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Category:Letter
MONTHYEARIR 05000390/20244032025-01-31031 January 2025 Watts Barr Nuclear Plant - Material Control and Accounting Program Inspection Report 05000390/2024403 and 05000391/2024403 (Public) CNL-25-009, And Watts Bar Nuclear Plant, Units 1 and 2 - Organization Topical Report, TVA-NPOD89-A, Revision 262025-01-29029 January 2025 And Watts Bar Nuclear Plant, Units 1 and 2 - Organization Topical Report, TVA-NPOD89-A, Revision 26 05000391/LER-2025-001, Loss of the Control Room Emergency Air Temperature Control System Due to Chiller Valve Malfunction2025-01-21021 January 2025 Loss of the Control Room Emergency Air Temperature Control System Due to Chiller Valve Malfunction IR 05000390/20254022025-01-17017 January 2025 Information Request for the Cybersecurity Baseline Inspection Notification to Perform Inspection 05000390/2025402 05000391/2025402 CNL-25-012, Supplement to 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 .2025-01-16016 January 2025 Supplement to 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 . IR 05000391/20240402025-01-16016 January 2025 95001 Supplemental Inspection Report 05000391/2024040 and Follow-Up Assessment Letter CNL-24-076, Response to Request for Additional Information Regarding Application to Modify the Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13)2025-01-16016 January 2025 Response to Request for Additional Information Regarding Application to Modify the Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13) ML25006A1172025-01-10010 January 2025 – Review of the Fall 2023 Steam Generator Tube Inspection Report (EPID-L-2024-LRO-0022) CNL-25-001, License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2024-12-27027 December 2024 License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML24312A0052024-12-23023 December 2024 Issuance of Amendment No. 171 Regarding Extension of Facility Operating License Expiration Date to Recapture Low-Power Operating License Testing Time ML24312A3222024-12-23023 December 2024 Issuance of Amendment Nos. 334, 357, & 317; 368 & 362; 172 & 77 Regarding Revision to TS 5.4 & 5.7.1 CNL-24-009, Brown Ferry Nuclear Plant, Units 1, 2 & 3; Sequoyah Nuclear Plant, Units 1 & 2 and Watts Bar Plant, Units 1 & 2 - Triennial Decommission Funding Plans for Independent Spent Fuel Storage Installations2024-12-17017 December 2024 Brown Ferry Nuclear Plant, Units 1, 2 & 3; Sequoyah Nuclear Plant, Units 1 & 2 and Watts Bar Plant, Units 1 & 2 - Triennial Decommission Funding Plans for Independent Spent Fuel Storage Installations CNL-24-082, Central Emergency Control Center Emergency Plan Implementing Procedure Revision2024-12-17017 December 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revision ML24285A2072024-12-17017 December 2024 Amendment Nos. 170 and 76 Regarding the Revision of Technical Specifications 2.0, 3.0, 3.1, 3.2, 3.3, 3.4, and 5.9.5 by Adopting Various Technical Specifications Task Force Travelers IR 05000390/20243012024-12-16016 December 2024 NRC Operator License Examination Report 05000390/2024301 and 05000391/2024301 ML24346A3982024-12-16016 December 2024 NRC Examination Results Summary Examination Report 05000390/2024301 and 05000391/2024301 ML24352A0092024-12-16016 December 2024 Investigation Summary, Office of Investigations Case Number 2-2023-007 CNL-24-072, Request for Alternative to Extend First Containment Inservice Inspection Interval2024-12-12012 December 2024 Request for Alternative to Extend First Containment Inservice Inspection Interval CNL-24-075, Response to Request for Additional Information for Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah and Watts Bar (SQN-TSTS-23-02 and W2024-11-27027 November 2024 Response to Request for Additional Information for Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah and Watts Bar (SQN-TSTS-23-02 and WBN ML24297A4632024-11-21021 November 2024 – Environmental Assessment and Finding of No Significant Impact Related to Recapture of Low-Power Testing Time CNL-24-080, Response to Request for Additional Information Regarding Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis (WBN-19-011)2024-11-20020 November 2024 Response to Request for Additional Information Regarding Application to Revise Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis (WBN-19-011) IR 05000390/20240032024-11-13013 November 2024 Integrated Inspection Report 05000390/2024003, 05000391/2024003 & 07201048/2024001 CNL-24-021, Application to Revise Technical Specification Limiting Condition of Operation 3.5.2, ECCS – Operating, Note 1 to Include Residual Heat Removal Pump Flow Paths (SQN-TS-23-04 and WBN-TS-23-020)2024-11-12012 November 2024 Application to Revise Technical Specification Limiting Condition of Operation 3.5.2, ECCS – Operating, Note 1 to Include Residual Heat Removal Pump Flow Paths (SQN-TS-23-04 and WBN-TS-23-020) CNL-24-014, License Amendment Request to Revise the Sequoyah Nuclear Plant, Units 1 and 2 and Watts Bar Nuclear Plant, Units 1 and 2 Technical Specifications to Use Online Monitoring Methodology (SQN-TS-24-02 and WBN-TS-23-22)2024-11-0404 November 2024 License Amendment Request to Revise the Sequoyah Nuclear Plant, Units 1 and 2 and Watts Bar Nuclear Plant, Units 1 and 2 Technical Specifications to Use Online Monitoring Methodology (SQN-TS-24-02 and WBN-TS-23-22) IR 05000390/20250102024-11-0404 November 2024 Notification of an NRC (FPTI) (NRC Inspection Report 05000390/2025010 0500039/ 2025010) (RFI) CNL-24-064, Response to Request for Additional Information Regarding the Watts Bar Nuclear Plant, Unit 2 Steam Generator Tube Inspection Report for U2R52024-11-0404 November 2024 Response to Request for Additional Information Regarding the Watts Bar Nuclear Plant, Unit 2 Steam Generator Tube Inspection Report for U2R5 CNL-24-074, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-10-23023 October 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24290A1202024-10-17017 October 2024 Operator Licensing Examination Approval 05000390/2024301 and 05000391/2024301 ML24282B0412024-10-15015 October 2024 Request for Withholding Information from Public Disclosure for Sequoyah Nuclear Plant, Units 1 and 2 and Watts Bar Nuclear Plant, Units 1 and 2 ML24261C0062024-10-0404 October 2024 Correction to Amendment No. 134 to Facility Operating License No. NPF-90 and Amendment No. 38 to Facility Operating License No. NPF-96 ML24260A1682024-10-0404 October 2024 Regulatory Audit Summary Related to Request to Add and Revise Notes Related to Technical Specification Table 3.3.2-1, Function 5 ML24284A1072024-09-26026 September 2024 Affidavit for Request for Withholding Information from Public Disclosure for Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2 CNL-24-047, Decommitment of Flood Mode Mitigation Improvement Systems2024-09-24024 September 2024 Decommitment of Flood Mode Mitigation Improvement Systems CNL-24-060, Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description2024-09-24024 September 2024 Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description ML24262A0602024-09-23023 September 2024 Summary of August 19, 2024, Meeting with Tennessee Valley Authority Regarding a Proposed Supplement to the Tennessee Valley Authority Nuclear Quality Assurance Plan CNL-24-065, Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-09-18018 September 2024 Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions 05000390/LER-2024-002, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation2024-09-0505 September 2024 Automatic Reactor Trip Due to Main Generator Protection Relay Actuation IR 05000390/20240052024-08-28028 August 2024 Updated Inspection Plan and Assessment Follow-Up Letter for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390-2024005 and 05000391-2024005 ML24218A1442024-08-27027 August 2024 Issuance of Amendment Nos. 169 and 75 Regarding Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate IR 05000390/20244022024-08-20020 August 2024 – Security Baseline Inspection Report 05000390-2024402 and 05000391/2024402 - Public CNL-24-061, Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08),2024-08-19019 August 2024 Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08), ML24219A0262024-08-12012 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 05000391/LER-2024-003, Inoperability of Both Trains of Unit 2 Low Head Safety Injection2024-07-11011 July 2024 Inoperability of Both Trains of Unit 2 Low Head Safety Injection ML24185A0292024-07-0303 July 2024 Requalification Program Inspection Notification Letter 2025-01-31
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000391/LER-2025-001, Loss of the Control Room Emergency Air Temperature Control System Due to Chiller Valve Malfunction2025-01-21021 January 2025 Loss of the Control Room Emergency Air Temperature Control System Due to Chiller Valve Malfunction 05000390/LER-2024-002, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation2024-09-0505 September 2024 Automatic Reactor Trip Due to Main Generator Protection Relay Actuation 05000391/LER-2024-003, Inoperability of Both Trains of Unit 2 Low Head Safety Injection2024-07-11011 July 2024 Inoperability of Both Trains of Unit 2 Low Head Safety Injection 05000391/LER-2024-002, Re Automatic Reactor Trip Due to Steam Generator 3 Level LO-LO2024-05-0606 May 2024 Re Automatic Reactor Trip Due to Steam Generator 3 Level LO-LO 05000391/LER-2024-001, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation2024-03-27027 March 2024 Automatic Reactor Trip Due to Main Generator Protection Relay Actuation 05000391/LER-2023-003-01, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control2024-02-29029 February 2024 Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control 05000391/LER-2023-003, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control2023-10-0303 October 2023 Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control 05000390/LER-2023-001-01, Inadequate 10 CFR 50.59 Results in Failure to Obtain Prior NRC Approval for Condition Prohibited by Technical Specifications2023-09-27027 September 2023 Inadequate 10 CFR 50.59 Results in Failure to Obtain Prior NRC Approval for Condition Prohibited by Technical Specifications 05000391/LER-2023-002, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation2023-08-24024 August 2023 Automatic Reactor Trip Due to Main Generator Protection Relay Actuation 05000391/LER-2023-001, Unanalyzed Condition Related to Loss of the 2A Emergency Diesel Generator During a Postulated Appendix R Fire2023-07-20020 July 2023 Unanalyzed Condition Related to Loss of the 2A Emergency Diesel Generator During a Postulated Appendix R Fire 05000390/LER-2023-001, Interpretation of Technical Specification (TS) Table 1.1-1 Leads to a Condition Prohibited by TS2023-07-0303 July 2023 Interpretation of Technical Specification (TS) Table 1.1-1 Leads to a Condition Prohibited by TS 05000391/LER-2021-001, Automatic Reactor Trip on Main Turbine Trip Caused by Main Feed Pump Trip Due to Low Condenser Vacuum2021-05-10010 May 2021 Automatic Reactor Trip on Main Turbine Trip Caused by Main Feed Pump Trip Due to Low Condenser Vacuum 05000390/LER-2021-001, Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open2021-04-20020 April 2021 Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000391/LER-2020-004, Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking2021-01-0707 January 2021 Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking 05000390/LER-2020-005, Automatic Start of the Emergency Diesel Generators Due to an Equipment Failure During Transfer of Power Source for the 2A-A Shutdown Board2021-01-0404 January 2021 Automatic Start of the Emergency Diesel Generators Due to an Equipment Failure During Transfer of Power Source for the 2A-A Shutdown Board 05000391/LER-2020-003, Re Low RHR Flow in Mode 6 Results in a Condition Prohibited by Technical Specifications2020-12-21021 December 2020 Re Low RHR Flow in Mode 6 Results in a Condition Prohibited by Technical Specifications 05000391/LER-2020-002, Re Two Pressurizer Safety Valves Outside of Technical Specification Limits Due to Set Point Drift2020-12-17017 December 2020 Re Two Pressurizer Safety Valves Outside of Technical Specification Limits Due to Set Point Drift 05000390/LER-2020-003, Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open2020-09-10010 September 2020 Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000391/LER-2020-001, Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open2020-07-15015 July 2020 Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000390/LER-2020-002, Automatic Start of the Emergency Diesel Generators Due to an Equipment Failure During Transfer of Power Source for the 1B-B Shutdown Board2020-07-14014 July 2020 Automatic Start of the Emergency Diesel Generators Due to an Equipment Failure During Transfer of Power Source for the 1B-B Shutdown Board 05000390/LER-2020-001, Manual Reactor Trip Due to Lowering Steam Generator Level Caused by a Hand Station Failure2020-04-17017 April 2020 Manual Reactor Trip Due to Lowering Steam Generator Level Caused by a Hand Station Failure 05000390/LER-2019-004, Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open2020-01-13013 January 2020 Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000390/LER-2019-003, Manual Reactor Trip Due to Main Feedwater Regulating Valve Failing Closed2019-10-21021 October 2019 Manual Reactor Trip Due to Main Feedwater Regulating Valve Failing Closed 05000391/LER-2019-002, Breach Due to Penetration Boot Seal Separation Results in Shield Building Inoperability2019-08-19019 August 2019 Breach Due to Penetration Boot Seal Separation Results in Shield Building Inoperability 05000390/LER-2019-002, Loss of Control Room Emergency Air Temperature Control System Due to Air Filter Failure2019-08-0707 August 2019 Loss of Control Room Emergency Air Temperature Control System Due to Air Filter Failure 05000391/LER-2019-001, Regarding Manual Reactor Trip Due to Main Feedwater Regulating Valve Failing Closed2019-07-18018 July 2019 Regarding Manual Reactor Trip Due to Main Feedwater Regulating Valve Failing Closed 05000390/LER-2018-006, Containment Air Return Fan Inoperable for a Time Period Longer than Allowed by Technical Specifications Due to an Inadequate Post Maintenance Test2019-02-11011 February 2019 Containment Air Return Fan Inoperable for a Time Period Longer than Allowed by Technical Specifications Due to an Inadequate Post Maintenance Test 05000390/LER-2018-005, Regarding Manual Reactor Trip Due to Failure of Reactor Coolant Pump to Transfer to Normal Power2018-12-19019 December 2018 Regarding Manual Reactor Trip Due to Failure of Reactor Coolant Pump to Transfer to Normal Power 05000390/LER-2018-004, Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open2018-11-13013 November 2018 Control Room Emergency Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000391/LER-2018-005, For Watts Bar Nuclear Plant, Unit 2, Automatic Reactor Trip Due to Turbine Control System Card Failure and Throttle Valve Closure2018-10-22022 October 2018 For Watts Bar Nuclear Plant, Unit 2, Automatic Reactor Trip Due to Turbine Control System Card Failure and Throttle Valve Closure 05000391/LER-2018-004, Failure to Implement Annunciator Response Process Results in a Condition Prohibited by Technical Specifications2018-09-21021 September 2018 Failure to Implement Annunciator Response Process Results in a Condition Prohibited by Technical Specifications 05000391/LER-2018-003, Reactor Trip Due to Main Generator Differential Relay Actuation2018-08-21021 August 2018 Reactor Trip Due to Main Generator Differential Relay Actuation 05000391/LER-2018-002, Loss of Shield Building Vacuum Due to Equipment Failure2018-07-0909 July 2018 Loss of Shield Building Vacuum Due to Equipment Failure 05000390/LER-2018-002-01, Shield Building Inoperability Due to Annulus Vacuum Transient2018-06-26026 June 2018 Shield Building Inoperability Due to Annulus Vacuum Transient ML18144A9982018-05-24024 May 2018 Ventilation System Inoperable Due to Main Control Room Door Being Left Open 05000390/LER-2018-002, Regarding Shield Building Inoperability Due to Annulus Vacuum Transient2018-03-19019 March 2018 Regarding Shield Building Inoperability Due to Annulus Vacuum Transient 05000390/LER-2017-016, Regarding System Actuations Due to Opening of Feeder Breaker to the 1B-B 6.9 Kv Shutdown Board2018-02-20020 February 2018 Regarding System Actuations Due to Opening of Feeder Breaker to the 1B-B 6.9 Kv Shutdown Board 05000391/LER-2017-006, Regarding Manual Reactor Trip in Response to Indication of Multiple Dropped Control Rods2018-02-0909 February 2018 Regarding Manual Reactor Trip in Response to Indication of Multiple Dropped Control Rods 05000391/LER-1917-005, Regarding Unplanned Emergency Core Cooling System Injection Into the Reactor Coolant System Due to Personnel Error2018-01-25025 January 2018 Regarding Unplanned Emergency Core Cooling System Injection Into the Reactor Coolant System Due to Personnel Error 05000390/LER-1917-015, Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications2018-01-0808 January 2018 Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications 05000390/LER-1917-014, Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function2017-12-20020 December 2017 Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function 05000390/LER-1917-013, Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications2017-11-0606 November 2017 Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications 05000390/LER-1917-011, Regarding Failure to Enter Technical Specification 3.6.3 for Containment Lsolation Valve2017-10-23023 October 2017 Regarding Failure to Enter Technical Specification 3.6.3 for Containment Lsolation Valve 05000390/LER-1917-012, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications2017-10-23023 October 2017 Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications 05000390/LER-1917-010, Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board2017-10-10010 October 2017 Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board 05000391/LER-1917-004, Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication2017-09-25025 September 2017 Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication 05000390/LER-1917-008, Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation2017-08-14014 August 2017 Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation 05000390/LER-1917-007, Regarding Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance2017-08-0808 August 2017 Regarding Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance 05000390/LER-1917-006, Regarding Structural Degradation of 161 Kv Line Pole Leads to a Condition Prohibited by Technical Specifications2017-07-31031 July 2017 Regarding Structural Degradation of 161 Kv Line Pole Leads to a Condition Prohibited by Technical Specifications 05000390/LER-1917-005, Re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications2017-07-10010 July 2017 Re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications 2025-01-21
[Table view] |
LER-1917-013, Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications |
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text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 November 6, 2A17 10 cFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391 Subject: Licensee Event Report 39012017-013-00, lncorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications This submittal provides Licensee Event Report (LER) 39012017-013-00. This LER provides details concerning a condition where a damper which would not fully close resulted in a condition prohibited by the Technical Specifications (TS). This condition is being reported as a condition prohibited by TS in accordance with 10 CFR 50.73(aX2Xi)(B).
There are no regulatory commitments contained in this letter. Please direct any questions concerning this matter to Kim Hulvey, wBN Licensing Manager, al (423) 3os-zrzo.
Respectfully,%
Paul Simmons Site Vice President Watts Bar Nuclear Plant Enclosure cc: See Page 2
U.S. Nuclear Regulatory Commission Page 2 November 6, 2017 cc (Enclosure):
NRC Regional Administrator - Region ll NRC Senior Resident Inspector - Watts Bar Nuclear Plant
NRC FORM 366 (04-2017\\
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o-IW, U.S. NUCLEAR REGULATORY COMMISSION LTCENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO.3150-0104 ExptRES: oil31t2o2o
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME Watts Bar Nuclear Plant, Unit 1
- 2. DOCKET NUMBER 05000390
- 3. PAGE OF 5
1
- 4. TITLE
_ lncorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTHI ORy I Venn YEAR I t.-trt#s,-
REV NO.
MONTH I DAY YEAR FACILIry NAME Watts Bar Nuclear Plant, Unit 2 IOSOOOSg1 09 06 I 2017 2017 013 - 00 06 11 2017 FACILITY NAME I
oocxgr NLIITrtaER
- 9. OPERATING MODE 1 1. THIS REPORT lS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR g: (Check alt that appty) 1 tr 20 z2o1 (b) tr 2o.2zo3(a)(3Xr) tr 50.73(aX2)(iiXA) n 50 73(a)(2XviiiXA) n 20 2zo1(d) tr 20 z2o3(a)(3)(ii) t] 50 73(aX2XiiXB) tr 50.73(aX2Xviii)(B) tr zo 2zo3(aX1) n 20.2203(a)(4) n 50 73(ax2xiii) n 50 73(aX2Xix)(A) n 20 z2o3(aX2)(i) tr 50.36(cX1)(i)(A) tr 50.73(a)(2)(iv)(A) tr 50 73(aX2Xx)
- 10. POWER LEVEL 100 tr zo 22o3(a)(2Xii) tr 50 36(cX1X.iXA) tr 50.73(aX2XvXA) tr rc T1(aX4) tr 20 2203(aX2)(iii) n so s6(cX2) tl 50 73(a)(2)(v)(B) tr rcT1(a)(s) tr 2o.z2o3(aX2Xiv) n 50.46(ax3xii) tr 50 73(a)(2)(v)(c) tr rcrr(a)(1) tr zo zzo3(a)(2Xv) tr 50 73(aX2XiXA) n so 73(aX2)(vXD) tr fi 77(aX2Xi) tr 2o.zzo3(aX2)(vi)
X 50.73(aX2Xi)(B) tr 50 73(aX2Xvii) tr fi 77(ax2xii) tr 50 73(a)(2)(iXC) tr OTHER Specify in Abstract betow or in directly to the shield building exhaust vent. Following an accident, potential radioactivity releases within the AB are processed by the ABGTS units prior to release to the environment.
ABGTS performs the primary safety-related function of maintaining the AB at a minimum negative pressure of -0.25 inches water gauge (wg) to assure that the guidelines in 10 CFR 100 and General Design Criteria (GDC) 19 are not exceeded. Sufficient air is drawn from the ABSCE to establish and maintain the desired building negative pressure. The negative pressure chosen for post-accident operation is sufficiently low to ensure that airborne contamination in the AB is not released to the environment yltltgut being processed by the ABGTS units. Although the maximum permissible ABSCE leakage rate is 9900 cubic feet per minute (cfm) at -0.25 inches wg with respect to the outside environment, ABGTS will maintain a minimum fan capacity of 9300 cfm at all times. This will allow for adequate airflow for a maximum allowable ABSCE infiltration rate of 7930 cfm, and margin for a postulated one inch service air line break (1370cfm) due to a Safe Shutdown Earthquake (SSE), at a negative differential pressure of
- - 0.25 inches wg.
The inability of ABGTS Train A to meet the required greater than or equal to 1370 cfin total vacuum relief line airflow required by the design does not necessarily render ABGTS Train A incapable of performing its specified safety function during a Loss of Coolant Accident (LOCA) (maintaining ABSCE at i negativJ pressure, -0.25 inches wg and -0.5 inches wg). The difference between the actual vacuum relief measured value (corrected for pressure) and 1370 cfrn (from a postulated one inch air line break) provides the margin foTABSCE breaching permits during plant operation. The ABSCE breach margin is calculated using ABGTS total flow, vacuum relief line flow, and AB pressures. Therefore, once either train of ABGT-S's breach margin is exceeded that unit of ABGTS becomes inoperable and unable to perform its specifi ed safety function.
Testing performed on September 6,2017 showed the failure of damper 2-FCO-30-108 to fully close impacted the ability to accommodate the postulated one inch service air line break when the calculated breaches through the ABSCE exceeded 63.9 square inches for Train A. This breach margin was exceeded between July 7, 2017 at2030 EDT and September 5,2017 at 1645 EDT. Following correction of damper 2-FCO-30-108, the ABSCE breach margin increased to over 170 square inches.
ASSESSMENT OF SAFETY CONSEQUENCES
As described in the previous section, the ABGTS Train A was not able to perform its safeg function for a design basis accident with a one inch service airline failure. The postulation of a seismic event and a Design Basis Event (DBE) are not assumed to occur concurrently, but are considered for the purposes of design and operability. The consequences of this event are judged to be low, and are not modeled in the plant Probabilistic Risk Assessment (PRA).
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event Train B of ABGTS remained operable during this time period.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residuai heat, control the release of radioactive material, or mitigate the consequences of an accident V.
Train B of ABGTS remained operable during this time period.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service Train A of ABGTS was operable based on ABSCE breaches in effect when the condition of damper 2-FCO-30-108 was identified. The condition identified on September 6, 2017 was corrected on September 19,2017.
VI. CORRECTIVE ACTIONS
This event was entered into the Tennessee Valley Authority (TVA) Corrective Action Program and is being tracked under Condition Report (CR) 1335791.
A. lmmediate Corrective Actions The issue improperly adjusted damper was corrected after identification.
B. Corrective Actions to Prevent Recurrence or to Reduce Probability of Similar Events Occurring in the Future Corrective actions will be provided in a supplement to this report.
VII. PREVIOUS SIMILAR EVENTS AT THE SAME SITE
LER 390/2009-001 reported that Surveillance lnstructions used to test the ABGTS were ndt adequate in that the closure of non-safety ventilation dampers may have masked the performance of safety related dampers. ln addition, as part of this report, two failures of a temporary boundary door used to facilitate Unit 2 construction were reported. Corrective actions included revising procedures to ensure that non-safety dampers could not mask the performance of safety related dampers during testing. The event described in LER 390-2017-013 is different in that it is related to performance of post maintenance inspections, not surveillance procedures.
VIII. ADDITIONAL INFORMATION
None.
IX. COMMITMENTS None.Page 5 of,,, 5,,,
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05000391/LER-1917-001, Regarding Containment Airlock Function Lost Due to Equalizing Valve Not Closing | Regarding Containment Airlock Function Lost Due to Equalizing Valve Not Closing | | 05000391/LER-1917-002, Regarding Manual Reactor Trip as a Result of a Secondary Plant Transient | Regarding Manual Reactor Trip as a Result of a Secondary Plant Transient | 10 CFR 50.73(a)(2) | 05000391/LER-1917-003, Regarding Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure | Regarding Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2) | 05000391/LER-1917-004, Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication | Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication | 10 CFR 50.73(a)(2) | 05000390/LER-1917-004, Regarding Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup | Regarding Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup | 10 CFR 50.73(a)(2) | 05000391/LER-1917-005, Regarding Unplanned Emergency Core Cooling System Injection Into the Reactor Coolant System Due to Personnel Error | Regarding Unplanned Emergency Core Cooling System Injection Into the Reactor Coolant System Due to Personnel Error | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-1917-005, Re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications | Re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2) | 05000390/LER-1917-006, Regarding Structural Degradation of 161 Kv Line Pole Leads to a Condition Prohibited by Technical Specifications | Regarding Structural Degradation of 161 Kv Line Pole Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2) | 05000390/LER-1917-007, Regarding Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance | Regarding Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(viii)(B) | 05000390/LER-1917-008, Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation | Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000390/LER-1917-010, Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board | Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board | 10 CFR 50.73(a)(2) | 05000390/LER-1917-011, Regarding Failure to Enter Technical Specification 3.6.3 for Containment Lsolation Valve | Regarding Failure to Enter Technical Specification 3.6.3 for Containment Lsolation Valve | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) | 05000390/LER-1917-012, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications | Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications | 10 CFR 50.73(a)(2) | 05000390/LER-1917-013, Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications | Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-1917-014, Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function | Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function | | 05000390/LER-1917-015, Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications | Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) |
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