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Category:Report
MONTHYEARWBL-24-047, Analysis of Capsule V from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program2024-09-25025 September 2024 Analysis of Capsule V from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program WBL-24-022, Cycle 5 Steam Generator Tube Inspection Report2024-05-16016 May 2024 Cycle 5 Steam Generator Tube Inspection Report WBL-24-007, Technical Specification (TS) 5.7.2.15 - Explosive Gas and Storage Tank Radioactivity Monitoring Program2024-03-0505 March 2024 Technical Specification (TS) 5.7.2.15 - Explosive Gas and Storage Tank Radioactivity Monitoring Program CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) ML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information WBL-23-018, Revised Pressure and Temperature Limits Report (PTLR)2023-04-10010 April 2023 Revised Pressure and Temperature Limits Report (PTLR) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu WBL-22-034, Update to Fire Protection Report Figures Redacted2022-08-0101 August 2022 Update to Fire Protection Report Figures Redacted WBL-21-057, Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-12-16016 December 2021 Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report CNL-21-018, Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance .2021-12-0909 December 2021 Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance . WBL-21-056, Revised Pressure and Temperature Limits Report (PTLR)2021-12-0909 December 2021 Revised Pressure and Temperature Limits Report (PTLR) ML21244A3452021-09-20020 September 2021 Proposed Alternative IST RR 9 to the Requirements of the ASME OM Code for Test Plan Group 6 Relief Valves CNL-21-055, Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-07-20020 July 2021 Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report CNL-21-040, Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-03-23023 March 2021 Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) ML21060A9132021-03-17017 March 2021 Final Environmental Assessment and Finding of No Significant Impact for Initial and Updated Decommissioning Funding Plans for Watts Bar ISFSI CNL-21-011, Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-02-25025 February 2021 Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) WBL-21-006, Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-02-11011 February 2021 Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report WBL-20-066, Revised Pressure and Temperature Limits Report (PTLR)2020-12-16016 December 2020 Revised Pressure and Temperature Limits Report (PTLR) CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User2020-12-16016 December 2020 10 CFR 71.95 Report for 3-60B Casks User CNL-20-074, Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2020-08-28028 August 2020 Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) WBL-20-004, Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program2020-04-16016 April 2020 Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program CNL-19-082, License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2019-10-10010 October 2019 License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) WBL-19-026, Revised Pressure and Temperature Limits Report (PTLR)2019-04-0404 April 2019 Revised Pressure and Temperature Limits Report (PTLR) ML19039A0492019-02-0808 February 2019 Amd 1 to USAR Chapter 9 Auxiliary System NRC Additional Redactions ML19003A5692019-01-16016 January 2019 Review of the Fall 2017 Steam Generator Tube Inspection Report ML18242A0382018-08-30030 August 2018 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report CNL-18-092, Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02)2018-08-0101 August 2018 Application to Revise the Technical Specifications to Adopt TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (WBN-TS-18-02) CNL-18-007, Seismic Probabilistic Risk Assessment Supplemental Information2018-04-10010 April 2018 Seismic Probabilistic Risk Assessment Supplemental Information ML17356A2692017-12-20020 December 2017 Construction Lessons Learned Report ML17313A1282017-11-0909 November 2017 Revised Pressure and Temperature Limits Report (PTLR) CNL-17-134, Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations2017-10-13013 October 2017 Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations ML17272A0192017-09-29029 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17263A1162017-09-20020 September 2017 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report ML17209A5542017-07-28028 July 2017 Cycle 14 Steam Generator Tube Inspection Report CNL-17-050, Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index2017-05-30030 May 2017 Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index ML16215A1042016-08-0202 August 2016 Technical Specification (TS) 5.9.8 - Post Accident Monitoring System Report ML16113A0202016-04-22022 April 2016 Submittal of Title 10, Code of Federal Regulations 50.59 Summary Report CNL-16-038, Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information2016-03-31031 March 2016 Application to Revise Technical Specification 4.2.1, Fuel Assemblies and Radioactive Waste System Design Basis Source Term Supplement to Response to NRC Request for Additional Information CNL-16-034, TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program2016-02-19019 February 2016 TMI Task Action I.D.1 Commitment Closure Regarding Control Room Design Review Special Program CNL-15-263, Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory2015-12-29029 December 2015 Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Supplemental Information Related to the Onsite Regulatory Audit at Pacific Northwest National Laboratory CNL-15-204, Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit2015-09-21021 September 2015 Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit CNL-15-165, Submittal of Electromagnetic Interference (EMI) Survey Results2015-08-20020 August 2015 Submittal of Electromagnetic Interference (EMI) Survey Results CNL-15-143, The Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report2015-07-31031 July 2015 The Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report CNL-15-131, Individual Plant Examination of External Events (IPEEE) Report, Revision 32015-07-15015 July 2015 Individual Plant Examination of External Events (IPEEE) Report, Revision 3 CNL-15-106, 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information2015-07-0808 July 2015 3, Sequoyah and Watts Bar, Units 1 & 2 - Provides Service List Update for Routine Information CNL-15-097, Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2015-06-16016 June 2015 Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML15121A6562015-05-0101 May 2015 NRC Region II - CIB1 Watts Bar 2 Ip&S 194 Additional Questions Request List CNL-15-039, Severe Accident Management Alternatives for Reactor Coolant Pump Seals2015-04-10010 April 2015 Severe Accident Management Alternatives for Reactor Coolant Pump Seals CNL-15-043, Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Acc2015-03-25025 March 2015 Flood Hazard Reevaluation Report for Watts Bar, Response to NRC Request for Information Per Title 10 of CFR 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid 2024-09-25
[Table view] Category:Technical
MONTHYEARWBL-24-047, Analysis of Capsule V from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program2024-09-25025 September 2024 Analysis of Capsule V from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program WBL-24-022, Cycle 5 Steam Generator Tube Inspection Report2024-05-16016 May 2024 Cycle 5 Steam Generator Tube Inspection Report CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information WBL-23-018, Revised Pressure and Temperature Limits Report (PTLR)2023-04-10010 April 2023 Revised Pressure and Temperature Limits Report (PTLR) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu WBL-21-057, Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-12-16016 December 2021 Cycle 4 Mid-Cycle Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report WBL-21-056, Revised Pressure and Temperature Limits Report (PTLR)2021-12-0909 December 2021 Revised Pressure and Temperature Limits Report (PTLR) CNL-21-018, Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance .2021-12-0909 December 2021 Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance . CNL-21-055, Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-07-20020 July 2021 Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report CNL-21-040, Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-03-23023 March 2021 Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) CNL-21-011, Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)2021-02-25025 February 2021 Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002) WBL-21-006, Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report2021-02-11011 February 2021 Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report WBL-20-066, Revised Pressure and Temperature Limits Report (PTLR)2020-12-16016 December 2020 Revised Pressure and Temperature Limits Report (PTLR) CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User2020-12-16016 December 2020 10 CFR 71.95 Report for 3-60B Casks User CNL-20-074, Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2020-08-28028 August 2020 Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) WBL-20-004, Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program2020-04-16016 April 2020 Analysis of Capsule U from Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program CNL-19-082, License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2019-10-10010 October 2019 License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) WBL-19-026, Revised Pressure and Temperature Limits Report (PTLR)2019-04-0404 April 2019 Revised Pressure and Temperature Limits Report (PTLR) ML19039A0492019-02-0808 February 2019 Amd 1 to USAR Chapter 9 Auxiliary System NRC Additional Redactions ML17356A2692017-12-20020 December 2017 Construction Lessons Learned Report CNL-17-134, Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations2017-10-13013 October 2017 Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations CNL-17-050, Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index2017-05-30030 May 2017 Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index CNL-15-204, Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit2015-09-21021 September 2015 Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit CNL-15-143, The Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report2015-07-31031 July 2015 The Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report CNL-15-097, Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2015-06-16016 June 2015 Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident CNL-15-039, Severe Accident Management Alternatives for Reactor Coolant Pump Seals2015-04-10010 April 2015 Severe Accident Management Alternatives for Reactor Coolant Pump Seals ML15030A5082015-01-30030 January 2015 Tritium Production Program, Updated Plans for Cycle 13 Operation and Updated Evaluation of the Radiological Impacts of Tritium Permeation Into the Reactor Coolant System ML14100A0392014-04-0202 April 2014 Submittal of Pre-Operational Test Instruction CNL-14-038, Tennessee Valley Authoritys Seismic Hazard & Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Recommendation 2.1 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Acci2014-03-31031 March 2014 Tennessee Valley Authoritys Seismic Hazard & Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Recommendation 2.1 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accide ML13338A6832013-11-26026 November 2013 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Watts Bar Nuclear Plant, Units 1 and 2, TAC MF0950 and MF1177 ML13196A3762013-07-0909 July 2013 Submittal of Pre-op Test Instructions ML13206A0042013-06-24024 June 2013 Methodology for Evaluating the Potential for Multiple Dam Failures Due to Seismic Events ML13115A0362013-04-11011 April 2013 Engineering Information Record 51-9198783-000, Watts Bar WBN1C11 SG Inspection 180-Day Report ML13148A0142013-04-0404 April 2013 Preoperational Test, 2-PTI-068-13, Rev. 1, Shutdown from Outside the Main Control Room ML13162A3102013-04-0303 April 2013 2-PTI-002-01, Rev 000, Condensate System ML13081A0022013-03-13013 March 2013 Revised Watts Bar Nuclear Plant Unit 1/Unit 2 As-Designed Fire Protection Report. Part 1 of 2 ML13081A0032013-03-13013 March 2013 Revised Watts Bar Nuclear Plant Unit 1/Unit 2 As-Designed Fire Protection Report. Part 2 of 2 ML13162A3112013-02-25025 February 2013 2-PTI-026-01, Rev 000, High Pressure Fire Protection ML13044A1142013-01-31031 January 2013 Multiple Spurious Operation Evaluation Report R1976-20-01, Dated January 2013, Revision 2 ML13050A3982013-01-31031 January 2013 2-PTI-072-01, Rev 000, Containment Spray Pump Value Logic Test ML13162A3122012-11-16016 November 2012 2-PTI-003A-03, Rev 000, Main Feedwater System Functional Test ML12298A0592012-10-18018 October 2012 Submittal of 2-PTI-099-05, Rev 0, Overpower Delta-T & Overtemperature Delta-T Turbine Runback. ML13050A3972012-08-20020 August 2012 2-PTI-068-04, Rev 000, Pressurizer Relief Tank ML12236A1652012-07-19019 July 2012 Application to Revise Watts Bar Nuclear Plant (WBN) Unit 1 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis (WBN-UFSAR-12-01) ML12215A3382012-02-29029 February 2012 Enclosure 2, WCAP-17309-NP, Rev. 1, Watts Bar, Unit 2 Evaluation for Tube Vibration Induced Fatigue ML12073A3922012-02-29029 February 2012 WNA-VR-00283-WBT-NP, Rev. 7, Nuclear Automation Watts Bar Unit 2 NSSS Completion Program I&C Projects Iv&V Summary Report for the Post Accident Monitoring System. Attachment 2 ML12073A3592012-02-28028 February 2012 WBT-D-3769 Np, Common Q Pams Secure Development and Operational Environment Sser 23 Appendix Hh Action Item 98 Requests for Additional Information ML12073A2252012-02-28028 February 2012 Attachment 6, TVA Calculation WBPEVAR8807025, Revision 8, Bypassed and Inoperable Status Indication Logic Input Indications (Letter Item 4) ML12034A1662012-01-31031 January 2012 WBT-D-3753 NP-Enclosure - Clarification of Dielectric Withstand Testing in Response to WNA-CN-00157-WBT 2024-09-25
[Table view] |
Text
A 20004-019 (11/20/2012)
AREVA AREVA NP Inc.Engineering Information Record Document No.: 51 -9198783 -000 Watts Bar WBNIC11 SG Inspection 180-Day Report Page 1 of 16 A AREVA 20004-019 (11/20/2012)
Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Safety Related? [; YES F-1 NO Does this document establish design or technical requirements?
1 YES [7 NO Does this document contain assumptions requiring verification?
YES IX] NO Does this document contain Customer Required Format? YES NO Signature Block Pages/Sections Name and P/LP, R/LR, Prepared/Reviewed/
Title/Discipline Signature A-CRF, A Date Approved or Comments Victor Newman P All Principal Engineer David C. Ingram R All Technical Consultant
- 2) 4 '47,3 James P. Campbell /1 AI A All Technical Manager, .C&S Engineering Note: P/LP designates Preparer (P), Lead Preparer (LP)R/LR designates Reviewer (R), Lead Reviewer (LR)A-CRF designates Project Manager Approver of Customer Requested Format (A-CRF)A designates Approver/RTM
-Verification of Reviewer Independence Page 2 A AREVA 20004-019 (11/20/2012)
Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Record of Revision Revision Pages/Sections/
No. Paragraphs Changed Brief Description
/ Change Authorization 000 All Original Release 4 I-Page 3 A AR EVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Table of Contents Page SIG NATURE BLOCK .............................................................................................................................
2 RECO RD O F REVISIO N .......................................................................................................................
3 LIST O F TABLES ..................................................................................................................................
5 LIST O F FIG URES ................................................................................................................................
5 1.0 INTRO DUCTIO N ........................................................................................................................
6 2.0 180-DAY STEAM GENERATOR TUBE INSPECTION REPORT ...........................................
8
3.0 REFERENCES
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16 Page 4 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report List of Tables Page TABLE 2-1: TABLE 2-2: TABLE 2-3: TABLE 2-4: TABLE 2-5: TABLE 2-6: EDDY CURRENT INSPECTION SCOPE ......................................................................
8 NUMBER OF INDICATIONS DETECTED FOR EACH DEGRADATION MECHANISM...
10 NDE TECHNIQUES USED FOR EACH DEGRADATION MECHANISM
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11 SERVICE-INDUCED INDICATIO NS .............................................................................
12 NUMBER OF TUBES PLUGGED FOR EACH DEGRADATION MECHANISM
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15 TOTAL NUMBER AND PERCENTAGE OF TUBES PLUGGED TO DATE ..................
15 List of Figures Page FIGURE 1-1: TUBE SUPPORT ARRANGEMENT FOR WATTS BAR-1 STEAM GENERATORS
........ 7 Page 5 A AR EVA Document No.: 51-9198783-000 Watts Bar WBNlC11 SG Inspection 180-Day Report
1.0 INTRODUCTION
During the Watts Bar Unit 1 (WBN1) Fall 2012 refueling outage (designated as 1C1I), inspections of all four WBN1 steam generators (SGs) were performed.
These inspections included eddy current inspections of the SG tubing as well as primary and secondary side visual inspections.
This report documents the "Watts Bar lCl 1 SG Inspection 180-Day Report" as required by the WBN1 Technical Specifications.
Commercial operation of the original steam generators began in 1996. Operation of the replacement steam generators (RSG) Westinghouse/Model 68AXP SG design began following the Unit 1 Cycle 7 refueling outage in 2006 when the SGs were replaced.
The steam generators operated 1.184 EFPY in their first cycle after replacement which was the end of Cycle 8. An ISI (the first) was performed during that refueling outage. Cycle 11 is the second ISI of the replacement SGs (No inspections were performed in Cycles 9 or 10). The steam generators have operated a total of 5.254 EFPY since their replacement in the fall of 2006. They have operated 4.070 EFPY since the previous ISI.Figure 1-1 below provides the arrangement of the tube support structures for the WBN1 SGs.Page 6 A ARE VA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Figure 1-1: Tube Support Arrangement for Watts Bar-1 Steam Generators All U-Bend Support Straps are 2" Wide* TWO-PHASE EXIT REGION CROSS FLOW Tube Supports are fomaed by 2" and V" Lattice Bars AXIAL-COLD SIDE RECIRCULATING ENTRANCE CROSS FLOW FEEDWATER ENTRANCE CROSS FLOW HOT SIDE RECIRCULATING ENTRANCE CROSS FLOW -Note: VS = Vertical Strap, DS = Diagonal Strap Horizontal supports are a lattice grid design (ATSG)Page 7 A AR EVA Document No.: 51-9198783-000 Watts Bar WBNlC11 SG Inspection 180-Day Report 2.0 180-DAY STEAM GENERATOR TUBE INSPECTION REPORT In accordance with WBN1 Technical Specification 5.7.2.12, "Steam Generator Program", and Technical Specification 5.9.9, "Steam Generator Tube Inspection Report", this report documents the scope and results of the 1C 11 SG inspections.
There are eight specific reporting requirements (labeled "a" through "h" below). Each reporting requirement is followed with the required information based on the inspections performed during the 1C1 1 outage.a. The Scope of the Inspections Performed on Each SG The 1C 11 outage bobbin coil inspection was initially planned for 52.6% of the in-service tubes. The bobbin inspection included all tubes with prior indications of degradation.
Subsequent scope expansions to bound wear and foreign objects resulted in a final inspection of 57.8% of in-service tubes.In addition to the bobbin coil inspections, 5186 array coil inspections were also performed.
The array coil exams were aimed at detection of foreign objects and foreign object wear near the top of the tubesheet and support wear up to the C06 cold leg support (see Figure 1-1). Expansions were conducted to encompass wear and loose parts. Expansions in some areas were performed with array probe to have the benefit of enhanced PLP detection.
Table 2-1: Eddy Current Inspection Scope BOBBIN S/G Exam SIG E Tests Analyzed Planned Expanded and Completed 1-1 Full Length rows 5+ 2577 247 2824 H/L candycane rows 1-4 132 0 132 C/L straight rows 1-4 132 0 132 1-2 Full Length rows 5+ 2637 305 2942 H/L candycane rows 1-4 132 0 132 C/L straight rows 1-4 132 74 206 1-3 Full Length rows 5+ 2522 259 2781 H/L candycane rows 1-4 132 0 132 C/L straight rows 1-4 132 0 132 1-4 Full Length rows 5+ 2538 237 2775 H/L candycane rows 1-4 132 0 132 C/L straight rows 1-4 132 7 139 TOTAL 11330 1129 12459 Page 8 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C1 1 SG Inspection 180-Day Report Table 2-1: Eddy Current Inspection Scope (continued)
Array Top of Tubesheet S/G Planned Expanded Tests Analyzed Exams Exams and Completed 1-1 HL = H01-HTE 607 0 607 CL = C01-CTE/C08-CTE 607 44 651 1-2 HL = H01-HTE 607 0 607 CL = C01-CTE/C08-CTE/
VS2-CTE 607 51 658 1-3 HL = H01-HTE 605 0 605 CL = C01-CTE/C08-CTE 605 48 653 1-4 HL = H01-HTE 608 0 608 CL = C01-CTE/C08-CTE 608 189 797 TOTAL 4854 332 5186+Point Special Interest and historical preplan RPC S/G Locations to Tests Analyzed Inspect and Completed 1-1 Hot Leg SI 54 54 Cold Leg SI 1 1 U-bend Si 4 4 1-2 Hot Leg Sl 17 17 Cold Leg SI 3 3 U-Bend SI 5 5 1-3 Hot Leg SI 12 12 Cold Leg Sl 6 6 U-Bend SI 1 1 1-4 Hot Leg SI 10 10 Cold Leg SI 5 5 U-Bend SI 0 0 TOTAL 118 118 All Sco es Summary S/G Tests Analyzed and Completed 1-1 4405 1-2 4570 1-3 4322 1-4 4466 TOTAL 17763 Visual Inspection Results In addition to the eddy current inspections, visual inspections were also performed on both the primary and secondary sides. Primary side inspections included all previously installed plugs, the bowl cladding, and the divider plate. The plug inspections showed no anomalies indicative of improper installation or in-Page 9 A AR EVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report service degradation.
The inspections of the bowl cladding and the divider plate showed no visual evidence of degradation of the cladding, divider plate, or divider plate welds.Secondary side visual inspections were performed on the steam drums in SGs 12 and 13 to investigate the structural integrity of the steam drum components.
There were no anomalies reported in either steam drum. The visual to the extent possible included the perforated riser pipes, steam dryers, drains, drain cups, auxiliary feed water nozzle and sprayers, transmitter penetrations, hardware screws and nuts between the separators, upper dryers, and the Marmon clamp. All components were covered evenly with 1/16" of magnetite.
Steam drum inspections in SGs 11 and 14 were not inspected due to schedule and ALARA considerations.
Secondary side visual inspections were also performed at the top of the tubesheet in all SGs for the detection of foreign objects and for determining the effectiveness of water lancing. The cleanliness inspection at the top of the tubesheet resulted in additional sludge lancing. In bundle visuals were also performed by inserting the probe down several columns in each SG.SG 4 was the only SG where foreign material was identified on the tubesheet.
There were three pieces of wire identified.
The pieces of wire appeared to be TIG wire. One retrieved from the cold leg annulus 3 3 not identified by eddy current was 3%" x /64". Another piece retrieved, identified by eddy current, was 2%" x 3/64". Per analysis guidelines, once the foreign object was removed and the signal was confirmed to no longer be present, the PLP records (it was seen on two adjacent tubes) were permanently removed from the ECT database.
The third piece, not identified by eddy current, was 3" x /64". It could not be retrieved after two attempts.
It appeared to be held in place due to the smaller gap between the tube and a stay rod (stay rods have a larger diameter than the tubes). The inability to retrieve the loose part resulted in the affected tubes bounding the loose part being plugged and stabilized.
The PLP identification and disposition for tube plugging and stabilization is documented in AREVA CR 2012-7370.
One other indication in SG 4 identified by eddy current was examined visually.
There was no appearance of a loose part. It was removed from the database.The feedwater distribution boxes in each SG were inspected for the presence of foreign objects. These boxes have 0.29" diameter holes that serve to restrict foreign material of subsequent size from entering the steam generators.
A number of foreign objects were removed from the boxes. SG 4 had the largest amount of material where 35 small objects were removed. Some of the objects appeared to be Furmanite.
PLP indications were recorded in SG 14 tubes 1-60, 1-62, and 1-64 but were attributed to the presence of water lancing equipment in the SG at the time of the ECT exam.b. Active Degradation Mechanisms Found Volumetric wear was the only degradation mechanism detected during the 1 Cl 1 inspection.
All of the wear indications detected were located at support structures either in the U-bend at vertical or diagonal straps or in the lattice structure of the Advanced Tube Support Grids (ATSG). No wear indications attributable to foreign objects were reported.
Table 2-2 below shows the number of indications reported during the lCl 1 inspection.
Table 2-2: Number of Indications Detected for Each Degradation Mechanism Page 10 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Location of Wear SG 11 SG12 SG13 SG14 Total Indication Tubes Ind Tubes Ind Tubes Ind Tubes Ind Tubes Ind U-bend support 0 0 2 3 3 3 2 2 7 8 Detected Tube Support Grid 6 6 13 20 14 19 18 27 51 72 U-bend support 0 0 0 0 0 0 0 0 0 0 Plugged Tube Support Grid 1 1 3 6 4 9 6 15 14 31 Returnedto U-bend support 0 0 2 3 3 3 2 2 7 8 Service Tube Support Grid 5 5 10 14 10 10 12 12 37 41 Note: All tubes with a wear indication of 15% or greater were plugged. Some of the tubes had multiple indications of wear which varied in size (see Table 2-4). All pluggable degradation was related to cold leg support wear.c. Nondestructive Examination (NDE) Techniques Used for Each Degradation Mechanism Table 2-3 below provides the NDE techniques that were used for the detection of each degradation mechanism that was considered as existing or potential for the 1C1 1 inspection.
Table 2-3: NDE Techniques Used for Each Degradation Mechanism Degradation Mechanism Detection Technique U-Bend / Vertical Strap Wear Bobbin Horizontal
/ Tube Support Grid Wear Bobbin Tube-to-Tube Wear Bobbin Foreign Object Wear Bobbin / Array In addition to the detection techniques shown in the above table, +PointTM probes were also used for confirmation, characterization, and length sizing of wear indications at the vertical straps and horizontal grid supports.d. Location, Orientation (if Linear), and Measured Sizes (if Available) of Service Induced Indications Table 2-4 below provides a listing of all service-induced indications reported during the 1 C11 inspection including the estimated depths from the bobbin coil. If length sizing was available, this information is also provided as reported from the +PointTm inspections.
Page 11 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C1 1 SG Inspection 180-Day Report Table 2-4: Service-Induced Indications SG ROW COL IND %TW LOCATION Inch VOLTS Circ extent Ax extent Characterization Disposition 11 7 92 TWD 9 C04 0.93 0.31 ATSG Wear 11 9 104 TWD 12 H07 -0.87 0.39 ATSG Wear 11 29 124 TWD 13 C06 0.77 0.44 ATSG Wear 11 37 122 TWD 26 C05 0.84 1.26 0.53 0.54 ATSGWear Plugged 11 89 96 TWD 11 C03 -0.92 0.35 ATSG Wear 11 105 64 TWD 11 C05 -0.88 0.36 ATSG Wear 12 22 123 TWD 9 C03 0.84 0.3 ATSG Wear 12 24 123 TWD 11 C03 0.84 0.39 ATSG Wear Plugged 12 24 123 TWD 17 C07 -0.88 0.72 ATSG Wear Plugged 12 28 119 TWD 8 C06 -0.82 0.26 ATSG Wear 12 30 123 TWD 12 C05 -0.89 0.38 ATSG Wear 12 30 123 TWD 8 C06 -0.84 0.23 ATSG Wear 12 30 123 TWD 14 C06 0.84 0.48 ATSG Wear 12 30 123 TWD 13 C07 -0.86 0.44 ATSG Wear 12 31 122 TWD 24 C06 -0.82 1.15 0.28 0.45 ATSGWear Plugged 12 31 122 TWD 25 C07 -0.85 1.18 0.15 0.24 ATSG Wear Plugged 12 31 122 TWD 11 C08 -0.83 0.4 ATSG Wear Plugged 12 62 113 TWD 12 C07 -0.99 0.37 ATSG Wear 12 85 100 TWD 8 C06 -0.88 0.28 ATSG Wear 12 87 98 TWD 10 C03 -0.87 0.28 ATSG Wear 12 87 98 TWD 13 C06 -0.88 0.41 ATSG Wear 12 88 95 TWD 14 C03 -1.01 0.49 ATSG Wear 12 89 94 TWD 8 C03 -0.93 0.21 ATSG Wear 12 90 93 TWD 28 C06 -0.98 1.21 0.28 0.45 ATSGWear Plugged U Bend Support 12 91 58 TWD 8 VS2 0.8 0.22 Wear 12 91 94 TWD 8 C03 -0.89 0.22 ATSG Wear U Bend Support 12 100 63 TWD 12 VS2 -0.75 0.37 Wear U Bend Support 12 100 63 TWD 7 VS2 0.55 0.2 Wear 12 105 66 TWD 9 C03 0.84 0.27 ATSGWear 13 15 4 TWD 8 C06 -0.9 0.22 ATSGWear 13 19 2 TWD 11 C04 -0.86 0.38 ATSG Wear 13 23 124 TWD 12 C06 0.88 0.46 ATSG Wear 13 61 114 TWD 5 C06 0.82 0.21 ATSG Wear 13 86 29 TWD 21 C03 -0.95 0.77 0.61 0.53 ATSG Wear Plugged 13 86 29 TWD 9 C03 0.81 0.24 ATSG Wear Plugged Page 12 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report Table 2-4: Service-Induced Indications SG ROW COL IND %TW LOCATION Inch VOLTS Circ extent Ax extent Characterization Disposition 13 86 97 TWD 12 C02 -0.83 0.48 ATSG Wear 13 86 99 TWD 8 C04 0.89 0.29 ATSG Wear 13 87 96 TWD 24 C03 -0.87 1.11 0.49 0.49 ATSG Wear Plugged 13 87 98 TWD 26 C02 -0.82 1.41 0.54 0.56 ATSG Wear Plugged 13 87 98 TWD 18 C04 -0.87 0.8 ATSG Wear Plugged 13 87 98 TWD 17 C04 0.79 0.73 ATSG Wear Plugged 13 87 98 TWD 26 C05 0.79 1.4 0.56 0.58 ATSG Wear Plugged 13 88 95 TWD 21 C03 -0.77 0.64 0.49 0.53 ATSG Wear Plugged 13 88 95 TWD 32 C03 0.84 1.48 0.51 0.5 ATSGWear Plugged 13 89 96 TWD 8 C03 -0.82 0.28 ATSG Wear 13 92 35 TWD 10 C03 -0.95 0.3 ATSG Wear 13 92 91 TWD 8 C05 -0.96 0.17 ATSG Wear 13 97 42 TWD 9 C04 0.78 0.34 ATSG Wear U Bend Support 13 98 71 TWD 9 VS4 0.82 0.22 Wear U Bend Support 13 100 55 TWD 8 VS2 -0.85 0.29 Wear U Bend Support 13 101 58 TWD 12 DS2 -0.86 0.4 Wear 14 13 78 TWD 11 C10 0.72 0.39 ATSGWear 14 23 124 TWD 11 C06 -0.82 0.38 ATSG Wear 14 27 124 TWD 11 C03 -0.85 0.38 ATSG Wear 14 35 122 TWD 9 C03 0.9 0.3 ATSG Wear 14 49 110 TWD 8 C12 -0.91 0.25 ATSG Wear 14 53 118 TWD 8 C02 -0.8 0.26 ATSG Wear 14 63 114 TWD 11 C06 -0.98 0.4 ATSG Wear 14 84 99 TWD 18 C03 0.82 0.74 ATSG Wear Plugged 14 85 100 TWD 8 C04 -0.84 0.26 ATSG Wear 14 86 99 TWD 25 C04 -0.98 1.24 0.48 0.51 ATSG Wear Plugged 14 86 99 TWD 10 C06 -0.94 0.35 ATSG Wear Plugged 14 89 96 TWD 10 C03 -0.84 0.31 ATSG Wear 14 90 95 TWD 14 C03 0.68 0.52 ATSG Wear Plugged 14 90 95 TWD 10 C04 -0.95 0.33 ATSG Wear Plugged 14 90 95 TWD 28 C06 -0.94 1.56 0.68 0.54 ATSGWear Plugged 14 90 95 TWD 8 C06 0.81 0.26 ATSG Wear Plugged 14 90 95 TWD 8 C08 -1.07 0.28 ATSG Wear Plugged 14 91 94 TWD 9 C03 0.82 0.31 ATSG Wear 14 92 35 TWD 19 C03 -1.06 0.8 ATSG Wear Plugged U Bend Support 14 92 81 TWD 8 VS2 0.9 0.3 Wear Page 13 A AREVA Document No.: 51-9198783-000 Watts Bar WBNlCI1 SG Inspection 180-Day Report Table 2-4: Service-Induced Indications SG ROW COL IND %TW LOCATION Inch VOLTS Circ extent Ax extent Characterization Disposition 14 93 36 TWD 11 C05 0.85 0.35 ATSG Wear U Bend Support 14 93 70 TWD 10 VS2 0.88 0.29 Wear 14 97 40 TWD 15 C03 -0.84 0.56 ATSG Wear Plugged 14 97 40 TWD 8 C04 0.85 0.26 ATSG Wear Plugged 14 97 40 TWD 9 C05 0.82 0.27 ATSG Wear Plugged 14 97 40 TWD 8 C06 -0.82 0.26 ATSG Wear Plugged 14 97 86 TWD 9 C03 0.85 0.39 ATSG Wear 14 100 45 TWD 10 C03 -0.95 0.35 ATSG Wear Plugged 14 100 45 TWD 19 C05 -0.96 0.78 ATSG Wear Plugged Page 14 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C1 1 SG Inspection 180-Day Report e. Number of Tubes Plugged During the Inspection Outage for Each Active Degradation Mechanism Table 2-5 below provides the numbers of tubes plugged for each degradation mechanism detected.Table 2-5: Number of Tubes Plugged for Each Degradation Mechanism ATSG Wear Foreign Total SG (preventative)
Object 11 1 0 1 12 3 0 3 13 4 0 4 14 6 *6 12* No tube degradation was identified with the foreign object. Tubes were plugged and stabilized to bound the loose part.f. Total Number and Percentage of Tubes Plugged to Date Table 2-6: Total Number and Percentage of Tubes Plugged to Date Inspection SG 11 SG 12 SG 13 SG 14 Total Prior to service 0 1 0 1 2 1C8 2008 2 1 3 1 7 ICI 2012 1 3 4 12 20 Total 3 5 7 14 29 Percent 0.06% 0.1% 0.14% 0.27% 0.14%Page 15 A AREVA Document No.: 51-9198783-000 Watts Bar WBN1C11 SG Inspection 180-Day Report g. The Results of Condition Monitoring, Including the Results of Tube Pulls and In-Situ Testing Tube Inteqrity As required by the WBN1 Steam Generator Program, a condition monitoring (CM) assessment was performed
[1]. The only tube degradation detected during the IC1 1 inspection was wear at support structures (U bend supports and ATSGs). The deepest indication had an estimated depth of 32%TW from the bobbin coil exam. This indication was located at an ATSG. The maximum NDE length of any ATSG wear flaw was 0.68". The CM limit for a flaw of this length is approximately 55% TWD. This CM limit includes uncertainties for material properties, NDE depth sizing, and the burst pressure relationship.
Since the deepest flaw has an NDE depth less than the CM limit, the structural integrity performance criterion was met for the operating period prior to 1 Cl .Since wear indications will leak and burst at essentially the same pressure, accident-induced leakage integrity at a much lower accident pressure differential is also demonstrated.
Operational leakage integrity was demonstrated by the absence of any detectable primary-to-secondary leakage during the operating period prior to 1 C 1.Since tube integrity was demonstrated analytically, in-situ pressure testing was not required nor performed during the 1Ci I outage. Likewise, no tube pulls were planned nor performed during 1C11.h. The Effective Plugging Percentage for All Plugging in Each SG There are no sleeves installed in the WBN1 SGs. Therefore, the effective plugging percentage is the same as the plugging percentage shown in Table 2-6.
3.0 REFERENCES
- 1. AREVA Document 51-9196412-000 "Watts Bar Unit 1 Condition Monitoring for ICI I and Final Operational Assessment for Cycles 12, 13, and 14" Page 16