Letter Sequence RAI |
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Results
Other: CNL-14-077, Responses to Degraded Voltage Issue Requests for Additional Information, ML102290358, ML102290360, ML102290361, ML102290362, ML102290363, ML102290364, ML102290365, ML102290366, ML102290367, ML102290368, ML102290383, ML102290384, ML102290385, ML102290386, ML102290387, ML102290388, ML102290389, ML102290390, ML102290391, ML102290392, ML102290403, ML102290404, ML102290405, ML102290406, ML102290407, ML102290408, ML102290409, ML102290410, ML102290411, ML102290414, ML102290439, ML102290440, ML102290441, ML102290442, ML102290443, ML102290444, ML102290445, ML102290446, ML102290447, ML102290448, ML102290458, ML102290459, ML102290460, ML102290461, ML102290462, ML102290463, ML102290464, ML102290465, ML102290466... further results
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MONTHYEARML1001916862010-01-11011 January 2010 Final Safety Analysis Report, Amendment 97, Section 8 - Electric Power Through 8E-12 Project stage: Request ML1005500072010-03-11011 March 2010 Request for Additional Information Related to Licensee'S Final Safety Analysis Report Amendment 95 Related to Section 4.2.2, Reactor Vessel Internal Components Project stage: RAI ML1010405732010-04-0909 April 2010 Response to NRC Request for Additional Information Regarding Licensee'S Final Safety Analysis Report (FSAR) Amendment Related to Section 4.2.2, Reactor Vessel Internal Components. Project stage: Response to RAI ML1011303512010-05-0505 May 2010 Request for Withholding Information from Public Disclosure (Tac No. ME2731) Project stage: Withholding Request Acceptance ML1014500842010-06-23023 June 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Reactor Systems, Nuclear Performance and Code Review, Ansd Plant Systems Project stage: RAI ML1015402502010-06-24024 June 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Nuclear Performance and Code Review, Plant Systems, and Testing (Tac Nos. ME2731 and ME3091) Project stage: RAI ML1017903992010-06-28028 June 2010 Request for Additional Information (RAI) Regarding Licensee'S Final Safety Analysis Report Amendment Related to Section 4.2.2, Reactor Vessel Internal Components Project stage: Request ML1016200062010-06-29029 June 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Nuclear Performance and Code Review (Tac No. ME2731) Project stage: RAI ML1015304742010-07-0202 July 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Mechanical and Civil Engineering Systems Project stage: RAI ML1016000262010-07-0808 July 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report, Amendment Related to Quality and Vendor Branch Review Project stage: RAI ML1015303542010-07-12012 July 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Electrical Engineering Systems Project stage: RAI ML1018001562010-07-15015 July 2010 Request for Additional Information Regarding Licensee'S Final Safety Analysis Report Amendment Related to Chapter 8, Electric Power Project stage: RAI ML1023003272010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-4A (Dwg No. 1-47W845-5, R38) Project stage: Other ML1022904652010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-2 (Dwg No. 1-47W845-2, R76) Project stage: Other ML1023003282010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-4B (Dwg No. 1-47W845-7, R14) Project stage: Other ML1023003292010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-5 (Dwg No. 1-47W611-67-1, R8) Project stage: Other ML1023003302010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-6 (Dwg No. 1-47W611-67-2, R5) Project stage: Other ML1023003312010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-8 (Dwg No. 1-47W611-67-4, R5) Project stage: Other ML1023003322010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-9 (Dwg No. 1-47W611-67-5, R10) Project stage: Other ML1023003332010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-10 (Dwg No. 1-47W610-67-1, R27) Project stage: Other ML1023003342010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-10 Sh a (Dwg No. 1-47W610-67-1A, R16) Project stage: Other ML1023003352010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-11 (Dwg No. 1-47W610-67-2, R15) Project stage: Other ML1023003432010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-12 Sh a (Dwg No. 1-47W610-67-3A, R7) Project stage: Other ML1023003442010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-13 (Dwg No. 1-47W610-67-4, R17) Project stage: Other ML1023003452010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-11 Sh a (Dwg No. 1-47W610-67-2A, R8) Project stage: Other ML1023003462010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-12 (Dwg No. 1-47W610-67-3, R12) Project stage: Other ML1023003472010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-14 (Dwg No. 1-47W610-67-5, R14) Project stage: Other ML1023003482010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-14 (Dwg No. 1-47W610-67-5A, R2) Project stage: Other ML1023003492010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-16 (Dwg No. 1-47W859-4, R23) Project stage: Other ML1023003502010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-17 (Dwg No. 1-47W859-3, R18) Project stage: Other ML1023003512010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-18 (Dwg No. 1-47W859-2, R36) Project stage: Other ML1023003522010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-19 (Dwg No. 1-47W859-1, R47) Project stage: Other ML1023003602010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-23 (Dwg No. 1-47W611-70-1, R9) Project stage: Other ML1023003612010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-20 (Dwg No. 1-47W610-70-1, R23) Project stage: Other ML1023003622010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-21 (Dwg No. 1-47W610-70-2, R29) Project stage: Other ML1023003632010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-21A (Dwg No. 1-47W610-70-2A, R15) Project stage: Other ML1023003642010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-22 (Dwg No. 1-47W610-70-3, R17) Project stage: Other ML1023003652010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-24 (Dwg No. 1-47W611-70-2, R11) Project stage: Other ML1023003672010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-25 (Dwg No. 1-47W611-70-3, R4) Project stage: Other ML1023003682010-08-0505 August 2010 Final Safety Analysis Report Figure 9.2-25A (Dwg No. 1-47W611-70-4, R4) Project stage: Other ML1023003692010-08-0505 August 2010 Final Safety Analysis Report Figure 9.3-1 (Dwg No. 1-47W610-32-1, R15) Project stage: Other ML1022903652010-08-0505 August 2010 Final Safety Analysis Report Figure 8.3-7 (Dwg No. 1-45W760-211-9, R16) Project stage: Other ML1023003972010-08-0505 August 2010 Final Safety Analysis Report Figure 9.5-25A (Dwg No. 1-47W610-82-2, R7) Project stage: Other ML1023003982010-08-0505 August 2010 Final Safety Analysis Report Figure 9.5-25B (Dwg No. 1-47W610-82-3, R7) Project stage: Other ML1023003992010-08-0505 August 2010 Final Safety Analysis Report Figure 9.5-25C (Dwg No. 1-47W610-82-4, R7) Project stage: Other ML1022903682010-08-0505 August 2010 Final Safety Analysis Report Figure 8.3-14B (Dwg No. 1-45W760-82-1, R17) Project stage: Other ML1022903672010-08-0505 August 2010 Final Safety Analysis Report Figure 8.3-9 (Dwg No. 1-45W760-211-11, R13) Project stage: Other ML1022903662010-08-0505 August 2010 Final Safety Analysis Report Figure 8.3-8 (Dwg No. 1-45W760-211-10, R12) Project stage: Other ML1022903622010-08-0505 August 2010 Final Safety Analysis Report Figure 8.1-3 (Dwg No. 1-45W700-1, R28) Project stage: Other ML1022903582010-08-0505 August 2010 Final Safety Analysis Report Figure 8.1.2 (Dwg No. 1-15E500-1, R31) Project stage: Other 2010-06-28
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Category:Letter
MONTHYEARML24219A0262024-08-12012 August 2024 Request for Withholding Information from Public Disclosure IR 05000390/20240022024-08-0707 August 2024 Integrated Inspection Report 05000390/2024002 and 05000391/2024002 Rev ML24204A2652024-07-25025 July 2024 Regulatory Audit Summary Related to Request to Revise Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24170A8002024-07-15015 July 2024 Issuance of Amendment Nos. 168 and 74 Regarding Revision to Technical Specification Table 1.1-1 for Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts IR 05000390/20244402024-07-12012 July 2024 95001 Supplemental Inspection Supplemental Report 05000390-2024440 and 05000391-2024440 and Follow-Up Assessment Letter ML24131A0012024-07-0202 July 2024 Issuance of Amendment Nos. 167 and 73 Regarding Adoption of Technical Specification Task Force Traveler TSTF-427-A, Revision 2 CNL-24-052, Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-06-27027 June 2024 Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-24-018, License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS2024-06-25025 June 2024 License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc ML24100A7642024-05-16016 May 2024 Issuance of Amendment No. 166 Regarding Revision to Technical Specification 3.8.2, AC Sources-Shutdown, to Remove Reference to C-S Diesel Generator (CNL-23-062) IR 05000390/20240012024-05-14014 May 2024 Integrated Inspection Report 05000390/2024001 and 05000391/2024001 CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000391/20240072024-04-30030 April 2024 Assessment Follow-up Letter for Watts Bar Nuclear Plant, Unit 2 – Report 05000391/2024007 ML24120A1182024-04-29029 April 2024 – Notification of NRC Supplemental Inspection (95001) and Request for Information CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 ML24087A1912024-04-18018 April 2024 Exemption from Select Requirements of 10 CFR Part 73, Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-24-010, License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19)2024-04-17017 April 2024 License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19) CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update ML24072A0052024-04-15015 April 2024 Issuance of Amendment Nos. 165 and 72 Regarding Increase in the Maximum Number of Tritium Producing Burnable Absorber Rods and Supporting Changes, and Revision to the Updated Final Safety Analysis Report CNL-24-004, Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13)2024-04-0404 April 2024 Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13) IR 05000390/20244012024-04-0202 April 2024 – Security Baseline Inspection Report 05000390/2024401 and 05000391/2024401 - (Public) CNL-24-020, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements2024-04-0101 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements CNL-24-008, Guarantee of Payment of Deferred Premiums - 2023 Annual Report2024-03-27027 March 2024 Guarantee of Payment of Deferred Premiums - 2023 Annual Report CNL-24-007, Annual Insurance Status Report2024-03-27027 March 2024 Annual Insurance Status Report CNL-24-025, Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule2024-03-25025 March 2024 Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule ML24081A0262024-03-21021 March 2024 Emergency Plan Implementing Procedure Revisions ML24079A0312024-03-19019 March 2024 Wb 2024-301, Corporate Notification Letter (210-day Ltr) CNL-24-031, Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-03-18018 March 2024 Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance CNL-24-029, Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 372024-03-14014 March 2024 Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 37 CNL-24-028, Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2024-03-14014 March 2024 Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000390/20240102024-03-0808 March 2024 Age-Related Degradation Inspection Report 05000390/2024010 and 05000391/2024010 CNL-24-012, Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-02-28028 February 2024 Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance IR 05000390/20230062024-02-28028 February 2024 Annual Assessment Letter for Watts Bar Nuclear Plant Units 1 and 2 - Report 05000390/2023006 and 05000391/2023006 CNL-24-023, Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 252024-02-20020 February 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 25 IR 05000390/20230042024-02-13013 February 2024 Integrated Inspection Report 05000390/2023004 and 05000391/2023004 and Apparent Violation ML24038A3342024-02-0707 February 2024 Change to South Texas Project Electric Generating Station (Stepgs) Emergency Plan ML23319A2452024-01-29029 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) 2024-08-07
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML24155A1372024-05-29029 May 2024 Email from K. Green to S. Hughes Request for Additional Information Related to License Amendment Request to Revise Residual Heat Removal Flow Rate ML24120A1182024-04-29029 April 2024 – Notification of NRC Supplemental Inspection (95001) and Request for Information ML24116A2012024-04-17017 April 2024 Nrctva ISFSI CBS (RFI) ML24045A0312024-02-14014 February 2024 NRR E-mail Capture - Request for Additional Information Related to the Exemption Request for the 10 CFR Part 73 Enhanced Weapons Rule ML23166A1142023-06-15015 June 2023 Document Request for Watts Bar Nuclear Plant - Radiation Protection Inspection - Inspection Report 2023-03 ML23067A2372023-03-0808 March 2023 WB_2023-02_RP_inspection_doc_request ML23030A3512023-01-25025 January 2023 Notification of Watts Bar Nuclear Plant - Design Bases Assurance Inspection (Programs) and Initial Information Request ML22343A0692022-12-0808 December 2022 NRR E-mail Capture - Request for Additional Information - Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, License Amendment Request to Revise Technical Specification 3.4.12 (L-2022-LLA-0103) ML22227A0272022-08-11011 August 2022 NRR E-mail Capture - Request for Additional Information Related to Alternative Requests RP-11 for Sequoyah Nuclear Plant, Units 1 and 2, and IST-RR-9 for Watts Bar Nuclear Plant, Units 1 and 2 ML22144A1002022-05-12012 May 2022 NRR E-mail Capture - Request for Additional Information Related to Tva'S Request to Revised the TVA Plants' Radiological Emergency Plans ML22115A1402022-04-25025 April 2022 NRR E-mail Capture - Requests for Confirmation of Information and Additional Information Regarding Watts Bar Nuclear Plant, Unit 2 Exemption Request Re 10 CFR Part 26 (L-2022-LLE-0017) ML22083A2372022-03-24024 March 2022 NRR E-mail Capture - Request for Additional Information and Confirmation of Information Related to Tva'S Request for Changes to Watts Bar Nuclear Plant, Units 1 and 2, Technical Specification 3.7.8 ML22056A3802022-02-25025 February 2022 Document Request for Watts Bar Nuclear Plant - Radiation Protection Inspection - Inspection Report 2022-02 ML21267A1392021-09-23023 September 2021 Document Request for Upcoming RP Inspection at Watts Bar ML21221A2602021-08-0909 August 2021 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Revise Watts Bar, Unit 1 Tech Specs Related to Continuous Opening of the Auxiliary Building Secondary Containment Enclosure Boundary ML21102A1312021-04-19019 April 2021 Request for Withholding Information from Public Disclosure for Watts Bar Nuclear Plant, Unit 1 ML21095A0422021-04-0202 April 2021 NRR E-mail Capture - Added Clarification to RAI 2 for Thot LAR ML21095A0402021-04-0202 April 2021 NRR E-mail Capture - Request for Additional Information Re Generic Letter 95-05 90-Day Report and LAR to Adjust Growth Rate for Thot (EPIDs L-2021-LRO-0003 and L-2021-LLA-0026) ML21095A0412021-04-0101 April 2021 NRR E-mail Capture - Revised Draft RAI - Combined RAI Set for Watts Bar Unit 2 90-Day Report and Thot LAR (EPIDs L-2021-LRO-0003 and L-2021-LLA-0026) ML21095A0442021-04-0101 April 2021 NRR E-mail Capture - Revised Draft RAI - Combined RAI Set for Watts Bar Unit 2 90-Day Report and Thot LAR (EPIDs L-2021-LRO-0003 and L-2021-LLA-0026) ML21091A0772021-04-0101 April 2021 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Revise Technical Specification 5.7.2.19, Containment Leakage Rate Testing Program ML21095A0462021-03-22022 March 2021 NRR E-mail Capture - Draft Request for Additional Information Regarding Tva'S Generic Letter 95-05 90-Day Report for Watts Bar Unit 2 ML21039A6402021-02-0808 February 2021 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Revise the Watts Bar Nuclear Plant, Unit 1 Technical Specifications Related to Steam Generator Tube Inspection Frequency ML21012A2032021-01-11011 January 2021 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Revise the Watts Bar UFSAR to Use Alternate Probability of Detection ML20350B5592020-12-15015 December 2020 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Adopt Traveler TSTF-490 Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec ML20338A3202020-12-0303 December 2020 Notification of an NRC Fire Protection Team Inspection (NRC Inspection Report 05000390/2021011 and 05000391/2021011) and Request for Information ML20322A4412020-11-17017 November 2020 NRR E-mail Capture - Request for Additional Information Regarding Tva'S Request to Revise TS 3.7.11 Related to the MCR Chiller Replacement ML20322A4392020-11-0505 November 2020 NRR E-mail Capture - Draft Request for Additional Information Regarding Tva'S Request to Revise TS 3.7.11 Related to the MCR Chiller Replacement ML20308A3512020-11-0202 November 2020 Request for Additional Information on WBN Request for Exemption from 10 CFR Part 73, Appendix B, Section VI for the Conduct of an Annual Force-on-Force Exercise (EPID L-2020-LLE-0165 (COVID-19)) ML20253A1782020-09-0909 September 2020 Emergency Preparedness Program Inspection Request for Information ML20266G4592020-08-14014 August 2020 Notification of Inspection and Request for Information ML20196L8622020-07-14014 July 2020 NRR E-mail Capture - Watts Bar Nuclear Plant, Units 1 and 2 - Request for Additional Information Regarding Request to Implement the Full Spectrum LOCA Methodology ML20086G4802020-03-26026 March 2020 NRR E-mail Capture - Request for Additional Information for WBN2 Request Measurement Uncertainty Recapture Power Uprate (L-2019-LLS-0000) - Part 2 ML20085G3572020-03-25025 March 2020 NRR E-mail Capture - Request for Additional Information for WBN2 Request Measurement Uncertainty Recapture Power Uprate (L-2019-LLS-0000) ML20084M1942020-03-24024 March 2020 NRR E-mail Capture - Request for Additional Information for WBN2 Request Measurement Uncertainty Recapture Power Uprate (L-2019-LLS-0000) ML20083J3952020-03-12012 March 2020 NRR E-mail Capture - Draft Request for Additional Information for WBN2 Request Measurement Uncertainty Recapture Power Uprate (L-2019-LLS-0000) ML19340A6842019-12-0505 December 2019 NRR E-mail Capture - Request for Additional Information for WBN2 Request for One-Time Extension of Completion Time for TS 3.7.8 (L-2019-LLA-0020) ML19218A0302019-08-0505 August 2019 NRR E-mail Capture - Watts Bar Nuclear Plant - Second-Round Request for Additional Information Related to Application to Revise Technical Specifications Regarding DC Electrical Systems, TSTF-500, Revision 2 ML19218A0282019-07-25025 July 2019 NRR E-mail Capture - Watts Bar Nuclear Plant - Draft Second-Round Request for Additional Information Related to Application to Revise Technical Specifications Regarding DC Electrical Systems, TSTF-500, Revision 2 ML19186A4352019-07-0505 July 2019 NRR E-mail Capture - Watts Bar Nuclear Plant - Correction to Final Request for Additional Information Related to Application to Adopt 10 CFR 50.69 ML19169A3592019-06-18018 June 2019 NRR E-mail Capture - Watts Bar Nuclear Plant - Final Request for Additional Information Related to Application to Adopt 10 CFR 50.69 ML19148A7912019-05-28028 May 2019 NRR E-mail Capture - Sequoyah Nuclear Plant and Watts Bar Nuclear Plant - Final Request for Additional Information Related to Request for Alternative to OM Code Requirements ML19106A0462019-04-15015 April 2019 NRR E-mail Capture - Watts BAR, Units 1 and 2 Request for Additional Informatin (RAI) Regarding Changes to Technical Specifications Sections 3.8.1, 3.8.7, 3.8.8, and 3.8.9 ML19071A3542019-03-0808 March 2019 NRR E-mail Capture - Watts Bar Nuclear Plant - Final Request for Additional Information Related to Request to Adopt TSTF-425 to Relocate Specific Surveillance Frequency Requirements to Licensee-Controlled Program ML18313A2202018-11-0707 November 2018 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection ML18282A6372018-10-0909 October 2018 NRR E-mail Capture - RAIs (Final) - LAR to Revise the Steam Generator Technical Specifications for Watts Bar Nuclear Plant, Unit 2 ML18270A2362018-09-26026 September 2018 NRR E-mail Capture - Watts Bar Units 1 and 2 RAIs - Modify TS 3.8.9 Completion Time for Inoperable 120V AC Vital Buses (L-2018-LLA-0050) ML18240A0702018-08-27027 August 2018 NRR E-mail Capture - RAI for Watts Bar Unit 2 Tpbars LAR and Watts Bar Units 1 and 2 LAR Related to Fuel Storage ML18199A1822018-07-17017 July 2018 NRR E-mail Capture - Request for Additional Information Regarding Watts Bar Unit 1 Extension of Surveillance Requirement Intervals ML18123A3942018-05-0303 May 2018 NRR E-mail Capture - Request for Additional Information Unit 2 Cycle 1 Steam Generator Tube Inspection Report 2024-05-29
[Table view] |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 July 2, 2010 Mr. Ashok S. Bhatnagar Senior Vice President Nuclear Generation Development and Construction Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SUB..WATTS BAR NUCLEAR PLANT, UNIT 2 -REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSEE'S FINAL SAFETY ANALYSIS REPORT AMENDMENT RELATED TO MECHANICAL AND CIVIL ENGINEERING SYSTEMS (TAC NO. ME2731)
Dear Mr. Bhatnagar:
By letter dated January 11, 2010 (NRC Agencywide Document Access and Management System Accession No. ML 100191686), to the U.S. Nuclear Regulatory Commission (NRC), the Tennessee Valley Authority provided an update (Amendment No. 97) to the Final Safety Analysis Report (FSAR) for the Watts Bar Nuclear Plant (WBN), Unit 2. That update contained changes to a number of sections of the WBN Unit 2 FSAR, including Sections 3.9.1, 3.9.2, 3.9.3, and 5.5.1. The NRC staff has reviewed these four sections and has identified additional information that is needed to complete the technical review of the operating license application.
A response is required within 30 days of receipt of this letter. If you should have any questions, please contact me at 301-415-1457.
Sincerely, oel S. Wiebe, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.
Request for Additional cc w/encl: Distribution via REQUEST FOR ADDITIONAL INFROMATION WATTS BAR NUCLEAR PLANT, UNIT 2 FINAL SAFETY ANALYSIS REPORT AMENDMENT NOS. 95, 96, AND 97 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 By letter dated January 11, 2010 (NRC Agencywide Document Access and Management System Accession No. ML 100191686), to the U.S. Nuclear Regulatory Commission (NRC), the Tennessee Valley Authority (TVA) provided an update (Amendment
- 97) to the Final Safety Analysis Report (FSAR) for the Watts Bar Nuclear Plant (WBN), Unit 2. This update contained changes to a number of sections of the WBN Unit 2 FSAR, including Sections 3.9.1, 3.9.2, 3.9.3, and 5.5.1. The NRC staff has reviewed these four sections and has identified additional information that is needed to complete the technical review of the operating license application.
EMCB Request for Additional Information (RAn 3.9-1 The NRC staff noted a number of instances in the review of Sections 3.9.1, 3.9.2, 3.9.3 and their corresponding tables and figures of Amendment No. 97 to the WBN Unit 2, Final Safety Analysis Report (FSAR) (Reference
- 1) where editorial modifications may be necessitated in subsequent revisions to the WBN Unit 2 FSAR. Please review the following NRC staff notations and rectify, as necessary. On page 3.9-18 of Reference 1, continuing to page 3.9-19, the first two paragraphs of Section 3.9.2.5.6, "Results and Acceptance Criteria," are duplicates of the first two paragraphs of the following section (3.9.2.5.7), also titled "Results and Acceptance Criteria." On page 3.9-36 of Reference 1, superfluous spaces exist between the word "Table" and "3.9-17." On page 3.9-44 of Reference 1, the primary membrane plus primary bending stress limit should be "1.1 S" versus the current "1.1.S." On page 3.9-63 of Reference 1, the title of Table 3.9-5 should be revised to state that the limits are "Maximum Deflections" versus the current wording of "Maximum Defections." On page 3.9-63 of Reference 1, Note 1 references Westinghouse Commercial Atomic Power (WCAP)-5890 with a corresponding superscript of number 21, indicating that this refers to Reference
- 21. Page 3.9-58 of Reference 1 indicates that this WCAP report is Reference 22, not Reference
- 21. If this is not erroneous, please provide additional justification in conjunction with RAI 3.9.2-3 below. Enclosure
-2On page 3.9-77 of Reference 1, the third note corresponding to Table 3.9-16 should be revised to correct the misspelling of "Non-pressure" and "other justifiable" versus the current wording of "Non-pressur" and "othe justifiable." EMCB RAI 3.9.1-1 In Supplemental Safety Evaluation Report (SSER) 6 (Reference 3), the NRC staff noted that the licensee's piping evaluation for a postulated main feedwater header rupture transient, which results in a water hammer event due to a rapid check valve closure, included an assumption that certain feedwater piping system supports failed when the loads exceeded their calculated capacities; this was listed as an open item in SSER 6 (tracked as Outstanding Issue 20(a)). In SSER 13 (Reference 6), the staff noted that the analyses performed, which postulated pipe support failures, was acceptable based on the difficulty involved with making subsequent pipe support modifications and the low probabilistic nature involved with the water hammer transient.
Additionally, as part of the closure of this open item, SSER 13 also included a copy of a report performed by Brookhaven National Laboratory (BNL) regarding this issue. BNL was contracted by the NRC to evaluate the licensee's piping analyses performed to demonstrate compliance with the criteria of Appendix F of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code. BNL concluded that the licensee's piping analyses performed for the feedwater loops inside containment were sufficient and demonstrated that the piping system would maintain its structural integrity when subjected to the dynamic loading associated with the water hammer event. Please describe the applicability of the conclusions made by the NRC staff and the contractor (BNL) regarding the piping analyses described above as they relate to the current WBN Unit 2 refurbishment efforts. Please indicate whether the same issues exist with the inability to modify certain piping supports within containment and whether the piping analyses for the WBN Unit 2 feedwater loops are the same as those analyses performed in support of WBN Unit 1. If these analyses are dissimilar, please summarize and provide justification for any portions of the analyses that are not exactly the same and whether the results of these dissimilar analyses demonstrate that the feedwater piping loops meet the acceptance criteria of the code of record for this piping system. EMCB RAI 3.9.2-1 In Section 3.9.2.3 of Reference 1, it is indicated that Sequoyah Nuclear Plant Unit 1 and Trojan Nuclear Power Plant (Trojan) "...have been instrumented to provide prototype data applicable to Watts Bar" for the purposes of evaluating the flow induced oscillatory pressure effects on the reactor vessel internals.
Additionally, it is concluded, based on scale model test results and "...preliminary results from Trojan... ," that plants with neutron shielding pads exhibit less core barrel vibration than plants with thermal shields. Based on the fact that Trojan ceased operations in the year 1 992, please discuss the applicability of the statements above, which are currently included in Reference
- 1. If these data was captured during Trojan's operational state, please describe how this operating experience has been applied to the design or operational characteristics of any of the reactor vessel internals.
Additionally, please indicate whether additional results, other than the "preliminary results" mentioned in Reference 1, were utilized to provide additional information regarding the comparison between plants with neutron shielding pads and plants with thermal shields as they relate to core barrel excitation.
-3 EMCB RAI The analyses methods described in Section 3.9.2.5 of Reference 1, "Dynamic System Analysis of the Reactor Internals Under Faulted Conditions," were approved for use by a previous license amendment request submitted for WBN Unit 1. These methods incorporate the use of the MULTIFLEX, LATFORCE, FORCE-2 and WECAN computer codes to model the complex, nonlinear thermal-hydraulic loadings induced on the reactor vessel internals under upset loading conditions.
Please confirm that the inputs used to analyze these conditions for WBN Unit 2 are the same inputs as those used to analyze the loadings induced on the WBN Unit 1 reactor vessel internals.
If any variances exist between the WBN Unit 1 and WBN Unit 2 inputs for these codes, including primary and secondary loadings, flow parameters, mass models, finite element formulations, or other input parameters, provide justification for the variation and its effects on the ability of the WBN Unit 2 reactor vessel internals to meet the acceptance criteria provided in Table 3.9-5. Additionally, please clarify whether the references to "Watts Bar Unit 1" on pages 3.9-15, 3.9-19, and 3.9-20 (2) are correctly referring to WBN Unit 1 for purposes of comparing analyses or whether these instances are incorrect (i.e., these references should state WBN Unit 2 and not WBN Unit 1). EMCB RAI 3.9.2-3 Table 3.9-5 of Reference 1, "Maximum Def[I]lections Under Design Basis Event (in)," provides the maximum allowable and no loss-of-function limits for the reactor vessel internals under design basis loading conditions.
Note 1 to Table 3.9-5 indicates that WCAP-5890 provides limiting criteria for internals deflection based on stress levels induced in the internals structures.
Please discuss whether the acceptance criteria provided in Table 3.9-5 are based on WCAP-5890.
If these criteria are based on this WCAP report, please provide the bases for the regulatory acceptance of this report. If these criteria are based on a methodology other than the WCAP report, please provide additional information regarding the development of these deflection limits and the bases for the regulatory acceptance of this alternate methodology.
EMCB RAI 3.9.2-4 Please provide justification for the variance between the WBN Unit 1 and WBN Unit 2 allowable and no loss-of-function deflection limits as this variance relates to the upper barrel expansion and compression limits and the no loss-of-function limit for the upper package axial deflection.
This justification should include information regarding whether there are variations in the analyses methodologies for determining the WBN Units 1 and 2 reactor vessel internals faulted loads (as requested in RAI 3.9.2-2).
Additionally, this justification should indicate whether there are variations in the acceptance criteria for the WBN Units 1 and 2 deflection limits. EMCB RAI 3.9.3-1 In SSER 4 (Reference 2), the NRC staff noted that a sampling program was initiated by TVA to determine whether the compressive stresses imposed on short column pipe supports exceeded the buckling criteria margin established by the NRC. The NRC staff accepted the sampling program and determined that TVA had adequately addressed the t\IRC design criteria for Class 2 and 3 pipe supports; this resolved Outstanding Issue 2. Please confirm the applicability
-of the sampling program discussed in Reference 3 as it relates to Class 2 and 3 pipe supports at WBN Unit 2. If this sampling program was not used in support of the WBN Unit 2 refurbishment effort, please discuss the current criteria used for demonstrating that these pipe supports maintain sufficient margin against critical buckling of short column pipe supports.
EMCB RAI 3.9.3-2 In SSER 6 (Reference 3), the NRC staff noted its concerns regarding the licensee's use of earthquake experience data to seismically qualify Category I(L) piping and identified this concern as Outstanding Issue 19(h). In SSER 8 (Reference 5), the NRC staff noted that the licensee had developed screening criteria to identify items in Category I(L) piping systems that may require further evaluation based on this earthquake experience data. Additionally, the licensee indicated that bounding stress cases would be performed to demonstrate the conservatism of these screening criteria.
The NRC staff found this screening criteria adequate for demonstrating the seismic ruggedness of Category I(L) piping. Please confirm that this screening has been performed for the WBN Unit 2 refurbishment efforts. If this screening method was not utilized in the seismic qualification of the WBN Unit 2 Category I(L) piping, please discuss the criteria that has been used to seismically qualify these piping systems and discuss the regulatory acceptance bases for this alternate criteria.
EMCB RAI 3.9.3-3 In addition to the screening methods used for Category I(L) piping systems described in RAI 3.9.3-2, SSER 8 also describes TVA's criteria used for the evaluation of Category I(L) piping supports.
The NRC staff noted in SSER 8 that TVA had indicated it would utilize a factor of safety of three in their evaluation of concrete expansion anchor bolts for these pipe supports.
The NRC staff accepted the use of this safety factor value for validating the existing design of concrete expansion anchors used in this piping system based on TVA's implementation of recommendations including additional concrete inspection, anchor spacing, and concrete edge distance in conjunction with the eXisting anchor bolts. The NRC staff also noted in SSER 8 that for future Category I(L) piping, the required safety factors for these piping systems found in the former Office of Inspection and Enforcement (IE) Bulletin 79-02, should be utilized.
Please discuss whether the existing, applicable Category I(L) piping supports at WBN Unit 2 have been evaluated in the manner described in SSER 8. If these supports have been evaluated in a dissimilar manner, please provide justification for the departure from the methods described in Reference
- 4. EMCB RAI 5.5.1-1 Please discuss whether TVA has committed to perform an augmented inservice inspection of the reactor coolant pump (RCP) flywheel.
If no commitment has been made, please provide justification that the potential for excessive vibration on the reactor coolant pump flywheels will be adequately addressed to minimize the possibility of RCP shaft or flywheel failure.
-5 References
- 1) Letter from M. D. Jesse, Exelon Generation Company, LLC, to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN) -Unit 2 -Final Safety Analysis Report (FSAR), Amendment 97," dated January 11, 2010. (Accession Nos. ML 100191421 (letter), ML 100191684 (Section 3.8.5-3.11))
- 2) NUREG-0847, Supplement 4, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," dated March 31, 1985. (Accession No. ML072060524)
- 3) NUREG-0847, Supplement 6, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," dated April 30, 1991. (Accession No. ML072060464)
- 4) NUREG-0847, Supplement 7, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," dated September 30, 1991. (Accession No. ML072060471)
- 5) NUREG-0847, Supplement 8, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," dated January 31, 1992. (Accession No. ML072060478)
- 6) NUREG-0847, Supplement 13, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," dated April 30, 1994. (Accession No. L072060484)
Mr. Ashok S. Bhatnagar Senior Vice President Nuclear Generation Development and Construction Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SUB..WATTS BAR NUCLEAR PLANT, UNIT 2 -REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSEE'S FINAL SAFETY ANALYSIS REPORT AMENDMENT RELATED TO MECHANICAL AND CIVIL ENGINEERING SYSTEMS (TAC NO. ME2731)
Dear Mr. Bhatnagar:
By letter dated January 11, 2010 (NRC Agencywide Document Access and Management System Accession No. ML 100191686), to the U.S. Nuclear Regulatory Commission (NRC), the Tennessee Valley Authority provided an update (Amendment No. 97) to the Final Safety Analysis Report (FSAR) for the Watts Bar Nuclear Plant (WBN), Unit 2. That update contained changes to a number of sections of the WBN Unit 2 FSAR, including Sections 3.9.1, 3.9.2, 3.9.3, and 5.5.1. The NRC staff has reviewed these four sections and has identified additional information that is needed to complete the technical review of the operating license application.
A response is required within 30 days of receipt of this letter. If you should have any questions, please contact me at 301-415-1457.
Sincerely, IRA! Joel S. Wiebe, Senior Project Manager Watts Bar Special ProjeCts Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.
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