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MONTHYEARRA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. Project stage: Request RA-16-0003, Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report...2016-02-0303 February 2016 Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report... Project stage: Request RA-16-0020, Withdrawal of License Amendment Request Regarding Methodology Reports DPC-NF-2010 and DPC-NE-2011-P2016-04-0707 April 2016 Withdrawal of License Amendment Request Regarding Methodology Reports DPC-NF-2010 and DPC-NE-2011-P Project stage: Withdrawal RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P Project stage: Supplement 2016-02-03
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H.B. Robinson, Unit 2 - Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-PML16125A420 |
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Site: |
Harris, Robinson |
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Issue date: |
05/04/2016 |
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From: |
Elnitsky J Duke Energy Progress |
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To: |
Document Control Desk, Office of Nuclear Reactor Regulation |
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Shared Package |
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ML16125A444 |
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References |
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RA-16-0023, TAC MF7337, TAC MF7338 DPC-NE-1008, Rev. 0, DPC-NE-2011, Rev. 2, DPC-NF-2010, Rev. 3 |
Download: ML16125A420 (338) |
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Category:Legal-Affidavit
MONTHYEARRA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-22-0290, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology2023-08-30030 August 2023 License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology RA-23-0043, Refuel 33 (R2R33) Inservice Inspection Program Ninety Day Owner'S Activity Report and Analytical Evaluations2023-03-30030 March 2023 Refuel 33 (R2R33) Inservice Inspection Program Ninety Day Owner'S Activity Report and Analytical Evaluations RA-22-0210, Supplement to License Amendment Request to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-07-28028 July 2022 Supplement to License Amendment Request to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-20-0382, 30-Day Report Pursuant to 10 CFR 50.46, Changes to or Errors in an Acceptable Loss of Coolant Evaluation Model2020-12-17017 December 2020 30-Day Report Pursuant to 10 CFR 50.46, Changes to or Errors in an Acceptable Loss of Coolant Evaluation Model RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in . RA-19-0453, Response to Request for Additional Information Regarding License Amendment Request to Modify Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type2019-12-20020 December 2019 Response to Request for Additional Information Regarding License Amendment Request to Modify Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type RA-19-0138, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections2019-07-23023 July 2019 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections RA-19-0017, License Amendment Request to Modify the Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type2019-04-10010 April 2019 License Amendment Request to Modify the Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type RA-18-0185, Response to Request for Additional Information (RAI) Regarding 10 CFR 50.46 Annual Report, Including Revised Robinson Large Break Loss of Coolant Accident Report2018-12-10010 December 2018 Response to Request for Additional Information (RAI) Regarding 10 CFR 50.46 Annual Report, Including Revised Robinson Large Break Loss of Coolant Accident Report RA-18-0016, Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses2018-06-0505 June 2018 Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses HNP-18-045, Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval2018-04-18018 April 2018 Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval RA-17-0048, Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2)2017-10-30030 October 2017 Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2) RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P ML16112A2752016-04-19019 April 2016 Robinson Fuel Meth Pre-sub Meeting - Presentation Affidavit for Closed Pre-Submittal Meeting with Duke Energy Progress to Discuss Fuel Reload Design Methodology Reports and Proposed LAR Re H.B. Robinson and Shearon Harris Plants HNP-15-038, License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change2015-12-17017 December 2015 License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0041, Attachment 1, Affidavit and Attachment 3, Pre-Submittal Meeting Presentation Materials on DPC-NF-2010-A and DPC-NE-2011-P Methodologies (Redacted)2015-09-18018 September 2015 Attachment 1, Affidavit and Attachment 3, Pre-Submittal Meeting Presentation Materials on DPC-NF-2010-A and DPC-NE-2011-P Methodologies (Redacted) RA-15-0037, Transmittal of Response to NRC Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 52015-09-0909 September 2015 Transmittal of Response to NRC Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5 RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RA-15-0004, Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology2015-03-0505 March 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology RNP-RA/13-0072, Request for Site-Specific Information Related to Use of Gothic Be Withheld from Public Disclosure in Accordance with 10 CFR 2.390(a)(4)2013-08-0808 August 2013 Request for Site-Specific Information Related to Use of Gothic Be Withheld from Public Disclosure in Accordance with 10 CFR 2.390(a)(4) ML12073A2922012-02-24024 February 2012 License Amendment Request for Revision to Technical Specification Core Operating Limits Report for Realistic Large Break LOCA Analysis RA-11-008, Progress Energy - Evidence of Guarantee of Payment of Deferred Premiums2011-04-14014 April 2011 Progress Energy - Evidence of Guarantee of Payment of Deferred Premiums RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation RNP-RA/07-0126, Request for Technical Specifications Change to Section 3.6.8 Isolation Valve Seal Water System2007-11-29029 November 2007 Request for Technical Specifications Change to Section 3.6.8 Isolation Valve Seal Water System ML0721203582007-07-20020 July 2007 Supplemental Declarations of Robert L. Smith, Judy Hogan, Worth Glover, Jr., Tony A. Hackney, Elbert Green, Derry J. Smith, Jr. and Beverly Ann D'Aquanni RNP-RA/05-0062, Response to Request for Additional Information Regarding Technical Specifications Change Request to Section 3.8.4 DC Sources - Operating2005-07-13013 July 2005 Response to Request for Additional Information Regarding Technical Specifications Change Request to Section 3.8.4 DC Sources - Operating RNP-RA/05-0039, Response to NRC Request for Additional Information Regarding Loss of Coolant Accident Alternative Source Term Dose Analysis2005-05-26026 May 2005 Response to NRC Request for Additional Information Regarding Loss of Coolant Accident Alternative Source Term Dose Analysis ML0217001992002-06-0404 June 2002 Joint Appendix Volume III: Pages 1299 - 1556, Petition to Review a Final Decision of the Nrc. Affidavit of Eric A. Mccartney, Michael J. Devoe, Benjamin Wesley Morgan, and Deposition of Gordon R. Thompson Regarding Contention EC-6 Volmen 3 ML0037715352000-11-20020 November 2000 Affidavit of Gareth W. Parry, Stephen F. Lavie, Robert L. Palla and Christopher Gratton in Support of NRC Staff Brief and Summary of Relevant Facts, Data and Arguments Upon Which the Staff Proposes to Rely at Oral Argument on Environmental ML0036737512000-01-0404 January 2000 Affidavit of Kenneth C. Heck in Support of NRC Staff'S Written Summary ML18230A4841979-06-12012 June 1979 Transmitting Affidavit of J. A. Jones, Marked for Identification as Applicant Exhibit Pp, & Revisions to Applicant'S Proposed Findings of Fact & Conclusion of Law in Form of Supplemental Initial Decision (Construction Permits) ML18230A4911979-04-17017 April 1979 NRC Staff Motion to Accept Transcript Corrections ML18230A5241979-01-23023 January 1979 Applicant'S Response to Conservation Council of North Carolina and Wake Environment, Inc.'S Motion to Remand to Licensing Board for Further Hearings 2023-08-30
[Table view] Category:Letter
MONTHYEARIR 05000261/20244012024-09-11011 September 2024 Security Baseline Inspection Report 05000261/2024401 IR 05000400/20240112024-09-10010 September 2024 NRC Inspection Report 05000400/2024011 ML24242A2612024-08-29029 August 2024 Operator Licensing Examination Approval 05000261/2024301 IR 05000400/20240052024-08-26026 August 2024 Updated Inspection Plan for Shearon Harris Nuclear Power Plant, Unit 1 (Report 05000400-2024005) Rev 1 IR 05000261/20240052024-08-22022 August 2024 Updated Inspection Plan for H.B. Robinson Steam Electric Plant - Report 05000261/2024005 ML24169A2712024-08-14014 August 2024 – Issuance of Amendment No. 280 to Adopt TSTF-258-A, Revision 4, Regarding Changes to Technical Specification 5.7, High Radiation Area ML24059A4252024-08-14014 August 2024 Issuance of Amendment No. 202 Regarding Alignment of Certain Technical Specifications with Improved Standard Technical Specifications ML24213A0522024-08-0202 August 2024 Issuance of Amendment No. 201 to Extend Completion Time of Inoperable Reactor Coolant System Accumulator Using Consolidated Line Item Improvement Process IR 05000261/20240022024-08-0101 August 2024 Integrated Inspection Report 05000261/2024002 ML24212A3412024-07-31031 July 2024 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection ML24170A7312024-07-29029 July 2024 – Exemption from the Requirements of 10 CFR 50.55a(H)(2) Using the Risk-Informed Process for Evaluations Letter IR 05000400/20240022024-07-29029 July 2024 – Integrated Inspection Report 05000400/2024002 ML24183A0972024-07-12012 July 2024 ISFSI; Catawba 1, 2 & ISFSI; McGuire 1, 2 & ISFSI; Oconee 1, 2, 3 & ISFSI; Shearon Harris 1; H. B. Robinson 2 & ISFSI; and Radioactive Package Shipping Under 10 CFR 71 (71-266 & 71-345) – Review of QA Program Changes EPID L-2024-LLQ-0002 IR 05000400/20243022024-06-27027 June 2024 – NRC Operator Licensing Examination Approval 05000400/2024302 IR 05000400/20244012024-06-25025 June 2024 – Security Baseline Inspection Report05000400/2024401 ML24162A1372024-06-24024 June 2024 – Regulatory Audit Summary Related to the Review of Exemption Request from Certain Requirements in 10 CFR 50.55a(h)(2)(EPID L-2024-LLE-00040) ML24114A0152024-06-0303 June 2024 Unit 2 – Issuance of Amendment No. 279 Regarding Application of Leak-Before-Break Methodology for Auxiliary Reactor Coolant System Piping IR 05000261/20240112024-06-0303 June 2024 Focused Engineering Inspection- Commercial Grade Dedication Report 05000261/2024011 ML24136A1382024-05-20020 May 2024 – Notification of an NRC Fire Protection Team Inspection (FPTI) (NRC Inspection Report 05000400/2024011) and Request for Information (RFI) ML24116A2592024-05-14014 May 2024 Staff Evaluation Related to Aging Management Plan and Inspection Plan for Reactor Vessel Internals IR 05000261/20240012024-05-0909 May 2024 Integrated Inspection Report 05000261/2024001 ML24127A1592024-05-0808 May 2024 – Notification of Licensed Operator Initial Examination 05000400/2024302 IR 05000400/20240012024-05-0505 May 2024 Integrated Inspection Report 05000400/2024001 IR 05000261/20240102024-04-30030 April 2024 Biennial Problem Identification and Resolution Inspection Report 05000261/2024010 ML24100A0912024-04-10010 April 2024 Operator License Examination Report ML24058A2462024-03-18018 March 2024 – Supplemental Information Needed for Using the Risk-Informed Process for Evaluations for the Request for Exemption from Certain Requirements in 10 CFR 50.55a(h)(2) IR 05000400/20230062024-02-28028 February 2024 Annual Assessment Letter for Shearon Harris Nuclear Power Plant - NRC Inspection Report 05000400/2023006 IR 05000261/20230062024-02-28028 February 2024 Annual Assessment Letter for H.B. Robinson Steam Electric Plant Unit 2 - Report 05000261-2023006 ML24032A2632024-02-23023 February 2024 – Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0044 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000400/20243012024-02-21021 February 2024 – NRC Operator Licensing Examination Approval 05000400/2024301 IR 05000261/20243012024-02-0606 February 2024 – Notification of Licensed Operator Initial Examination 05000261/2024301 ML24033A0592024-02-0202 February 2024 Response to Request for Additional Information License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology IR 05000261/20230042024-01-31031 January 2024 Integrated Inspection Report 05000261/2023004 IR 05000400/20230042024-01-30030 January 2024 Integrated Inspection Report 05000400/2023004 ML24009A2432024-01-25025 January 2024 Unit, No. 2 - Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0047 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML24009A2712024-01-24024 January 2024 Revision to Reactor Vessel Material Surveillance Capsule Withdrawal Schedule ML23354A0052024-01-0808 January 2024 Request for Withholding Information from Public Disclosure, H. B. Robinson Steam Electric Plant, Unit No. 2 ML23342A0902023-12-0808 December 2023 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection IR 05000261/20234202023-11-30030 November 2023 Security Baseline Inspection Report 05000261/2023420 (Cover Letter with Report) IR 05000261/20230102023-11-28028 November 2023 Fire Protection Team Inspection Report 05000261/2023010 ML23317A3462023-11-14014 November 2023 Duke Fleet - Correction Letter to License Amendment Nos. 312 & 340 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 IR 05000400/20230032023-11-0909 November 2023 Integrated Inspection Report 05000400/2023003 IR 05000261/20230032023-11-0707 November 2023 Integrated Inspection Report 05000261 2023003 and 07200060 2023001 ML23226A0862023-10-12012 October 2023 Issuance of Amendment No. 277 Regarding Revision of TSs to Add High-High Steam Generator Level Function to Table 3.3.2-1 and Remove Obsolete Content from TSs 2.1.1.1 and 5.6.5.b ML23346A1322023-10-0606 October 2023 Communication from C-10 Research & Education Foundation Regarding NextEra Common Emergency Fleet Plan License Amendment Request and Related Documents Subsequently Published ML23234A1702023-10-0303 October 2023 Issuance of Amendment No. 199 Regarding Administrative Changes to the Renewed Facility Operating License and Technical Specifications ML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23195A0782023-08-29029 August 2023 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 IR 05000400/20230052023-08-23023 August 2023 Updated Inspection Plan for Shearon Harris Nuclear Power Plant, Unit 1 (Report 05000400/2023005) ML23235A0552023-08-23023 August 2023 Notification of an Fire Protection Team Inspection (FPTI) (NRC Inspection Report 05000261/2023010) and Request for Information (RFI) 2024-09-11
[Table view] Category:Report
MONTHYEARRA-24-0217, End of Cycle 25 (H1 R25) Inservice Inspection Program Owners Activity Report2024-09-0909 September 2024 End of Cycle 25 (H1 R25) Inservice Inspection Program Owners Activity Report RA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023 Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022 Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary) RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-21-0312, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2021-11-22022 November 2021 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) IR 05000261/20210052021-08-25025 August 2021 Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005) RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits2021-03-15015 March 2021 Notification of Permit Revision Request Regarding Copper and Zinc Limits RA-20-0381, CFR 50.54( Q) Screening Evaluation Form for Revisions2020-12-0808 December 2020 CFR 50.54( Q) Screening Evaluation Form for Revisions RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in . RA-20-0032, License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits2020-03-0606 March 2020 License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits RA-19-0403, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z2019-10-23023 October 2019 Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 RA-18-0179, 90-Day Special Report2018-10-0404 October 2018 90-Day Special Report ML18230A7562018-08-18018 August 2018 Fault Investigation, Responses to Mr. W. R. Butler'S Letter of May 16, 1975 ML18230A7592018-08-18018 August 2018 Fault Investigation, Progress Report, Volume 1 of 2 HNP-18-023, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes2018-05-0202 May 2018 Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes RNP-RA/18-0024, Report of Changes Pursuant to 10 CFR 50.59(d)(2)2018-04-0202 April 2018 Report of Changes Pursuant to 10 CFR 50.59(d)(2) RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) HNP-17-040, Snubber Program Plan2017-06-0101 June 2017 Snubber Program Plan HNP-17-006, Submittal of Cycle 20 Activity Report2017-01-30030 January 2017 Submittal of Cycle 20 Activity Report HNP-16-083, Mitigating Strategies Assessment Report for Flooding Hazard Information2016-12-21021 December 2016 Mitigating Strategies Assessment Report for Flooding Hazard Information ML16281A5102016-12-15015 December 2016 Staff Assessment of the Reactor Vessel Internals Aging Management Program Plans RNP-RA/16-0087, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-31031 October 2016 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors RNP-RA/16-0078, Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-0505 October 2016 Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RA-16-0024, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P2016-10-0303 October 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RNP-RA/16-0038, Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-1742016-05-25025 May 2016 Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-174 RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report HNP-16-017, Transmittal of Summary of a 10 CFR 50.54(q) Evaluation2016-02-29029 February 2016 Transmittal of Summary of a 10 CFR 50.54(q) Evaluation RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0047, Annual Report of Changes Pursuant to 10 CFR 50.462015-11-17017 November 2015 Annual Report of Changes Pursuant to 10 CFR 50.46 ML15301A5572015-11-0202 November 2015 Supplement to Staff Assessment of Response to 10 CFR 50.54(f) Information Request Flood Causing Mechanisms Reevaluations ML15280A1992015-10-19019 October 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review RNP-RA/15-0053, Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04)2015-08-19019 August 2015 Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04) RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RA-15-0022, Annual Report of Changes Pursuant to 10 CFR 50.462015-05-14014 May 2015 Annual Report of Changes Pursuant to 10 CFR 50.46 ML15126A0892015-04-30030 April 2015 EAL Bases HNP-15-027, Flood Hazard Reevaluation Report, Revision 12015-04-0101 April 2015 Flood Hazard Reevaluation Report, Revision 1 RNP-RA/15-0018, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel2015-02-26026 February 2015 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel RNP-RA/14-0037, Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System2014-04-0808 April 2014 Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System HNP-14-035, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.32014-03-27027 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 RNP-RA/14-0012, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3....2014-03-12012 March 2014 Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3.... RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident 2024-09-09
[Table view] Category:Technical
MONTHYEARRA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023 Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022 Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary) RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits2021-03-15015 March 2021 Notification of Permit Revision Request Regarding Copper and Zinc Limits RA-20-0381, CFR 50.54( Q) Screening Evaluation Form for Revisions2020-12-0808 December 2020 CFR 50.54( Q) Screening Evaluation Form for Revisions RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in . RA-20-0032, License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits2020-03-0606 March 2020 License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits RA-19-0403, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z2019-10-23023 October 2019 Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z RA-18-0179, 90-Day Special Report2018-10-0404 October 2018 90-Day Special Report RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) HNP-17-040, Snubber Program Plan2017-06-0101 June 2017 Snubber Program Plan ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML13365A2912014-02-19019 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML13364A2142014-02-12012 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14030A1022014-01-29029 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Shearon Harris Nuclear Power Plant, Unit 1, TAC No.: MF0874 ML14027A0632014-01-24024 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720 ML13267A2122013-09-30030 September 2013 Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML13270A1762013-09-24024 September 2013 Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety ML12331A1752012-11-26026 November 2012 Draft Bypass Fiber Quantity Test Plan ML12278A3992012-08-31031 August 2012 WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant ML1210707162012-05-17017 May 2012 Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant ML12067A1802012-02-29029 February 2012 ANP-3011Q1(NP), Revision 000, Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis. Enclosure 4 Response to Request for Additional Information (Non-Proprietary) ML12068A1332012-02-23023 February 2012 Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG ML1008904242010-03-0202 March 2010 NAI-1478-001, Revision 1, HNP CSAT Volume, Flow and Naoh Concentration Range Revisions. ML1008905932010-01-31031 January 2010 Application for Revision to Technical Specification Core Operating Limits Report References, ANP-2853(NP), Rev. 0, Realistic Large Break LOCA Summary Report. ML1002508582010-01-18018 January 2010 Holtec Report, No. HI-2043321, Rev. 6, Critically Safety Analyses of BWR Fuel Without Credit for Boraflex in the Racks at the Harris Nuclear Power Station. ML1003402842010-01-12012 January 2010 NAI-1478-001, Revision 0, HNP CSAT Volume, Flow and Naoh Concentration Range Revisions RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009 WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant. RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation ML0921500542009-05-31031 May 2009 ANP-2693(NP), Revision 0, Loss of Forced Reactor Coolant Flow Analysis for Harris Nuclear Plant, Unit 1, Enclosure 4 to Serial: HNP-09-068 ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 HNP-06-128, Engineering Report on: Evaluation of Charging-Safety Injection Pump Motor Bearing Temperatures at Progress Energy Harris Nuclear Plant2006-10-12012 October 2006 Engineering Report on: Evaluation of Charging-Safety Injection Pump Motor Bearing Temperatures at Progress Energy Harris Nuclear Plant ML0632802412006-10-0505 October 2006 Shearon Harris, Letter Report NAI-1282-001, Gothic Charging Pump Room Heatup Analysis for Shearon Harris. RNP-RA/06-0081, Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP2006-08-31031 August 2006 Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP RNP-RA/06-0027, ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.2006-03-31031 March 2006 ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis. ML0525105042005-08-31031 August 2005 Framatome Anp, Inc 77-5069740-NP-00, Shearon Harris Criticality Evaluation. ML0507004082004-02-20020 February 2004 EMF-3030(NP), Revision 0, Robinson Nuclear Plant, Realistic Large Break LOCA Analysis, February 2004, Non-Proprietary Version RNP-RA/03-0075, Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-62003-07-31031 July 2003 Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-6 RNP-RA/03-0031, Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 5042003-04-28028 April 2003 Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 504 ML0311207052003-04-18018 April 2003 Review of 90-day Steam Generator Tube Inservice Inspection Report for a Refueling Outage in 2001 2024-06-28
[Table view] |
Text
JOHN ELNITSKY Senior Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 704-382-4371 John.Elnitsky@duke
-energy.com
SUBJECT:
SUPPLEMENTAL INFORMATION FOR LICENSE AMENDMENT REQUEST REGARDING METHODOLOGY REPORT DPC
-NE-1008-P
REFERENCES:
Application to Revise Technical Specifications for Methodology Report DPC
-NE-1008-P Revision 0, "Nuclear Design Methodology Using CASMO
-5/SIMULATE
-3 for Westinghouse Reactors"Application t o Revise Technical Specifications to Adopt Methodology Reports DPC
-NF-2010 Revision 3 "Nuclear Physics Methodology for Reload Design" and DPC-NE-2011-P Revision 2, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors"Withdrawal of License Amendment Request Regarding Methodology Reports DPC
-NF-2010 and DPC
-NE-2011-PShearon Harris Nuclear Power Plant, Unit No. 1 and H. B. Robinson Steam Electric Plant, Unit 2 - Withdrawal of Requested Licensing Action Regarding Duke Energy Progress, Inc., Application to Revise Technical Specifications to Adopt Methodology Reports Submitted to NRC for Acceptance Review (CAC N os. MF7337 AND MF7338)
Attachment 1 Affidavit of John Elnitsky
Nuclear Design Methodology Using CASMO
-5/SIMULATE
-3 for Westinghouse ReactorsNuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors
Nuclear Design Methodology Using CASMO
-5/SIMULATE
-3 for Westinghouse ReactorsNuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors
EVALUATION OF THE PROPOSED CHANGE
McGuire Nuclear Station, Units 1 and 2 Issuance of Amendments Regarding Revision 1 to DPC
-NE-1005-P, Nuclear Design Methodology Using CASMO
-4/SIMULATE
-3 MOX (TAC Nos. MD7409 and MD7410)Catawba Nuclear Station, Units 1 and 2 Issuance of Amendments Regarding Revision 1 to DPC
-NE-1005-P, Nuclear Design Methodology Using CASMO
-4/SIMULATE
-3 MOX (TAC Nos. MD74 07 and MD7408)Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding the Use of CASMO-4/SIMULATE
-3 Methodology for Reactor Cores Containing Gadolinia Bearing Fuel (TAC Nos. ME4646 , ME4647, and ME4648)Topical Report on Physics Methodology for Reloads: McGuire and Catawba Nuclear Station , McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3222 and MB3223)Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3343 and MB3344)
Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station, Units 1 and 2, Re: Topical Report DPC
-NF-2010, Nuclear Physics Methodology for Reload Design, Revision 2Acceptance for Referencing of Topical Report DPC
-NE-2011-P, "Duke Power Company Nuclear Design Methodology for Core Operating Limits o f Westinghouse Reactors
Attachment 3 Proposed Technical Specification Changes (Mark-up)
Reporting Requirements 5.6 5.6 Reporting Requirements
5.6.2 Annual
Radiological Environmental Operating Report (continued)
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3 Radioactive
Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1. 5.6.4 DELETED 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: HBRSEP Unit No. 2 1. Shutdown Margin (SOM) for Specification 3.1.1; 2. Moderator Temperature Coefficient limits for Specification 3.1.3; 3. Shutdown Bank Insertion Limits for Specification 3.1.5; 4. Control Bank Insertion Limits for Specification 3.1.6; 5. Heat Flux Hot Channel Factor (Fa(Z)) limit for Specification 3.2.1; 6. Nuclear Enthalpy Rise Hot Channel Factor limit for Specification 3.2.2; (continued) 5.0-24 Amendment No. 212 No changes to this page. Included for information only. Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and 8. Boron Concentration limit for Specification 3.9.1. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents: HBRSEP Unit No. 2 1. Deleted 2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events," approved version as specified in the COLR. 3. XN-NF-82-21 (A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," approved version as specified in the COLR. 4. Deleted 5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow," approved version as specified in the COLR. 6. Deleted. 7. Deleted 8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR 9. XN-NF-621(A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR. 10. Deleted 11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR. 12. Deleted 13. Deleted. (continued) 5.0-25 Amendment No. 227 No changes to this page. Included for information only. Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
HBRSEP Unit No. 2 14. Deleted 15. Deleted 16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.B. Robinson Unit 2," approved version as specified in the COLR. 17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR. 18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR. 19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR. 20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR. 21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR. 22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs," approved version as specified in the COLR. 23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR. 24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR. (continued) 5.0-26 Amendment No. 227 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT <COLR) (continued)
Insert 1 (see next page) 25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water approved version as specified in the COLR. 26. BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR. 27. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,* approved version as specified in the COLR. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology," approved version as specified in the COLR. 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 Post Accident Monitoring
<PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring , the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued)
HBRSEP Unit No. 2 5.0-27 Amendment No. 2 4 4
No changes to this page. Included for information only. ADMIN I STRATIVE CONTROLS 6.9.l.6 CORE OPERATING LIMITS 6.9.1.6.l Core operating limits shall be e stablished and documented in the CORE OPERATING LIMITS REPORT CCOLR). plant prncedure PLP-106. prior to each reload cycle. or prior to any remaining portion of a reload cycle. for the fo 11 owing: a. b. c. d. e. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2. Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.
Shutdown Bank Insertion Limits for Specification 3/4.1.3.5. Control Bank Insertion Limits for Specification 3/4.1.3.6. Axial Flux Difference Limits for Specification 3/4.2.1. f. Heat Flux Hot Channel Factor. . K(Z). and VCZ) for Specification 3/4.2.2. g. Enthalpy Rise Hot Channel Factor. . and Power Factor Multiplier.
PF 6 H for Specification 3/4.2.3. h. Boron Concentration for Specification 3/4.9.1. 6.9.1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are Eerformed.
and the approved revision number shall be identified in the CO R. a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR. (Methodology for Sgecification 3.1.1.2 -SHUTDOWN MARGIN -MODES 3. 4 and 5. 3.1.1.3 -Moderator Temperature Coefficient.
3.1.3.5 -Shutdown Bank Insertion Limits. 3.1.3.6 -Control Bank Insertion Limits. 3.2.1 -Axial Flux Difference.
3.2.2 -Heat Flux Hot Channel Factor. 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor. and 3.9.1 -Boron Concentration).
- b. ANF-89-151(P)(A). "ANF-RELAP Methodology for Pressurized Water Reactors:
Analysis of Non-LOCA Chapter 15 Events." approved version as specified in the COLR. (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient.
3.1.3.5 -Shutdown Bank Insertion limits. 3.1.3.6 -Control Bank Insertion Limits. 3.2.1 -Axial Plux Difference.
3.2.2 -Heat Flux Hot Channel Factor. and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). c. XN-NF-82-21(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR. (Methodology for Specification 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). SHEARON HARRIS -UNIT 1 6-24 Amendment No. 94 No changes to this page. Included for information only. ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT !Continued)
- d. e. f. g. XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing,* approved version as specified in the COLR. (Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel. Factor). EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR. (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,*
1 Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012. (Methodology for Specification 3.2.1
- Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).
- XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR. (Methodology for Specification 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -C.ontrol Bank Insertion Limits, and 3.2.2 -Heat Flux Hot Channel Factor). SHEARON HARRIS -UNIT 1
3, 4
- 3. 4 15 Insert 2 ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued) (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). o. Mechanical Design Methodologies XN-NF-81-58(P}(A}, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR. ANF-81-58(P)(A}, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR. XN-NF-82-06(P}(A}, "Qualification of Exxon Nuclear Fuel for Extended Burn up," approved version as specified in the COLR. ANF-88-133(P}(A}, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Bumups of62 GWd/MTU," approved version as specified in the COLR. XN-NF-85-92(P}(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR. (see next page) EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR. (Methodologies for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor}. DPC-NE-2005-P-A, " Thermal-Hydraulic Statistical Core Design Methodology." approved version as specified in the COLR. (Methodology for Specification 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor} 6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear timits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. . 6.9.1.6.4 The CORE OPERA TING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shaU be submitted wi thin 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1. The report shall include: a. The scope of inspections performed on each SG, b. Degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism , SHEARON HARRIS -UNIT 1 6-24c Amendment No.1 4 8
Attachment 5 DPC-NE-1008, "Nuclear Design Methodology Using CASMO
-5/SIMULATE
-3 for Westinghouse Reactors" (Redacted
)
DD
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Cycle Table3-5 BOC, HZP Individual Control Bank Worth Comparisons Control Rod Ban k CBD CBC CBB CBA SBE SBD 3-21 SBC SBB Total SBA Worth a-c DDD D%&'&D
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Figure 3-12 cont'd Harris Unit 1Cycle18 Assembly Average Power Distribution Comparisons 100.0% FP, 192 EFPD, Control D at 218 SWD 3-32 a-c
Figure 3-12 cont'd Harris Unit 1 Cycle 18 Assembly Average Power Distribution Comparisons 100.0% FP, 485 EFPD, Control D at 218 SWD 3-35 a-c
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Attachment 6 DPC-NF-2010, Revision 3, "Nuclear Physics Methodology for Reload Design" and Technical Justification of Changes
K 1K 2
T 1K 1T 2K 2
=( )10= (/°)
, ,=( )10= (/°)
=( )10= (/%)
=+
=( )10= (/%)
( )=( )10= ()
Available Rod Worth Required Rod Worth
=1+Bias+UC+Ux1+Ux2+
l
== ==() ( 1)x in
=
= ()
=1+
=1() K a a
=1 Emthm th=BiasEmthn
=
DthCthMth
=
=()÷=()÷=()÷OSUTLX()=+
=1=[(())÷( 1)]KPKPP OSUTLD ()=+()
()=+()()=++()()=++()()=+()
UL(C) =UL(C) ()=++()
UL(C=()=+()
Incore instrument at corelocation L-05 is used for RVLIS
Technical Justification of Changes for Revision 3
Identification of the report location of each change refers back to the Revision 2a page numbering.
Attachment 8 DPC-NE-2011, Revision 2, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors" and Technical Justification of Changes (Redacted)
=1+Bias+(UC+Ux 1+Ux 2+...)
DNBM=MinMARP (x,y)RPP(x,y)MARP(x,y)100
CFMM=MinMAXLHRLHR(x,y,z)MAXLHR
LOCA LIMIT KW/FT CORE HEIGHT FT
% MARGIN % OFFSET LEGEND Bank D (swd)
LOCA MARGIN VS. AXIAL OFFSET
% MARGIN % OFFSET LOFA MARGIN VS. AXIAL OFFSET
F (x,y,z) K(Z)F (x,y,z) .K(Z)F F (x,y,z)UMTMTTILT[F (x,y,z)M(x,y,z)F(x,y,z)
F(x,y,z)NP(x,y,z)NP(x,y,z)M(x,y,z)
(,,)F(x,y,z)F F(x,y,z)
F(x,y,z)F(x,y,z)F(x,y,z)F(x,y,z)
F(x,y,z)
FFM(x,y,z)M(x,y,z)FF F (x,y,z)UMTMTTILT[F(x,y,z)M (x,y,z)]
F (x,y)MARP (x,y)1.0+1RRH(1.0P)
FFF (x,y)UMRTILT[F (x,y)M(x,y)] F (x,y)F
(,)F(x,y)F(x,y)
F(x,y)F(x,y)F(x,y)F(x,y)
F(x,y)
F FM(x,y)F
F (x,y,z)
F (x,y,z)
F (x,y,z)F (x,y,z)F (x,y,z)F(x,y,z)F (x,y,z)F(x,y,z)F (x,y, z)F (x,y, z)F(x,y,z)F(x,y,z)F (x,y,z)
F (x,y,z)F(x,y,z)
F (x,y,z)F(x,y,z)F (x,y, z)F (x,y,z)M (x,y,z)F (x,y, z)MBLP=( , ,)F (x,y, z)100%F (x,y,z)F (x,y, z)F (x,y, z) F (x,y)F (x,y,z)F (x,y)F (x,y)F (x,y)F (x,y) F(x,y)F (x, y) F (x,y)
F (x,y)F (x,y)F (x,y)M (x, y)F (x,y,z) ,F (x,y,z)F (x,y, z)F (x,y)F (x,y)
K(Z) Core Height, ft
MAX. ALLOWABLE TOTAL PEAK AXIAL PEAK LOCATION, X/L
% OF RATED THERMAL POWER FLUX DIFFERECE ( I) %
ROD BANK POSITION (STEPS WITHDRAWN)
PERCENT OF RATED THERMAL POWER
Technical Justification of Changes for Revision 2 (Redacted)
refers back to the Revision 1a page numbering
.
Section 1.3, Applicability of the Method:
Section 7,
References:
Appendix A:
Section 4.1 Change:
Section 4.5 Change:
Section 4.6 Change:
Appendix B Addition:
FF,F, (,)
Section 6.1 Change: Section 6.3 Change: Section 6.3.1 Change:
Section 6.1, equation on page 6-1 and removal of TILT definition: F (x,y,z)UMTMTTILT[F (x,y,z)M(x,y,z) Section 6.1, text and equation on page 6-3: FF(x,y,z)F(x,y,z)=F (x,y,z)M(x,y,z)/(UMTMTTILT)Section 6.2, equation on page 6-4: F (x,y,z)UMTMTTILT[F(x,y,z)M (x,y,z)]Section 6.2, text and equation on page 6-5: FF(x,y,z)
F(x,y,z)=F (x,y,z)M(x,y,z)/(UMTMTTILT)Section 6.3, equation on page 6-6 and removal of TILT definition: F (x,y)UMRTILT[F (x,y)M(x,y)]Section 6.3, text and equation on page 6-7: FM(x,y)F(x,y)F(x,y)=F (x,y)M(x,y)/(UMRTILT)
F(x,y,z)=F (x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y,z)=F (x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y)=F (x,y)M(x,y)/(UMRTILT)
(,,) F (x,y,z) F (x,y,z) NP(x,y,z)(,,)(,)F(x,y,z)F(x,y) Section 6.1.1 Change: 15-10-5 0 5 10 15 20 25-40-35-30-25-20-15-10-5 0 5 10 15 20 25 30 35 40LOCA FQ Margin (Limit
-FQ)/LIMIT*100AFD (%)Figure 1Example LOCA FQ Margin Versus AFDTypicalSteady State AFDOperating RangeNegative HFPAFD LimitPostive HFPAFD Limit FSection 6.3.1 Change: FSection 6.4.1 Change:
Section 6.4.2 Change: F(x,y,z)F(x,y)(,)F (x,y)F (x,y) RN P(x,y)
F (x,y,z)F (x,y,z)F(x,y,z)F(x,y,z)F (x,y,z)
F (x,y,z)F(x,y,z)F (x,y,z)F(x,y,z)
F (x,y,z)F(x,y,z)F (x,y,z)F(x,y,z)
F (x,y,z) ,F (x,y,z)F (x,y, z)F (x,y)F (x,y)
Change B-1: Appendix B Changes, third paragraph, page B-2:
Attachment 9 Application of the DPC
-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants
Page 1 Application of the DPC
-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants
1.0 Background
DPC-NE-2004-PA, "Core Thermal-Hydraulic Methodology Using VIPRE-01", Reference 1, describes Duke Energy's NRC
-approved steady state thermal hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. Key elements of the report include a description of the code inputs and correlations used to develop McGuire/Catawba VIPRE-01 models. The methodology used to determine allowable thermal-hydraulic operating limits in terms of core power level, reactor coolant system temperature and pressure, and three dimensional core power distributions is also presented. This methodology is used to determine allowable operating limits that provide DNB protection for nuclear plants that utilize a Westinghouse NSSS control and protection system. This method is described in Section 5 of Reference 1. Section 5 is applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba.
2.0 Applicability
The thermal-hydraulic methodology in Reference 1 is intended to interface with a Westinghouse plant reactor protection system (RPS) to ensure the required DNB design basis is satisfied. McGuire, Catawba, Harris and Robinson nuclear units all employ a Westinghouse RPS system to prevent fuel damage from occurring during Condition I and II events. DNB protection is accomplished through the use of the over
-temperature delta
-temperature (OTT) and over-power delta-temperature (OPT) trip functions, in combination with the following trips which limit the applicable range over which the OTT and OPT trip functions must provide DNB protection.
High pressurizer pressure trip Low pressurizer pressure trip Low reactor coolant flow trip
Core DNB limits are determined for a range of operating conditions to ensure that the DNB design basis is satisfied. The protected space in terms of power level and pressure is defined by the OTT/OPT trip, high pressurizer pressure trip, and low pressurizer pressure trip. An upper range on temperature is set by a combination of the Hot Leg boiling limit and the steam generator safety valve actuation. The low reactor coolant system (RCS) flow trip provides DNB protection by limiting the reactor coolant flow that must be considered.
The methodology described in Reference 1 requires the following three elements to be in place.
A NRC-approved VIPRE-01 model for the unit being evaluated A NRC-approved critical heat flux correlation for the fuel type A NRC-approved statistical DNBR limit for the fuel type
VIPRE-01 models have been developed to represent the Harris and Robinson core geometry and AREVA's HTP fuel product in current operation. The VIPRE-01 model developed, along with the code inputs and correlations used to the represent AREVA's Advanced W 17x17 HTP and 15x15 HTP fuel products are described in Appendices H and I of DPC-NE-2005-P-A, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology", Reference 2.
Page 2 The critical heat flux (CHF) correlation applicable to AREVA's Advanced W 15x15 and 17x17 HTP fuel designs is the HTP CHF correlation developed in Reference 3. The applicability of this correlation with the VIPRE-01 computer code was also demonstrated in Appendices H and I of Reference 2. In addition, Reference 2 describes the statistical core design (SCD) DNBR limit for each reactor considering plant specific uncertainties.
All three methodology components (VIPRE-01 model, CHF correlation and statistical DNBR limit) in Reference 2 are used to calculate core DNB limits. These limits are determined as a function of power level, reactor coolant system temperature, and reactor coolant system pressure using a reference radial peak power, FH, and the pin-by-pin power distribution specific to the applicable fuel design evaluated in Reference 2 Appendices H and I. Each point along the line corresponds to a constant DNBR value based on the SCD DNBR limit. DNB limit lines are provided in the core safety limits figure presented in Technical Specifications Figure 2.1-1 or provided in the cycle specific Core Operating Limits Report (COLR).
In summary, the Duke Energy thermal-hydraulic methodology used to provide DNB protection for Westinghouse plant instrumentation systems is not dependent upon the plant where it is applied. As a result, the method is as applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba. It does however, rely on the use of NRC-approved VIPRE-01 model, CHF correlation, and DNBR limit as specified in Reference 2 for each plant.
3.0 Maximum
Allowable Peaking Limits The Duke Energy maximum allowable total peaking (MATP) methodology is described in Section 5 of Reference 1. This method is independent of the reactor design or fuel type and consists of DNB calculations performed with an NRC-approved model, fuel design, critical heat flux correlation, and DNBR limit. The methodology is used to develop local fuel peaking limits to prevent DNB for a fixed reactor core operating condition defined by a thermal power level, pressure, RCS flow and reactor coolant temperature. The MATP limits developed are lines of constant MDNBR for a range of axial peak magnitudes with the location of the peak varied from the bottom to the top of the core. For a given axial peak magnitude (Fz) and axial location, the total peak (FH
- Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1.
The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function. The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient.
In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained.
Since NRC approval has been granted for the VIPRE
-01 model, CHF correlation, and DNBR limits in Page 3 References 2 and 3, application of the Duke Energy Core DNB limit and MATP methodology to the Harris and Robinson plants is appropriate. 4.0 Conclusion
The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved: VIPRE-01 model for the unit being evaluated Critical Heat Flux (CHF) correlation for the fuel type Statistical DNBR limit for the fuel type
Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system.
5.0 References
1.DPC-NE-2004-PA, Rev. 2a, "Core Thermal-Hydraulic Methodology Using VIPRE-01" 2.DPC-NE-2005-P-A, Rev. 5, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology" 3.EMF-92-153(P)(A), Rev. 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel" 4.DPC-NE-2011-PA, Rev. 1a, "Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors"