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Category:Letter
MONTHYEARML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues ML24008A2462024-01-18018 January 2024 Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions 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) CNL-23-052, Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability2024-01-0909 January 2024 Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability CNL-23-062, Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018)2024-01-0808 January 2024 Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018) ML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods CNL-23-069, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000390/20234412023-12-21021 December 2023 Plantfinal Significance Determination for a Security-Related Greater than Green Finding, Nov, and Assessment Follow-up, 05000390-2023441 and 05000391-2023441-Public CNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) IR 05000390/20234042023-12-14014 December 2023 Security Baseline Inspection Report 05000390/2023404 and 05000391/2023404 CNL-23-001, Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01)2023-12-13013 December 2023 Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01) ML23293A0572023-12-0606 December 2023 Issuance of Amendment Nos. 163 and 70 Regarding Adoption of TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control IR 05000390/20230102023-11-30030 November 2023 RE-Issue Watts Bar Nuclear Plant - Biennial Problem Identification and Resolution Inspection Report 050000390/2023010 and 05000391/2023010 and Apparent Violation CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000390/20230032023-11-13013 November 2023 Integrated Inspection Report 05000390/2023003 and 05000391/2023003 and Apparent Violation ML23312A1432023-11-0808 November 2023 Submittal of Dual Unit Updated Final Safety Analysis Report (UFSAR) Amendment 5 CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision ML23251A2002023-09-11011 September 2023 Request for Withholding Information from Public Disclosure for Watts Bar Nuclear Plant, Units 1 and 2 CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000390/20230052023-08-30030 August 2023 Updated Inspection Plan for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2023005 and 05000391/2023005 ML23233A0042023-08-28028 August 2023 Proposed Alternative to the Requirements of the ASME Boiler and Pressure Vessel Code for Upper Head Injection Dissimilar Metal Butt Welds IR 05000390/20230022023-08-16016 August 2023 Reissue - Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2023002 and 05000391/2023002 ML23220A1582023-08-0909 August 2023 Integrated Inspection Report 05000390/2023002 and 05000391/2023002 CNL-23-045, License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010)2023-08-0707 August 2023 License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010) CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) 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 CNL-23-055, Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills2023-07-25025 July 2023 Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills IR 05000390/20230112023-07-24024 July 2023 Quadrennial Focused Engineering Inspection (FEI) Commercial Grade Dedication Report 05000390 2023011 and 05000391 2023011 CNL-23-053, Tennessee Valley Authority - Radiological Emergency Plan Revisions2023-07-18018 July 2023 Tennessee Valley Authority - Radiological Emergency Plan Revisions CNL-23-020, Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06)2023-06-28028 June 2023 Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06) CNL-23-049, Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan .2023-06-26026 June 2023 Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan . ML23122A2322023-06-0707 June 2023 Issuance of Amendment Nos. 162 and 69 Regarding Change to Date in Footnotes for Technical Specification 3.7.11, Control Room Emergency Air Temperature Control System (Creatcs) CNL-23-044, Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out2023-06-0101 June 2023 Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out IR 05000390/20234032023-05-30030 May 2023 Cyber Security Inspection Report 05000390/2023403 and 05000391/2023403 ML23131A1812023-05-23023 May 2023 Correction to Amendment No. 161 to Facility Operating License No. NPF-90 CNL-23-042, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-05-16016 May 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000390/20220032023-05-0909 May 2023 Reissue Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2022003 and 05000391/2022003 ML23125A2202023-05-0505 May 2023 Issuance of Amendment No. 161 Regarding a Change to Footnotes for Technical Specification Table 1.1-1 Modes (Emergency Circumstances) IR 05000390/20230012023-05-0404 May 2023 Integrated Inspection Report 05000390/2023001 and 05000391/2023001 CNL-23-043, Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09)2023-05-0404 May 2023 Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09) CNL-23-032, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 412023-04-27027 April 2023 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 41 CNL-23-030, Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update2023-04-27027 April 2023 Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-23-033, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-04-24024 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision CNL-23-029, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-04-11011 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23072A0652023-04-0505 April 2023 Units 1 and 2 Issuance of Amendment Nos. 364 and 358; 160 and 68 Regarding a Revision to Technical Specification 3.4.12 ML23073A2762023-04-0303 April 2023 Individual Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing (EPID L-2023-LLA-0029) (Letter) CNL-23-023, Annual Insurance Status Report2023-03-30030 March 2023 Annual Insurance Status Report CNL-23-024, TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report2023-03-29029 March 2023 TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report 2024-01-09
[Table view] Category:Report
MONTHYEARCNL-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) L-19-034, Wafts Bar Nuclear Plant, Unit 1 - Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report2019-06-18018 June 2019 Wafts Bar Nuclear Plant, Unit 1 - Technical Specification (TS) 5.9.8 - Post Accident Monitoring System (Pams) Report L-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 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 CNL-14-212, Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-12-30030 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident 2024-01-03
[Table view] Category:Miscellaneous
MONTHYEARML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods WBL-22-034, Update to Fire Protection Report Figures Redacted2022-08-0101 August 2022 Update to Fire Protection Report Figures Redacted 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 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 ML17313A1282017-11-0909 November 2017 Revised Pressure and Temperature Limits Report (PTLR) 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 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-15-165, Submittal of Electromagnetic Interference (EMI) Survey Results2015-08-20020 August 2015 Submittal of Electromagnetic Interference (EMI) Survey Results 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 ML15121A6562015-05-0101 May 2015 NRC Region II - CIB1 Watts Bar 2 Ip&S 194 Additional Questions Request List 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 CNL-14-212, Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-12-30030 December 2014 Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML14163A6582014-09-18018 September 2014 Closeout of Generic Letter, 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML14212A6032014-07-31031 July 2014 WBRD-50-391/86-60 - Final Report and Revised Completion Schedule IR 05000391/19860602014-07-31031 July 2014 WBRD-50-391/86-60 - Final Report and Revised Completion Schedule ML14149A1502014-06-16016 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14133A5422014-05-23023 May 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML13246A0222013-08-28028 August 2013 Submittal of Pre-op Test Instruction ML13178A2812013-06-26026 June 2013 10 CFR 50.59 Summary Report Supplement ML13144A5762013-05-22022 May 2013 Watt Bar, Units 1 & 2, Report of Drug Testing Error in Accordance with 10 CFR 26.719(c)(1) ML13126A2942013-04-29029 April 2013 10 CFR 50.59 Summary Report ML13121A0602013-04-29029 April 2013 Commitment Summary Report ML13080A3632013-03-18018 March 2013 Enclosure 3, Summer 2011 Compliance Survey for Watts Bar Nuclear Plant Outfall Passive Mixing Zone ML13080A3662013-03-18018 March 2013 Enclosure 1, Summer 2010 Compliance Survey for Watts Bar Nuclear Plant Outfall Passive Mixing Zone ML13080A0732013-03-12012 March 2013 Extension Request Regarding the Flooding Hazard Reevaluation Report Required by NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation.. ML13175A1352013-03-0505 March 2013 2-PTI-092-03, Revision 0, Nuclear Instrumentation Source Range Noise Checks During Hot Functional Testing. ML12356A3172012-12-17017 December 2012 Submittal of Pre-op Test Instruction, 2-PTI-063-06, Revision 0, Safety Injection System Check Valve Test. ML12335A3402012-11-27027 November 2012 Tennessee Valley Authority - Fleet Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding the Flooding Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of ML13108A2842012-11-12012 November 2012 Unit 1, Fukushima Near-Term Task Force Recommendation 2.3: Seismic Response Report ML13175A1362012-11-0808 November 2012 2-PTI-099-06, Revision 0, Reactor Protection Setpoint Verification. ML13175A1342012-11-0101 November 2012 2-PTI-082-02, Revision 0, Rod Control - Non Hft. ML13175A1332012-10-22022 October 2012 2-PTI-085-01, Revision 0, Rod Control Functional Test. ML12236A1642012-07-19019 July 2012 Enclosure 1 Evaluation of Proposed Changes Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 1 ML12223A1832012-03-29029 March 2012 Environmental Protection Agency 2012 - Facility Detail Report - Environmental Facts Warehouse Fii - Moccasin Bend Wwtp ML11362A0562011-12-20020 December 2011 Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 7 (TAC No. MD6311), and Status of Generic Communications for Unit 2 - Revision 7 ML11341A1572011-11-30030 November 2011 Attachments 7 Through 9, WNA-CN-00157-WBT-NP, Revision 1, CAW-11-3316, and WBT-D-3566 Np, Incore Instrument System Signal Processing System Isolation Requirement ML11326A2842011-11-18018 November 2011 Commitment Summary Report ML11257A0502011-08-31031 August 2011 Attachment 7, WCAP-17427-NP, Rev. 1, Watts Bar Nuclear Plant Unit 2 Common Q Post Accident Monitoring System Computer Security Assessment, Attachment 8, Application for Withholding Proprietary Information from Public Disclosure and Attachme ML1104003852011-02-0707 February 2011 Enclosure 2, Appendix a, Hydrothermal Effects on the Ichthyhoplankton from the Watts Bar Nuclear Plant Supplemental Condenser Cooling Water Outfall in Upper Chickamauga Reservoir ML1104003842011-02-0707 February 2011 Enclosure 1, Hydrothermal Effects on the Ichthyoplankton from Watts Bar Nuclear Plant Supplemental Condenser Cooling Water Outfall in Upper Chickamauga Reservoir 2024-01-03
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Text
January 21, 2009 10 CFR 50.59(d)(2)
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket No. 50-390 Tennessee Valley Authority )
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - 10 CFR 50.59, CHANGES, TESTS, AND EXPERIMENTS AND COMMITMENT
SUMMARY
REPORTS, SUPPLEMENTAL REPORT This letter supersedes TVAs letter, Watts Bar Nuclear Plant (WBN) Unit 1 - 10 CFR 50.59, Changes, Tests, and Experiments and Commitment Summary Reports, dated September 24, 2008. This letter includes two additional 50.59 safety evaluations that should have been included in the September report. These omissions were identified in a recent self-assessment and have been entered for resolution into TVAs corrective action program as Problem Evaluation Report 160317. Additional clarifications were also made regarding the conclusions of two evaluations.
Pursuant to 10 CFR 50.59(d)(2), this letter provides the complete Summary Report of the implemented changes, tests, and experiments in which evaluations were performed in accordance with 10 CFR 50.59(c). The enclosure provides a summary of the evaluations associated with Updated Final Safety Analysis Report Amendment 7 (provided separately) and includes other evaluations implemented during the period from March 17, 2007 to June 30, 2008.
During this reporting period, there were no previous commitments that TVA has evaluated and revised using administrative controls that incorporate the Nuclear Energy Institutes (NEI) 99-04 Guideline For Managing NRC Commitments.
U.S. Nuclear Regulatory Commission Page 2 January 21, 2009 There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please call me at (423) 365-1824.
Sincerely, Original signed by M. K. Brandon Manager, Site Licensing and Industry Affairs Enclosure cc: See Page 3
U.S. Nuclear Regulatory Commission Page 3 January 21, 2009 Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
U.S. Nuclear Regulatory Commission Page 3 January 21, 2009 MKB:
Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 G. Arent, EQB 1B-WBN G. A. Boerschig, MOB 2R-WBN W. R. Campbell, LP 3R-C W. T. Coutu, LP 3R-C D. E. Jernigan, LP 3R-C M. T. McFadden, ADM 1Q-WBN G. W. Mauldin, EQB 2A-WBN L. E. Nicholson, LP 3R-C M. A. Purcell, LP 4K-C NSRB Support, LP 4K-C (including Advisors)
A. J. Scales, MOB 2M-WBN J. E. Semelsberger, EQB 2W-WBN M. D. Skaggs, ADM 1V-WBN J. F. Tortora, Jr., ADM 1B-WBN E. J. Vigluicci, WT 6A-K K. W. Whittenburg, SP 2B-C Sequoyah Licensing Files, OPS 4C-SQN EDMS, WT 3B-K M:\SUBMIT\Cycle 8 50.59 Report Supplement.pdf
ENCLOSURE WATTS BAR NUCLEAR PLANT UNIT 1 10 CFR 50.59
SUMMARY
REPORT
- 1. SA-SE Number: WBPLMN-07-008-1
- 2. SA-SE Number: WBPLMN-05-026-0
- 3. SA-SE Number: WBPLEE-07-022-1
- 4. SA-SE Number: WBPLMN-07-014-0
- 5. SA-SE Number: WBPLMN-08-006-0
- 6. SA-SE Number: WO 07-814239-000
- 7. SA-SE Number: WBPLMN-08-004-0
- 8. SA-SE Number: WBPLMN-08-005-0 E-1 of 9
SA-SE Number: WBPLMN-07-008-1 Implementation Date: 03/29/2007 Document Type: Affected Documents:
Title:
Temporary Temporary Alteration 1 Post-Accident Safety Function of Alteration Change 0002-065, R1 Emergency Gas Treatment Form (TACF) FSAR Change Package System (EGTS) 1915 TS Bases change package 2007-02 Description and Safety Assessments:
The following issue was documented in TVAs corrective action program as a problem evaluation report (PER):
Constant Operator monitoring will be required if, following an accident, the containment isolation phase A (CIA) signal is reset prior to annulus differential pressure decreasing to the setpoint needed for the EGTS pressure control loops isolation valves to open. EGTS will not perform its safety function without these valves opened as required. This was identified by Operators during simulator exercises and has been identified as a Priority 1 Operator Work Around.
The EGTS consists of two separate control loops. The hand switches of the control loops can be placed in the A-AUTO or A-AUTO STANDBY positions. The annulus pressure setpoint for the A-AUTO STANDBY position is -0.812 inches water gauge (wg). If a large break Loss of Coolant Accident (LOCA) occurs, the annulus pressure will increase from -5 inches wg to approximately -1 inch wg due to effects from the LOCA. The annulus pressure will continue to increase to the -0.812 inches wg setpoint since neither the annulus vacuum control system nor EGTS is operating to reduce the annulus pressure; i.e. the annulus vacuum control system is turned off by a safety injection system signal initiated subsequent to a LOCA. When the -0.812 inches wg EGTS setpoint is reached, both EGTS control loops will operate to filter the airborne vapors and particulates in the annulus. The off site and main control room doses will remain less than the 10 CFR 50, Appendix A, GDC 19 control room limits and the 10 CFR 100 offsite dose limits.
If a small break LOCA occurs, the annulus pressurization effects will not be as pronounced as with a large break LOCA. In this event, the annulus pressure may remain less than the -0.812 inches wg setpoint until the CIA signal is reset by operations. When the CIA signal is reset, the EGTS control loops in the A-AUTO STANDBY position are disabled. In this case, no EGTS train will operate.
Consequently, the EGTS will not function as described in the Updated Final Safety Analysis Report (UFSAR) and the EGTS safety functions will not be performed.
The primary safety function of the EGTS is to maintain the offsite and Main Control Room (MCR) doses less than the allowable 10 CFR 100 and 10 CFR 50, Appendix A General Design Criteria 19 limits during a LOCA. This safety function is accomplished during a LOCA by (1) keeping the annulus pressure below the outside atmospheric pressure and (2) removing airborne particulates and vapors that may contain radioactive nuclides from air drawn from the annulus and exhausted to the atmosphere. These safety functions will be accomplished with implementation of the TACF corrective actions assuming a credible failure of one train of EGTS exhausting to the outside environment for the duration of a LOCA accident.
These changes will increase the reliability because active components in the auxiliary relay circuit that are currently required to open the valves will be eliminated. No failure assumptions are changed, and no accident frequencies are increased. Specific changes addressed in this TACF do not create the possibility of a malfunction with different than currently evaluated results. Additionally, no fission product barriers are affected.
E-2 of 9
SA-SE Number: WBPLMN-05-026-0 Implementation Date: 06/25/2007 Document Type: Affected Documents:
Title:
Engineering EDC 51619-A Main Steam Valve Vault Airflow Document Change Rates (EDC)
Description and Safety Assessments:
The airflow rates of the exhaust fans serving the Main Steam Valve Vaults (MSVVs) are outside of the
+/-10% requirement stipulated by General Specification G-37. The fans are designed to maintain an adequate temperature environment for the main steam safety related valves, but are non-safety related. Design Change Notice (DCN) M18106-A added ductwork to the suction side of these fans, which caused the fan static pressures to be considerably increased. Because of this, air flow rates were significantly reduced. During the periodic tests, the fan speeds were increased to meet the design airflow rates.
Problem Evaluation Report (PER) 03-006901-000 revealed that the fan speeds were higher than those permitted in Technical Instruction-5.002, Flow Testing of Ventilation Systems. Further investigation revealed that the fan speed limits provided in Table 1 of the Technical Instruction may exceed the structural classification limits of the MSVV exhaust fan housings. This was confirmed through a review of the contract documentation and discussions with the manufacturer. It was advised to limit the fan speed to 535 rpm.
Temperature data was collected for a limited time period within the WBN MSVVs to justify the acceptability of the reduced total flow rates. In the north MSVV, temperature values remained unchanged, so the reduced exhaust fans test flow rates are considered acceptable. In the south MSVV, hot spot locations were generally greater than current values. These changes were updated via Drawing Change Authorization-51619-002. The hot spot changes do not affect the Minimum, Normal Average, Normal Maximum and/or Abnormal Maximum temperatures. Quality of life of components was re-evaluated based on the new hot spot temperatures. The changes required were as follows: 1) Four cables in binder WBNEQ-CABL-049 reduced their qualified life span from 34.5 years to 25.2 years. 2) The qualified life for the o-ring seals within the motor operated valves was changed from 35.39 years to 30.47 years. 3) Solenoid valve life span was reduced from 30.49 years to 27.32 years. 4) EQ Binder WBNEQ-IZS-003 was revised from 9.17 years to 8.17 years, EQ Binder WBNEQ-IZS-007 was revised from 7.94 years to 7.06 years, and the qualified life for certain limit switches listed in EQ Binder WBNEQ-IZS-005 was revised from 7.76 years to 7.05 years and from 9.89 years to 7.67 years.
The proposed changes do not introduce the possibility of a change in the frequency of an accident and no new failure modes are introduced. Operation of the MSVV exhaust fans within the acceptable limits based on the manufacturers specifications inherently reduces the possibility of their malfunction, and therefore this change will decrease the likelihood of a malfunction. Based on the analyses, the Heating Ventilation Air Conditioning system will still perform its intended design basis functions. No changes are made that would increase the consequences of any existing postulated malfunction, and the reduction in air flow rates does not have the potential to create a new accident.
No fission product barriers are unduly challenged due to this change.
E-3 of 9
SA-SE Number: WBPLEE-07-022-1 Implementation Date: 10/30/2007 Document Type: Affected Documents:
Title:
Design Change DCN 52233-A Tornado Watch/Warning Notice (DCN)
Final Safety Analysis Report (FSAR) Change Pkg. 1919 Description and Safety Assessments:
During the Component Design Basis Inspection (CDBI), the NRC identified a potential problem with the low temperature switches associated with the Diesel Generator (DG) room exhaust fans and the requirement that the fans be energized and running during a tornado watch/warning condition. Since the low temperature switches have a re-set setpoint value of 73°F and a setpoint of 68°F, the potential exists that attempts to start the fans and open the associated dampers may be unsuccessful, or that the fans may start initially, but later stop and close the dampers when they are required to be open. If the temperature switches have cooled down below their setpoint value of 68° F, the switches will prevent manual start of the fans until the re-set setpoint value of 73°F is reached. Problem Evaluation Report (PER) 120005 was initiated to document the issue and establish the required corrective actions. As part of the immediate actions for the PER, Abnormal Operating Instruction (AOI) 8, Tornado Watch or Warning, was revised to require installation of a jumper wire in each of the DG room exhaust fan Motor Control Center panels during a tornado watch. This was done in accordance with a Temporary Alteration (TA) created by a technical procedure per Site Programs and Procedures-9.5, Temporary Alterations. DCN 52233-A supersedes this TA by installing tornado bypass hand switches for each of the exhaust fans as explained below.
The design change is being issued to install a new tornado bypass hand switch in the control circuit for each of the diesel generator room exhaust fans in order to bypass the low temperature switch function during a tornado watch or warning in accordance with AOI-8. During normal operating conditions, the new hand switch will be placed in the NORMAL position allowing the low temperature switch to control their respective fan, as necessary, in order to maintain the diesel generator room exhaust temperature between 50°F and 120°F. In response to the requirements being added to Section 5.6 of System Description N3-30DB-4002 and captured in AOI-8, the new tornado bypass hand switch will be placed in the BYPASS position to allow the DG room exhaust fans to start and cause the associated inlet and exhaust dampers to open at temperatures less than the 68°F setpoint during tornado conditions.
If the outside temperature is expected to drop below 50°F during a tornado warning, then a start of each DG is required to be performed to ensure DG functionality. Monitoring the outside air temperature is required during tornado warnings to assure the DGs are started prior to a drop in temperature below 50°F. When tornado conditions are cancelled, the exhaust fans are required to be turned off and the hand switches will be returned to NORMAL promptly, and independently verified in accordance with AOI-8. Based on the evaluation of historical data, it is very unlikely that a tornado will occur with a temperature concurrent or lower than 50°F. Therefore, this change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR.
The change supports successful completion of AOI-8 actions to open each DG room exhaust fan supply and dampers and start each exhaust fan. The change does not result in new accidents or malfunctions, and it does not increase the frequency or consequences of accidents or malfunctions evaluated in the UFSAR. Additionally, no fission product barriers are challenged by the change.
E-4 of 9
SA-SE Number: WBPLMN-07-014-0 Implementation Date: 03/13/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52220-A ABI/CVI Intertie Notice (DCN)
Description and Safety Assessments:
DCN 52220-A modifies the Train A and Train B electrical circuits for high radiation in the refueling area logic bus and the Train A and Train B Solid State Protection System (SSPS) input for high radiation in the containment purge air exhaust, which initiates Containment Vent Isolation (CVI). The modification to these circuits allows WBN to maintain the Auxiliary Building Secondary Containment Enclosure (ABSCE) in the event of a high radiation signal in either the refueling area or the containment purge exhaust during refueling operations while the containment and/or annulus is open to the auxiliary building ABSCE spaces. This change will permit, but not require, operation of containment purge when moving irradiated fuel in the Auxiliary Building with containment and/or annulus hatches, personnel hatches, or penetrations open to the ABSCE spaces.
This change does not allow the equipment hatch to be open during fuel movement inside containment since Technical Specification 3.9.4, Refueling Operations, Containment Penetrations, requires the penetration to be closed during this time.
The electrical circuit modifications are implemented in two Auxiliary Relay Racks. One hand switch for each train will be added to the appropriate rack to allow the circuits to be swapped from a normal mode to a refueling mode. During normal operation, the circuits will perform the same logic as before the modification. Prior to starting refueling operations, the hand switch will be placed into the refueling mode, and it will be returned to the normal mode prior to entering Mode 4 from Mode 5.
During the refueling mode, the ABSCE isolation valves will be closed by receiving either the current high radiation signal from the spent fuel pool accident radiation monitors or upon a CVI signal. The Containment Ventilation Isolation Valves that are currently closed by a CVI signal will also be closed upon receiving a high radiation signal from the spent fuel pool accident radiation monitors. Isolation of these valves will establish the ABSCE boundary, prevent the potential for back flow from the Shield Building exhaust vent, and ensure that radioactive releases due to fuel handling accidents in containment will be processed by the Auxiliary Building Gas Treatment System (ABGTS) if they migrate into the Auxiliary Building.
The response time requirement for the spent fuel pool radiation monitors is not affected by this modification. A new response time requirement for the containment purge monitors during refueling operations is required. The new response time requirements do not affect the calculated off site or main control room dose analyses for the fuel handling accident.
This change complies with the safety and functional requirements specified in the applicable design bases documents and does not adversely affect the performance of any safety related equipment.
The proposed modifications do not increase the frequency or likelihood of accidents or malfunctions, increase the consequences of an accident or malfunction beyond the ten percent allowed by 10CFR50.59, or create a new type of accident. The design bases for fission product barriers will not be altered or exceeded and no new methods of evaluation were used in evaluating the proposed modifications.
E-5 of 9
SA-SE Number: WBPLMN-08-006-0 Implementation Date: 04/24/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52307-A Lower Compartment Coolers Notice (DCN) Blank Off Plate Installation Description and Safety Assessments:
The design function of the Lower Compartment Coolers (LCCs), along with the Control Rod Drive Mechanism (CRDM) coolers, is to maintain acceptable temperature within the reactor building lower compartment areas for the protection of equipment and control during normal reactor operation and shutdown. Any combination of three LCCs, or any combination of two LCCs and two CRDM coolers are required for safe shutdown per 10CFR50 Appendix R to prevent containment temperatures from exceeding established Equipment Qualification profiles used in the 10CFR50.49 program.
The LCC 1D-B upper outboard coil leaks at or near the header/tube interface and cannot be repaired in place. This design change will blank off the supply and return of the Essential Raw Cooling Water (ERCW) to this cooling coil, resulting in seven of the eight coils remaining functional. Blank off plates of stainless steel that are capable of withstanding the ERCW system design pressure will be installed at the supply and return flanges associated with the leaking coil. The change will isolate ERCW flow to the offending coil only and will not adversely impact continued flow to the remaining seven coils.
The piping system will remain seismically qualified with the blanking plates installed.
If the water leak in the offending coil is allowed to continue, an increase in the relative humidity levels inside lower containment is expected. This condition could mask water leaks inside lower containment. Failure of the blank off plates is less likely to occur than the failure of the existing coil components, and the installation of the plates eliminates the potential for the coil to rupture. This change will prevent one of the coils from performing its cooling function. Calculations have been revised to note that the capacity reduction of 1/8 on the cooler will not affect the current requirements.
Any combination of three LCCs or two LCCs and two CRDM coolers remains acceptable and does not affect Technical Specification requirements.
Proper operation of the LCC cooling coils is not a safety related function and is not required for the mitigation of any Updated Final Safety Analysis Report (UFSAR) Chapter 6 or 15 accidents. The blanking plate will be designed to meet the design conditions of the ERCW system. The design basis functions of the LCC 1D-B is maintained and is not adversely impacted by this proposed design change. Therefore, this change does not result in any new accidents or malfunctions, and does not result in increased frequency or consequences of accidents or malfunctions evaluated in the UFSAR.
In addition, no fission product barriers are challenged by this change. Safety related functions of the LCCs are not affected by this proposed modification.
E-6 of 9
SA-SE Number: WO 07-814239-000 Implementation Date: 05/29/2008 Document Type: Affected Documents:
Title:
Maintenance Work Work Order 07-814239-000 Fuel Cleaning Order (WO) Areva Procedure FS-235 Description and Safety Assessments:
During Cycle 7 operation, the WBN Unit 1 core exhibited symptoms of Crud Induced Power Shift (CIPS). The root cause of this is boiling near the upper surfaces of high-powered fuel rods, leading to the depositing of crud in these locations. The Ultrasonic Fuel Cleaning System (UFC) was developed by Dominion Engineering, Inc. with Electric Power Research Institute (EPRI) to remove the corrosion products from Pressurized Water Reactor (PWR) fuel using ultrasonic energy. Laboratory and field experience has shown an average crud removal rate of greater than 80%. All UFC operations will be performed underwater in the Unit 2 side of the transfer canal at WBN. Lifting and handling of the UFC system and its support stand will be performed in accordance with NUREG 0612 to mitigate the potential for damage to stored fuel.
The existing potential for pellet-clad interaction resulting from agitation of fuel pellet fragments that were cracked prior to cleaning has been evaluated and determined to be very unlikely. Normal administrative and procedural controls will be employed to ensure fuel is not damaged. Because of this evaluation and administrative control, the risk of damaged fuel or an increased risk of a fuel handling accident is negligible. Additionally, the cleaning apparatus is not located in an area where it can fall on any safety-related equipment besides the transfer canal liner.
The increased risk of crud particle release into the Unit 2 transfer canal is mitigated by the current operating procedures that require the fuel to be secured safely and cleaning suspended until pool visibility is restored. The potential to release dissolved radioactive gasses is minimized by design.
The average flow velocity in the cleaning apparatus is sufficiently high to entrain all gas bubbles smaller than 3 mm in diameter. These bubbles are mixed vigorously in the connection hose between the cleaner and the filter skid. This mixing action causes most of the gasses to re-dissolve into the bulk fluid. Any that are not re-dissolved have been shown to not cause a significant radiological hazard.
The possibility of a fuel assembly becoming stuck in the cleaning fixture is very remote. It is protected against by the small amount of force required to straighten a 1 inch bow in a fuel assembly, which can be provided by the sidewalls of the cleaning fixture. Also, the fuel can be indefinitely cooled in the cleaning apparatus by natural convection, as previously discussed.
The Spent Fuel Cooling and Cleaning System is a safety related system required to mitigate the consequences of Design Basis Events. During operation of the UFC system, fuel assemblies are placed in the fuel cleaning chambers and bombarded with high frequency pressure oscillations. This adds negligible heat loads to the system and will not prevent the spent fuel cooling system from performing its intended design function. Additionally, the UFC system is designed such that natural circulation will cool a fuel assembly in the UFC chamber in the event of a station blackout.
All fuel handling will be performed within the bounds of Updated Final Safety Analysis Report (UFSAR) procedures and interlocks. This change does not affect the consequences of an accident or malfunction because it does not affect the dose results of the accident analyses reported in the UFSAR. Because of the normal administrative and procedural controls in place, the increase in probability of a fuel handling accident is negligible. Potential accidents that have not been previously evaluated are not introduced by this change. No fission product barriers are affected by this activity.
E-7 of 9
SA-SE Number: WBPLMN 08-004-0 Implementation Date: 03/27/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52211-A Relocation of Radiation Monitor Notice (DCN) Safety Analysis Report for Transfer Canal Fuel Handling (SAR) Change Package Accident Detection 1927 Description and Safety Assessments:
A fuel handling accident (FHA) in the fuel transfer canal may not be detected by the radiation monitors located on the wall of the spent fuel pool below the 757 floor elevation. Consequently, the Auxiliary Building Gas Treatment System (ABGTS) would not be automatically established. As a result of this condition, calculation WBNTSR-009 was revised. The revised calculation determined that such an accident can be mitigated if spent fuel pool accident radiation monitor 0-RE-90-103-B is relocated above the refueling floor elevation and the exhaust vents located around the fuel transfer canal perimeter are isolated during movement of irradiated fuel in that area. Air intake must be isolated to ensure the activity released by a FHA in the transfer canal is not removed and exhausted by the ventilation system prior to radiation monitor detection.
This DCN is only relocating the radiation monitor, not changing its design functions. The mounting method for the new location has been evaluated and has been determined to meet seismic requirements.
Additionally, this DCN includes the isolation of air flow from the transfer canal by closing either a fire damper or a balancing damper during irradiated fuel movement in the transfer canal. Alternate ventilation requirements are required to ensure the fuel handling area ventilation system remains balanced to support both normal and post-accident conditions. The DCN requires opening one or two selected fuel handling area exhaust ductwork access panels and adding a sheet metal plate to restrict air flow through one access opening as needed. The access panel(s) will be secured in place using heavy gauge wire, and stainless steel screens will be installed to prevent debris from entering the ductwork. Static pressure will be measured upstream of the access panel both before and after isolation of flow from the transfer canal. This will verify required normal or post accident air flows are maintained.
Relocation of the spent fuel pool radiation monitor will not affect the safety related operation of the monitor. The proposed changes to the fuel handling area ventilation system as described above will not adversely affect establishment of the Auxiliary Building Secondary Containment Enclosure or operation of the ABGTS during FHA or Loss of Coolant Accident (LOCA) conditions. Thus, the proposed changes do not result in an increase in frequency of occurrence of accidents or likelihood of occurrence of malfunctions evaluated in the Updated Final Safety Analysis Report (UFSAR), and do not result in an increase in the consequences of an accident or malfunction. The proposed changes do not result in an accident of a different type or a malfunction with a different result. Further, the changes do not affect fission product barriers and the method of evaluation described in the UFSAR.
E-8 of 9
SA-SE Number: WBPLMN 08-005-0 Implementation Date: 04/14/2008 Document Type: Affected Documents:
Title:
Engineering Design Change Notice Number 2 Heater Drain Bypass to Document Change (DCN) 52270-A Condenser Level Control Valves (EDC)
Description and Safety Assessments:
When the Number 2 Heater Drain Bypass to Condenser level control valves (LCVs) are opened during high turbine load conditions, violent water hammer occurs. This design change of documentation only allows for the bypass LCVs to be opened only during low power operations to protect against severe water hammer incidences.
The three bypass to condenser lines from the No. 2 heater were not part of original plant design. A condition was identified where the No. 2 heaters did not drain to the No. 3 Heater Drain Tank (HDT) during rapid load reductions at low power operating conditions. To account for this, the bypass lines were installed. The LCVs are normally in the closed position and were designed to automatically open if the heater water level reaches a prescribed high setpoint. This design change will still allow the LCVs to be opened at low power levels. The change will require the bypass lines to be isolated during normal high turbine load operation to prevent water hammer.
If maintenance is needed on the main drain valves, the main drain path bypass valves can be used to maintain No. 2 heater water level. In the event the normal drain path malfunctions during high power operation, the bypass to condenser LCVs will no longer open automatically, and a heater string isolation could occur. It was determined that the reduced feedwater temperature caused by a heater string isolation will not compromise plant safety.
This design change has the potential to increase the frequency of loss of normal feedwater if all three strings were to isolate. Isolation of main feedwater from the steam generators (SGs) would require Operations to start the Auxiliary Feedwater (AFW) system manually, or AFW will automatically start on a low low SG level. The Loss of Normal Feedwater is categorized as a fault of moderate frequency, implying once a year occurrence. Historical operating experience at WBN shows that heater drain failures do not typically occur at a frequency greater than once per year. Further, no new failure modes are introduced that could increase the frequency of occurrence of an isolation of all three intermediate pressure heaters and no new components are added with this design change.
The design basis accident analyses evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected, and the system is not credited for accident mitigation. No equipment required for safe operation or shutdown is changed by this design. No fission product barriers are challenged by this change.
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