, Adopt National Fire Protection Association NFPA 805, Performance-Based Standard for Fire Protection for LWR Generating Plants (2001 Edition) ML16083A5642016-03-23023 March 2016 License Amendment Request 16-02, License Amendment Request to Revise Technical Specification 3.4.12., Low Temperature Overpressure Protection (Ltop) System. DCL-16-028, Technical Specifications B.4.17., Steam Generator (SG) Tube Integrity, LTOP Orifice-Key Design Features, and Additional Information2016-03-23023 March 2016 Technical Specifications B.4.17., Steam Generator (SG) Tube Integrity, LTOP Orifice-Key Design Features, and Additional Information ML16055A3592016-03-0909 March 2016 Issuance of Amendment Nos. 224 and 226, Revise Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and TS 5.6.5 (COLR) Paragon and Nexus Methodologies ML16032A0062016-02-19019 February 2016 Redacted, Issuance of Amendment Nos. 223 and 225, Revise UFSAR Section 15.4.2.2 to Incorporate Pressurizer Filling Analysis for Major Rupture of a Main Feedwater Pipe Accident Into Licensing Basis ML15245A5422015-09-30030 September 2015 Issuance of Amendment Nos. 220 and 222, Request to Revise Operating License Condition Related to Cyber Security Plan Milestone H Full Implementation Date ML15204A2222015-08-27027 August 2015 Issuance of Amendment Nos. 219 and 221, Adopt Technical Specifications Task Force Traveler TSTF-432, Revision 1, Change in Technical Specification End States (WCAP-16294) ML15162A8822015-07-0101 July 2015 Issuance of Amendment Nos. 218 and 220, Revise Surveillance Requirements in TS 3.8.1, AC Sources, Related to Diesel Generator Testing DCL-15-052, Response to NRC Request for Additional Information Regarding License Amendment Request 14-01, Revision to Technical Specification 3.8.1, AC Sources - Operating.2015-04-29029 April 2015 Response to NRC Request for Additional Information Regarding License Amendment Request 14-01, Revision to Technical Specification 3.8.1, AC Sources - Operating. ML14042A1932014-02-27027 February 2014 Issuance of Amendment Nos. 217 and 219, Editorial and Clarifying Changes to Technical Specification (TS) 3.7.10, Control Room Ventilation System (CRVS) and TS 5.6.5, Core Operating Limits Report (COLR) (TAC MF1959-MF1960) ML13130A2252013-05-10010 May 2013 Technical Specification 3.5.2 Safety Injection Pump 2-1 Issue Information ML13267A1352013-01-31031 January 2013 Attachment 6 to DCL-13-016, LTR-RAM-1-13-002 NP-Attachment, Revision 0, Justification for the Application of Technical Specifications Changes in WCAP-14333 and WCAP-15376 to the Tricon/Als Process Protection System at the Diablo Canyon Powe DCL-12-107, Supplement to PG&E Letter DCL-11-059, License Amendment Request 11-04, Revision to Technical Specification (TS) 3.6.6, 'Containment Spray and Cooling Systems,' TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' TS 3.8.1, 'Ac.....2012-11-0101 November 2012 Supplement to PG&E Letter DCL-11-059, License Amendment Request 11-04, Revision to Technical Specification (TS) 3.6.6, 'Containment Spray and Cooling Systems,' TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' TS 3.8.1, 'Ac..... DCL-12-057, Supplement to PG&E Letter DCL-12-017, Response to NRC Request for Additional Information: NRC Question 12012-05-31031 May 2012 Supplement to PG&E Letter DCL-12-017, Response to NRC Request for Additional Information: NRC Question 1 DCL-11-097, License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake2011-10-20020 October 2011 License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake DCL-11-059, Diablo Canyon Units 1 & 2, License Amendment Request 11-04, Revision to Technical Specification 3.6.6, Containment Spray and Cooling Systems, TS 3.7.5, Auxiliary Feedwater System, TS 3.8.1, AC Sources-Operating, TS 3.8.9, Distribution Sys2011-06-0101 June 2011 Diablo Canyon Units 1 & 2, License Amendment Request 11-04, Revision to Technical Specification 3.6.6, Containment Spray and Cooling Systems, TS 3.7.5, Auxiliary Feedwater System, TS 3.8.1, AC Sources-Operating, TS 3.8.9, Distribution Syste DCL-11-055, Supplement to License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs).2011-04-21021 April 2011 Supplement to License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs). DCL-11-018, License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs)2011-02-17017 February 2011 License Amendment Request 11-02, Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) DCL-10-028, Diablo Canyon, Units 1 and 2, Emergency License Amendment Request 10-02, One Time Revision to Technical Specification 3.8.1, AC Sources - Operating.2010-03-11011 March 2010 Diablo Canyon, Units 1 and 2, Emergency License Amendment Request 10-02, One Time Revision to Technical Specification 3.8.1, AC Sources - Operating. DCL-10-028, Emergency License Amendment Request 10-02, One Time Revision to Technical Specification 3.8.1, AC Sources - Operating.2010-03-11011 March 2010 Emergency License Amendment Request 10-02, One Time Revision to Technical Specification 3.8.1, AC Sources - Operating. DCL-10-018, License Amendment Request 10-01, Revision to Technical Specification 3.8.1, AC Sources - Operating.2010-02-24024 February 2010 License Amendment Request 10-01, Revision to Technical Specification 3.8.1, AC Sources - Operating. DCL-09-062, Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15.2009-09-0303 September 2009 Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15. ML0919601772009-07-0303 July 2009 License Amendment Request 09-02, Revision to Technical Specification 3.4.15, RCS Leakage Detection Instrumentation, Operability Requirements and Actions for RCS Leakage Detection Instrumentation DCL-09-009, License Amendment Request 09-01 - Revision to Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, Deletion of the High Negative Rate Trip Function2009-02-24024 February 2009 License Amendment Request 09-01 - Revision to Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, Deletion of the High Negative Rate Trip Function DCL-08-100, Supplement to License Amendment Request 07-03, Revision to Technical Specification 3.7.10, Control Room Ventilation System (CRVS)2008-11-25025 November 2008 Supplement to License Amendment Request 07-03, Revision to Technical Specification 3.7.10, Control Room Ventilation System (CRVS) ML0824102622008-10-30030 October 2008 Issuance of Amendment Nos. 200 and 201, Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk-Informed Initiative 5b) DCL-07-097, License Amendment Request 07-04, Proposed Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk Informed Technical Specifications Initiative 5b)2007-10-15015 October 2007 License Amendment Request 07-04, Proposed Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk Informed Technical Specifications Initiative 5b) DCL-07-075, Supplement to License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement, and Response to Request for Additional Information2007-08-0909 August 2007 Supplement to License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement, and Response to Request for Additional Information ML0713707332007-06-26026 June 2007 Tech Spec Pages for Amendment Nos. 197 and 198 TS 5.5.16, Containment Leakage Rate Testing Program, for Consistency with 10 CFR 50.55a(g)(4) ML0709900742007-06-25025 June 2007 Units, 1 and 2, Tech Spec Page for Amendments 196 and 197 Administrative Changes to the Technical Specification 5.5.8, Inservice Testing Program. ML0703500522007-04-17017 April 2007 Tech Spec Pages for Amendment Nos. 195 and 196 TS 3.4.1 Reactor Coolant System Pressure, Temperature, & Flow Departure from Nucleate Boiling Limits & TS 5.6.5 Core Operating Limits Report. ML0704404942007-03-21021 March 2007 Tech Spec Pages for Amendments 194 and 195 Regarding Steam Generator Tube Integrity Technical Specifications ML0701902142007-01-19019 January 2007 Units, 1 and 2 - Tech Spec Pages for Amendments 192 and 193 Technical Specification 1.1 Definitions and Technical Specification 3.4.16 Reactor Coolant System Specific Activity DCL-06-143, License Amendment Request 06-10 Revision to Technical Specification 5.5.8, Inservice Testing Program2006-12-29029 December 2006 License Amendment Request 06-10 Revision to Technical Specification 5.5.8, Inservice Testing Program DCL-06-136, License Amendment Request 06-05 Revision to Technical Specification 3.4.1, Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits, and Technical Specification 5.6.5, Core Operating....2006-12-29029 December 2006 License Amendment Request 06-05 Revision to Technical Specification 3.4.1, Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits, and Technical Specification 5.6.5, Core Operating.... DCL-06-142, License Amendment Request 06-09, Revision to Technical Specification 5.5.16, Containment Leakage Rate Testing Program (TSTF-343)2006-12-29029 December 2006 License Amendment Request 06-09, Revision to Technical Specification 5.5.16, Containment Leakage Rate Testing Program (TSTF-343) ML0635500312006-12-20020 December 2006 Tech Spec Pages for Amendment 192 Regarding TS 5.6.5, Core Operating Limits Report. DCL-06-129, Special Report 06-03 - 180 Day Report for Steam Generator Technical Specification Task Force 449, Thirteenth Refueling Outage2006-11-17017 November 2006 Special Report 06-03 - 180 Day Report for Steam Generator Technical Specification Task Force 449, Thirteenth Refueling Outage ML0631905402006-11-15015 November 2006 Technical Specifications, Issuance of Exigent Amendment No. 190 Revision to TS Section 3.8.4, DC Sources - Operating to Extend the Completion Time to Restore an Inoperable Station Battery 2023-02-09
[Table view] |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2016 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REVISION TO UPDATED FINAL SAFETY ANALYSIS REPORT FOR BEST ESTIMATE ANALYZER FOR THE CORE OPERATIONS-NUCLEAR POWER DISTRIBUTION MONITORING SYSTEM METHODOLOGY AND TECHNICAL SPECIFICATION 5.6.5 (CAC NOS. MF6120 AND MF6121)
Dear Mr. Halpin:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 224 to Facility Operating License No. DPR-80 and Amendment No. 226 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TSs) in response to your application dated April 16, 2015.
The amendments revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the DCPP UFSAR Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System." The amendments also revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with NRG-approved proprietary TR WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."
E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely, S.vv-o._ cy ' "~
Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323
Enclosures:
- 1. Amendment No. 224 to DPR-80
- 2. Amendment No. 226 to DPR-82
- 3. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. DPR-80
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 21. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. DPR-80 and Technical Specifications Date of Issuance: March 9 , 2O1 6
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT No. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 226 License No. DPR-82
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Pacific Gas and Electric Company (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 regulations 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; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:
(2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented prior to MODE 4 at the start of Cycle 20. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensee's letter dated April 16, 2015.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. DPR-82 and Technical Specifications Date of Issuance: March 9 , 2o1 6
ATTACHMENT TO LICENSE AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of the Facility Operating License Nos. DPR-80 and DPR-82, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 5.0-20 5.0-20 5.0-21 5.0-21
(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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 224 are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3) Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
- a. Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential; Amendment No. 224
(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; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.
(2) Technical Specifications (SSER 32. Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 226, are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3) Initial Test Program (SSER 31. Section 4.4.1)
Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment No. 226
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control F0 Surveillance Technical Specification, (Westinghouse Proprietary),
- 2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (Westinghouse Proprietary),
- 3. WCAP-8385, Power Distribution Control and Load Following Procedures, (Westinghouse Proprietary),
- 4. WCAP-10054-P-A, Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse Proprietary),
- 5. WCAP-10054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model, July 1997 (Westinghouse Proprietary),
- 6. WCAP-12945-P-A, Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis, June 1996 (Westinghouse Proprietary),
- 7. WCAP-12945-P-A, Addendum 1-A, Revision 0, "Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best Estimate LOCA Evaluation Models," December 2004. (Westinghouse Proprietary) (Unit 1 Only),
- 8. WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005. (Westinghouse Proprietary)
(Unit 2 Only),
- 9. WCAP-8567-P-A, "Improved Thermal Design Procedure,"
- 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and
- 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."
(continued)
DIABLO CANYON - UNITS 1 & 2 5.0-20 Unit 1 - Amendment No. ~.~,+94-,4-%,+98, 224 Unit 2 - Amendment No. ~.~.~.4-96,+99, 226
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Protection (LTOP) arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
- 1. Specification 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and
- 2. Specification 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System."
(continued)
DIABLO CANYON - UNITS 1 & 2 5.0-21 Unit 1 - Amendment No. ~. 498, 224 Unit 2 - Amendment No. ~ * .+w, 226
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By letter dated April 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15107A333), Pacific Gas and Electric Company (the licensee) submitted a license amendment request (LAR) to revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) power distribution monitoring system methodology described in the Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.2, "Power Distribution," to the method described in the Westinghouse Electric Company LLC (Westinghouse) proprietary topical report (TR) WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System," September 2012 (ADAMS Accession No. ML12270A386 for non-proprietary version) for Diablo Canyon Power Plant (DCPP), Units 1 and 2. This LAR also proposes to revise Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR),"
Section b to replace Westinghouse proprietary TR WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988, with U.S. Nuclear Regulatory Commission (NRC)-approved proprietary TR WCAP-16045-P-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 (ADAMS Accession No. ML042250322 for non-proprietary version), and NRG-approved proprietary TR WCAP-16045-P-A, Addendum 1-A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," August 2007 (ADAMS Accession No. ML072570329 for non-proprietary version).
2.0 REGULATORY EVALUATION
The regulatory requirements and criteria specifically applicable to the proposed license amendment are summarized and addressed below:
NRC Generic Letter 88-16 The guidance in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988 (ADAMS Accession No. ML031130447),
indicates that it is acceptable for licensees to control reactor physics parameter limits by Enclosure 3
specifying the calculation methodology. The generic letter indicates that such parameter limits may be removed from TSs and placed in a cycle-specific COLR. The COLR is defined in the TSs, and the Reporting Requirements in the TSs require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised. Generic Letter 88-16 also recommends that licensees include, in the TSs, a list of references for the NRG-approved methodologies, which are used to generate the cycle-specific parameter operating limits.
10 CFR 50.34 Section 50.34, "Contents of applications; technical information," of Title 10 to the Code of Federal Regulations ( 10 CFR) Part 50 requires that safety analysis reports include analysis and evaluation of the design and performance of structures, systems, and components for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload design process, licensees (or vendors) perform reload safety evaluations to ensure that their safety analyses remain bounding for the design cycle.
To confirm that the analyses remain bounding, licensees confirm that key inputs to the safety analyses (such as the critical power ratio) are conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a reanalysis or reevaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.
10 CFR 50.36 In Section 50.36, "Technical specifications," of 10 CFR Part 50, the Commission established its regulatory requirements related to the content of the TSs. Consistent with 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation:
- 1) safety limits, limiting safety system settings (LSSS), and limiting control settings; 2) limiting conditions for operations (LCOs); 3) surveillance requirements; 4) design features; and
- 5) administrative controls.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes Current TS 5.6.5 Section b states, in part:
- 10. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."
Revised TS 5.6.5 Section b would state, in part:
- 10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and
- 11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."
3.2 Technical Evaluation of the Application of the Revised BEACON Methodology The NRG-approved BEACON methodology described in Westinghouse TR WCAP-12472-P-A Addendum 4, "BEACON Core Monitoring and Operation Support System, Addendum 4", has been reviewed for application to DCPP, Units 1 and 2, as summarized below.
The BEACON core monitoring system currently in use at DCPP, Units 1 and 2, WCAP-12472-NP-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," January 2000 (ADAMS Accession No. ML003678155), employs a predictive three-dimensional nodal model of the reactor core calibrated to an in-core flux map, which is periodically obtained utilizing the movable in-core detector system {MIDS). The resultant nodal power distribution is frequently updated as part of the core follow calculation, and is adjusted for the axial offset measured by the ex-core detectors; the radial distribution of power is further corrected using a calorimetric process based on core exit thermocouple measurements.
Updating of the power distribution based on ex-core detector and core exit thermocouple readings is performed on a nearly continuous basis. A reasonably accurate representation of the core power distribution is determined through use of the MIDS. The flux traces obtained with MIDS are used to calibrate the BEACON predictive model, and also to calibrate both the ex-core detector response, and the core exit thermal couple readings. Between MIDS in-core flux scans, the ex-core and core exit thermocouple calibration factors are utilized in the BEACON power distribution calculation.
A key factor in the BEACON methodology is the uncertainties applied to the BEACON measured power distributions. One component of the measurement uncertainty is the variability of the thermocouple calibration factors (also referred to as mixing factors). The mixing factor uncertainty (standard deviation) is shown to exhibit a power level dependency due to the separate effects of decreased temperature difference (lff RPL) at lower power levels and changes in cross-flow patterns at reduced power levels.
In the current methodology, the mixing factor uncertainty is determined as a function of power level by accounting for the effects of decreased LHRPL and changes in cross-flow patterns separately. Additionally, an average thermocouple standard deviation at hot full-power conditions is determined from past performance of thermocouples from previous cycles and utilized to account for measurement uncertainty.
The updated methodology described in WCAP-12472-P-A, Addendum 4, utilizes core exit thermocouple data collected during initial power ascension following a refueling outage. This process simultaneously considers the effects of both LlT RPL and cross-flow patterns on mixing factor uncertainty, and eliminates the reliance on previous cycle thermocouple data, thus resulting in a more accurate and reliable evaluation of thermocouple uncertainties. Additionally, the use of current cycle data allows for the inclusion of any changes to the thermocouples since the previous cycle. These changes include the failure or replacement of individual thermocouples, relocation of a thermocouple or instrumentation system changes.
The NRC Safety Evaluation Report (SER) dated August 9, 2012, approving WCAP-12472-P-A, Addendum 4 (ADAMS Accession No. ML12150A147), states that the updated BEACON thermocouple methodology is acceptable. The NRC staff's review of the applicant's LAR shows that the licensee has considered the specific application of the NRG-approved methodology to
DCPP, Units 1 and 2, including the plant-specific configuration of the core exit thermocouple design and treatment of thermocouple failures.
In the NRC staff's audit report dated February 23, 2016 (ADAMS Accession No. ML16053A323), the staff noted that the licensee stated and confirmed that WCAP-12472-P-A, Addendum 4 may replace WCAP-12472-P-A, Addendum 1-A without loss of applicable licensing basis.
The proposed change in this LAR does not impact DCPP's compliance with the NRC SER conditions associated with the approval of the original BEACON WCAP-12472 and WCAP-12472-P-A, Addendum 1-A. Therefore, the three licensing conditions cited in the NRC SER documenting the staff's approval of the original BEACON topical report WCAP-12472-P-A, Addendum 1-A remains applicable with the implementation of Addendum 4 of WCAP-12472-P-A.
Based on the above, the NRC staff finds that the methodology described in WCAP-12472-NP-A, Addendum 4 may be implemented at DCPP, Units 1 and 2 as proposed by the licensee.
3.3 Technical Evaluation of the Application of WCAP-16045-P-A and addendum 3.3.1 PARAGON and NEXUS Code Descriptions The PARAGON code, described in WCAP-16045-P-A, is a neutron transport code, intended to replace PHOENIX-P for use in providing nuclear input data for the Westinghouse proprietary Advanced Nodal Code (ANC), which is a core simulator code. NEXUS (WCAP-16045-NP-A, Addendum 1A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology,"
August 2007) is an improvement to the PARAGON code system.
WCAP-16045-NP-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 and WCAP-16045-NP-A, Addendum 1-A are NRG-approved, based on extensive validation and verification. As described in the topical report and addendum, the data set used for validation and verification includes numerous criticality experiments and plant data. The plant data reflect a widely varied set of operating conditions that includes numerous fuel designs, lattice geometries, and burnable absorber loadings in use in the Westinghouse- and Combustion Engineering-designed nuclear steam supply systems. The verification was expanded to include comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology. The results showed excellent agreement between NEXUS predictions and available data.
3.3.2 Applicability to DCPP, Units 1 and 2 The NRC staff reviewed information contained in the UFSAR to verify that the database discussed in Section 3.3.1 of this SE includes fuel with similar characteristics to that used at DCPP. Chapter 4, "Reactor," of the UFSAR indicates that DCPP, Units 1 and 2, use 17x17 Westinghouse optimized fuel. This design feature is represented in the database described above; therefore, the NRC staff determined that the NEXUS code system is acceptable for use at DCPP. The only condition or limitation established in the SERs approving PARAGON and NEXUS precludes their use for mixed-oxide (MOX) cores. The NRC staff verified and
documented in its audit report dated February 23, 2016, that DCPP does not use MOX fuel in the core reload design.
The NRC staff determined, therefore, that a separate licensing action would be required in order to use MOX fuel at DCPP, so the MOX limitation in the SERs approving PARAGON and NEXUS is satisfied.
3.4 Additional Review Topics The addendum to WCAP-16045-P-A brings about a change to the way boron letdown curves are calculated and provide input into the overall nuclear design method. Because of this, in previous reviews of PARAGON-NEXUS/ANG applications, the NRC staff has found it necessary to verify that no changes were made to the analysis methods for post loss-of-coolant accident (LOCA) subcriticality and boric acid precipitation behavior.
The PARAGON code, as described in WCAP-16045-P-A, is a neutron transport code intended to replace PHOENIX-P for use in providing cross-section data to the Westinghouse proprietary ANC, a core simulator code. NEXUS, described in WCAP-16045-P-A, Addendum 1-A, is an improvement to the PARAGON/ANG code system that changes the method of communicating the nuclear data output from PARAGON to ANC. Rather than using boron letdown curves, NEXUS accounts for variations in the neutron spectrum by parameterizing the PARAGON cross-section output and reconstructing it within ANC.
WCAP-16045-P-A and addendum are NRG-approved, and the NRC staff has previously determined that they have been extensively validated and verified using a variety of experiments and plant data that account for a wide range of differences in operating conditions.
In Addendum 1-A, the verification was expanded to include the comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology, showing excellent agreement between NEXUS predictions and available data.
The use of WCAP-16045-NP-A, Addendum 1, for DCPP does not affect the inputs or method(s) for ensuring core subcriticality, both short- and long-term post-LOCA, thereby precluding the potential for return to power following a large break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the use of Addendum 1, the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection will continue to remain valid. Core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs and refueling boron concentration, will continue to be calculated using NRC approved methods.
The NRC staff concludes this is acceptable, because it confirms that the implementation of PARAGON and NEXUS at DCPP will not affect the calculation of post-LOCA boron requirements or emergency procedures to mitigate post-LOCA boric acid precipitation. Based on the above, the staff concludes that the existing post-LOCA analyses for subcriticality and long-term cooling remain applicable.
3.5 Summary The NRC staff has determined that the proposed changes in the UFSAR to the BEACON core monitoring system methodology described in WCAP-12472-P-A, Addendum 4 and the TS changes required to replace the PHOENIX-P code system with the PARAGON/NEXUS code system are acceptable for DCPP. This determination is based on 1) that the original BEACON TR WCAP-12472-P-A remains applicable, 2) that PARAGON and NEXUS are NRG-approved methods, which have been determined to be applicable to DCPP, and 3) that the DCPP post-LOCA subcriticality and boric acid precipitation analyses will be unaffected by the proposed changes. Based on the above, the staff concludes that the proposed TS revision is consistent with the guidance provided in GL 88-16, and that the generic qualification and validation of the PARAGON-NEXUS/ANG system satisfies the requirements of 10 CFR 50.34 and 10 CFR 50.36.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 published in the Federal Register on June 9, 2015 (80 FR 32628). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need 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: F. Forsaty, NRR Date: March 9, 2016
ML16055A359 *SE via memo OFFICE NRR/DORULPL4-1 /PM NRR/DORULPL4-1 /LA NRR/DSS/SRXB/BC (A)* NRR/DSS/SNPB/BC NAME SLingam JBurkhardt CJackson (EOesterle for) JDean DATE 2/26/16 2/25/16 2/18/16 3/7/16 OFFICE NRR/DSS/STSB/BC OGC - NLO w/comments NRR/DORULPL4-1/BC N RR/DORULPL4-1 /PM NAME RElliott JLindell RPascarelli (FLyon for) SLingam DATE 3/1/16 3/4/16 3/9/16 3/9/16