07-10-2017 | On May 10, 2017, at 0907 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant ( WBN) Unit 1 operations personnel discovered the 1B-B Safety Injection pump discharge isolation valve (1-ISV-63-527) closed. Technical Specification (TS) 3.5.2, ECCS - Operating, Condition A was immediately entered for one or more trains of the Emergency Core Cooling System ( ECCS) inoperable. TS 3.5.2 Condition A was exited at 0913 EDT when 1-ISV-63-527 was opened.
Investigation determined that the 1 B-B SI pump discharge isolation valve had been closed prior to Unit 1 entering Mode 3 on April 26, 2017, representing a condition prohibited by TS. During this time period, the 1A-A SI pump was inoperable for 21 minutes, representing a condition that could have prevented fulfillment of a safety function.
The cause of the mispositioned valve was the result of an individual failing to follow procedure use and adherence requirements during the performance of Emergency Diesel Generator ( EDG) Blackout testing. The safety injection pump discharge valve was closed to support the test but was not reopened following the testing. Corrective actions for this event include personal accountability actions, revision of the EDG blackout procedures to ensure the SI pump discharge valves are reopened, and additional station focus on procedure use, particularly use of Not Applicable (N/A) in performing procedures. |
---|
|
---|
Category:Letter
MONTHYEARCNL-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) 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 IR 05000390/20250102024-11-0404 November 2024 Notification of an NRC (FPTI) (NRC Inspection Report 05000390/2025010 0500039/ 2025010) (RFI) 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 IR 05000390/20243012024-10-17017 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 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 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 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-24-060, 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 CNL-24-047, Decommitment of Flood Mode Mitigation Improvement Systems2024-09-24024 September 2024 Decommitment of Flood Mode Mitigation Improvement Systems 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) 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 ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter 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 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 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 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-24-010, 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 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 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 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 2024-09-05
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000390/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 05000391/LER-2017-0052018-01-25025 January 2018 Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error, LER 17-005-00 for Watts Bar, Unit 2, Regarding Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error 05000390/LER-2017-0152018-01-0808 January 2018 Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications, LER 17-015-00 for Watts Bar, Units 1 and 2, Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications 05000390/LER-2017-0142017-12-20020 December 2017 Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function, LER 17-014-00 for Watts Bar, Unit 1, Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function 05000390/LER-2017-0122017-10-23023 October 2017 Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications, LER 17-012-00 for Watts Bar, Unit 1, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications 05000390/LER-2017-0112017-10-23023 October 2017 Failure to Enter Technical Specification 3.6.3 for Containment Isolation Valve, LER 17-011-00 for Watts Bar, Unit 1, Regarding Failure to Enter Technical Specification 3.6.3 for Containment lsolation Valve 05000390/LER-2017-0102017-10-10010 October 2017 Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board, LER 17-010-00 for Watts Bar Nuclear Plant, Unit 1 Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board 05000391/LER-2017-0042017-09-25025 September 2017 Manual Reactor Trip Due to Inoperable Rod Position Indication, LER 17-004-00 for Watts Bar Nuclear Plant, Unit 2 Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication 05000390/LER-2017-0042017-08-31031 August 2017 Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup, LER 17-004-01 for Watts Bar, Unit 1, Regarding Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup 05000390/LER-2017-0082017-08-14014 August 2017 Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation, LER 17-008-00 for Watts Bar, Unit 1, Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation 05000390/LER-2017-0072017-08-0808 August 2017 Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance, LER 17-007-00 for Watts Bar, Unit 1, Regarding Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance 05000390/LER-2017-0062017-07-31031 July 2017 Structural Degradation of 161 kV Line Pole Leads to a Condition Prohibited by Technical Specifications, LER 17-006-00 for Watts Bar, Unit 1, Regarding Structural Degradation of 161 kV Line Pole Leads to a Condition Prohibited by Technical Specifications 05000390/LER-2017-0052017-07-10010 July 2017 Isolation of the 1 B-B Safety Injection Pump Leads to Condition Prohibited by Technical Specifications, LER 17-005-00 for Watts Bar re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications 05000391/LER-2017-0032017-05-22022 May 2017 Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure, LER 17-003-00 for Watts Bar, Unit 2, Regarding Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure 05000391/LER-2017-0022017-05-12012 May 2017 Manual Reactor Trip as a Result of a Secondary Plant Transient, LER 17-002-00 for Watts Bar, Unit 2, Regarding Manual Reactor Trip as a Result of a Secondary Plant Transient 05000391/LER-2017-0012017-05-0303 May 2017 Containment Airlock Function Lost Due to Equalizing Valve Not Closing, LER 17-001-00 for Watts Bar, Unit 2, Regarding Containment Airlock Function Lost Due to Equalizing Valve Not Closing 05000390/LER-2017-0032017-03-0303 March 2017 Inadequate Operability Determination Leads to a Condition Prohibited by the Technical Specifications, LER 17-003-00 for Watts Bar, Unit 1, Regarding Inadequate Operability Determination Leads to a Condition Prohibited by the Technical Specifications 05000390/LER-2017-0022017-02-22022 February 2017 Incorrectly Hung Clearance Leads to a Condition Prohibited by the Technical Specifications, LER 17-002-00 for Watts Bar Nuclear Plant, Unit 1, Regarding: Incorrectly Hung Clearance Leads to a Condition Prohibited by the Technical Specifications 05000390/LER-2016-0112016-12-0909 December 2016 Loss of Centrifugal Charging Pump Due to Repeat Failure of Associated Room Cooler, LER 16-011-01 for Watts Bar Nuclear Plant, Unit 1 Regarding Loss of Centrifugal Charging Pump Due to Repeat Failure of Associated Room Cooler 05000391/LER-2016-0082016-10-28028 October 2016 Reactor Trip Resulting from Failure of 2B Main Bank Transformer, LER 16-008-00 for Watts Bar, Unit 2, Regarding Reactor Trip Resulting from Failure of 2B Main Bank Transformer 05000391/LER-2016-0052016-08-19019 August 2016 Main Feedwater Pump Trip on Loss of Condenser Vacuum Leads to Turbine Trip and Reactor Trip, LER 16-005-00 for Watts Bar, Unit 2, Regarding Main Feedwater Pump Trip on Loss of Condenser Vacuum Leads to Turbine Trip and Reactor Trip 05000390/LER-2016-0102016-08-0808 August 2016 Emergency Diesel Generator Crankcase Pressure Switches Not Analyzed to Withstand the Effects of a Tornado, LER-16-010-00 for Watts Bar Nuclear Plant, Units 1 and 2 Regarding Emergency Diesel Generator Crankcase Pressure Switches Not Analyzed to Withstand the Effects of a Tornado 05000391/LER-2016-0042016-08-0404 August 2016 Reactor Trip and Safety Injection Actuation Caused by Turbine Governor Valve Failure, LER 16-004-00 for Watts Bar, Unit 2, Regarding Reactor Trip and Safety Injection Actuation Caused by Turbine Governor Valve Failure 05000391/LER-2016-0032016-07-27027 July 2016 Turbine Driven Auxiliary Feedwater Pump Inoperable for Longer than Allowable Outage Time due to Governor Valve Spring Over-Tensioning, LER 16-003-00 for Watts Bar, Unit 2, Regarding Turbine Driven Auxiliary Feedwater Pump lnoperable for Longer Than Allowable Outage Time Due to Governor Valve Spring Over-Tensioning 05000390/LER-2016-0082016-07-15015 July 2016 Emergency Diesel Generator Manual Start Due to Loss of Voltage on the 6.9kV Shutdown Board 1B-B, LER 16-008-00 for Watts Bar, Unit 1, Regarding Emergency Diesel Generator Manual Start Due to Loss of Voltage on the 6.9kV Shutdown Board 1B-B 05000390/LER-2016-0092016-07-15015 July 2016 Failure to Complete Surveillance Requirements Causes a Condition Prohibited by the Technical Specifications, LER 16-009-00 for Watts Bar, Unit 1, Regarding Failure to Complete Surveillance Requirements Causes Conditions Prohibited by the Technical Specifications 05000391/LER-2016-0022016-07-11011 July 2016 Turbine Driven Auxiliary Feedwater Pump Inoperable for Longer than Allowable Outage Time due to Turbine Speed Control Failure, LER 16-002-00 for Watts Bar, Unit 2, Regarding Turbine Driven Auxiliary Feedwater Pump Inoperable for Longer Than Allowable Outage Time Due to Turbine Speed Control Failure 05000390/LER-2016-0062016-06-30030 June 2016 Undersized Room Cooler Fan Shaft Results in Loss of Centrifugal Charging Pump, LER 16-006-00 for Watts Bar, Unit 1, Regarding Undersized Room Cooler Fan Shaft Results in Loss of Centrifugal Charging Pump 05000390/LER-2016-0072016-06-20020 June 2016 Technical Specification Action Not Met for Rod Position Indication, LER 16-007-00 for Watts Bar Nuclear Plant, Unit 1 Regarding Technical Specification Action Not Met for Rod Position Indication 05000391/LER-2016-0012016-06-13013 June 2016 Loss of Automatic Containment Isolation for the Steam Generator Blowdown Sampling Lines, LER 16-001-00 for Watts Bar, Unit 2, Regarding Loss of Automatic Containment Isolation for the Steam Generator Blowdown Sampling Lines 2024-09-05
[Table view] |
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
I. PLANT OPERATING CONDITIONS BEFORE THE EVENT
Watts Bar Nuclear Plant (WBN) Unit 1 was at 46 percent rated thermal power (RTP) .
II. DESCRIPTION OF EVENT
A. Event Summary operations personnel discovered the 1B-B Safety Injection (SI) {EllS:BQ} pump discharge isolation valve {EIIS:V}(1-ISV-63-527) closed. Technical Specification (TS) 3.5.2, ECCS - Operating, Condition A was immediately entered for one or more trains of the Emergency Core Cooling System (ECCS) inoperable. TS 3.5.2 Condition A was exited at 0913 EDT when 1-ISV-63-527 was opened. Investigation determined that the 1B-B SI pump discharge isolation valve had been closed since prior to Unit 1 entering Mode 3 on April 26, 2017, representing a condition prohibited by TS. During this time period, the 1A-A SI pump was inoperable for 21 minutes, representing a loss of safety function.
This event is being reported to the Nuclear Regulatory Commission (NRC) under 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS and under 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of a safety function.
B. Inoperable Structures, Components, or Systems that Contributed to the Event No inoperable equipment contributed to this event.
C. Dates and Approximate Times of Occurrences Date Time Event (EDT) 4/11/2017 Preparations occur to perform 0-SI-82-4, 1B EDG Blackout Test, with Unit 1 in Mode 6. To prevent injection of water into the reactor coolant system, the SI pump discharge isolation valve was closed.
4/26/2017 1624 Unit 1 enters Mode 3 5/9/2017 1240 1A-A Emergency Diesel Generator (EDG) declared inoperable to check water in cylinders.
5/9/2017 1301 1A-A EDG declared operable.
5/10/2017 0907 1 B-B SI pump discharge isolation valve found closed, pump declared inoperable.
5/10/2017 0913 1B-B SI pump discharge isolation valve opened, pump declared operable.
D. Manufacturer and Model Number of Components that Failed During the Event Not applicable.
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
3. LER NUMBER
2017 - 00 005
E. Other Systems or Secondary Functions Affected
No other systems or secondary functions were affected .
F. Method of discovery of each Component or System Failure or Procedural Error This valve misposition was discovered by an operator performing routine operator rounds.
G. Failure Mode and Effect of Each Failed Component Not applicable.
H. Operator Actions
Upon discovering valve 1-ISV-63-527 isolated, the operator promptly opened the valve.
I. Automatically and Manually Initiated Safety System Responses Not applicable.
III. CAUSE OF THE EVENT
A. The cause of each component or system failure or personnel error, if known.
The test director for the EDG Blackout Test failed to follow procedure use and adherence requirements related to the application of Not Applicable (N/A), and did not obtain Section Manager concurrence for the use of N/A.
B. The cause(s) and circumstances for each human performance related root cause.
The test director for the EDG Blackout Test failed to follow procedure use and adherence requirements related to the application of Not Applicable (N/A), and did not obtain Section Manager concurrence for the use of N/A.
IV. ANALYSIS OF THE EVENT
On April 11, 2017, WBN Unit 1 was in Mode 6 during the Unit 1 fourteenth refueling outage (U1R14). The station was making preparations to perform 0-S1-82-4, 1B EDG Blackout Test. Blackout testing is normally conducted with the unit in Mode 5, however, due to a delay in the U1R14 schedule, the decision was made to conduct the testing in Mode 6. O-S1-82-4 Appendix B aligns the SI System for the blackout testing to ensure that water is not inadvertently injected into the core. Section 3.1 of Appendix B accomplishes this task by ensuring that the Cold Overpressure Mitigation System (COMS) clearance is in place. Normally, a COMS clearance places a hold order on the breakers and hand switches of the system pumps with the capability to inject high pressure water into the core. Appendix B, Section 3.1, Step [2] assumes this clearance for COMS is in place. Appendix B, Section 3.1, Step [3] takes the additional step of ensuring that the 1B-B SI Pump discharge valve (1-ISV-63-527-B) is closed and tagged. This is accomplished by used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
3. LER NUMBER
2017 - 00 005 verifying that the COMS clearance in Step [2] already holds 1-ISV-63-527-B closed and tagged or that the clearance is modified to include this valve.
0-S1-82-4 is written with the assumption that the unit will be in Mode 5 during the performance of the surveillance. An appendix to the procedure restores system alignment following conduct of the testing.
That appendix does not contain restoration steps for 1-ISV-63-527-B because the procedure assumes configuration control for this valve is maintained under the COMS clearance. The surveillance essentially transfers responsibility for configuration control to the COMS clearance, which was not required in this case due to the test being performed in Mode 6. This lack of configuration control in 0-S1-82-4 was a latent error introduced in the procedure in 2004.
The test director recognized that the unit was in Mode 6 and that the COMS clearance was neither required nor hanging at the time of the test. 0-S1-82-4 Appendix B, Section 3.1, Step [2] and [3] were marked N/A during the preparation. Step [2] is marked with a note that states "COMS not required in Mode 6". Step [3] is marked with a note that states "Valve verified closed but not tagged. Mode 6 does not require valve to be tagged for COMS". The test director failed to follow procedure use and adherence requirements. Specifically, the Section Manager concurrence was not obtained prior to moving to the next step in the procedure. Additionally, the test director did not consider the effect of N/A on these steps with regard to configuration control. The system restoration appendix was not reviewed to ensure adequate restoration steps were in place to restore 1-ISV-63-527-B to its required open position for normal operation.
Contributing to the event, while not required, performance of an 18 month locked valve verification and a system alignment verification were waived during the outage.
V. ASSESSMENT OF SAFETY CONSEQUENCES
Both trains of SI were required to be in service to comply with TS 3.5.2 following re-entry into Mode 3 on April 26, 2017. During the time until the valve misposition was identified on May 10, 2017, the 1A-A SI was operable except for a 21 minute period when its associated EDG was inoperable while it was checked to confirm water was not present in the engine cylinders. Therefore, during this 21 minute period, both SI trains were considered inoperable. An evaluation concluded the change in core damage probability from the 1 B-B SI pump being isolated for 14 days considering the brief period where the 1A-A SI pump was also unavailable, was less than 1E-7 during this time period, and the risk significance of this event was very small.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event The 1A-A SI pump was operable during the period in question except for a 21 minute 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 residual heat, control the release of radioactive material, or mitigate the consequences of an accident Not applicable.
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
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 The isolated 1B-B SI pump was returned to service seven minutes after discovery. The 1B-B SI pump was out of service from April 26, 2017 until May 10, 2017, or just over fourteen days.
VI. CORRECTIVE ACTIONS
This event was entered into the Tennessee Valley Authority (TVA) Corrective Action Program and is being tracked under Condition Report (CR) 1294133.
A. Immediate Corrective Actions
Upon discovering the 1B-B SI pump discharge isolation valve closed, the valve was immediately opened.
B. Corrective Actions to Prevent Recurrence or to Reduce Probability of Similar Events Occurring in the Future The EDG blackout procedures will be revised to ensure the SI pump discharge valves are reopened at the completion of testing. Additional management focus has been applied since this event related to procedure use and adherence, particularly in the application of N/A associated with procedure use.
VII. PREVIOUS SIMILAR EVENTS AT THE SAME SITE
clearance was placed on the wrong fuses for a containment purge valve. This led to the purge valve not having power removed to its actuator while leak testing was being performed. While this configuration control issue was also associated with human performance (failure to identify the proper fuse location), it was not associated with procedural compliance.
containment penetration was not isolated within four hours. The event described in this LER is different in that the correct actions to comply with the TS were understood, but a human performance error resulted in the correct actions not being performed.
Concerns with procedural use and adherence are a station focus area and are described in Section VI.B of this LER.
VIII. ADDITIONAL INFORMATION
None.
IX. COMMITMENTS
None.
|
---|
|
|
| | Reporting criterion |
---|
05000390/LER-2017-010 | Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board LER 17-010-00 for Watts Bar Nuclear Plant, Unit 1 Regarding Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9kV Shutdown Board | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000390/LER-2017-011 | Failure to Enter Technical Specification 3.6.3 for Containment Isolation Valve LER 17-011-00 for Watts Bar, Unit 1, Regarding Failure to Enter Technical Specification 3.6.3 for Containment lsolation Valve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-001 | Failure of Emergency Raw Cooling Water Pump Reverse Rotation Keys Represents Potential Common Cause Inoperability Watts Bar, Units 1 and 2 - Cancellation of Licensee Event Report 390/2017-001 Related to Essential Raw Cooling Water Pump Reverse Rotation Keys Representing a Potential Common Cause lnoperability | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000391/LER-2017-001 | Containment Airlock Function Lost Due to Equalizing Valve Not Closing LER 17-001-00 for Watts Bar, Unit 2, Regarding Containment Airlock Function Lost Due to Equalizing Valve Not Closing | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material | 05000391/LER-2017-002 | Manual Reactor Trip as a Result of a Secondary Plant Transient LER 17-002-00 for Watts Bar, Unit 2, Regarding Manual Reactor Trip as a Result of a Secondary Plant Transient | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-2017-012 | Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications LER 17-012-00 for Watts Bar, Unit 1, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-002 | Incorrectly Hung Clearance Leads to a Condition Prohibited by the Technical Specifications LER 17-002-00 for Watts Bar Nuclear Plant, Unit 1, Regarding: Incorrectly Hung Clearance Leads to a Condition Prohibited by the Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000391/LER-2017-003 | Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure LER 17-003-00 for Watts Bar, Unit 2, Regarding Automatic Start of Auxiliary Feedwater System Due to Main Condenser Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-2017-013 | 1 OF 5 LER 17-013-01 for Watts Bar Nuclear Plant, Units 1 and 2 Regarding Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-003 | Inadequate Operability Determination Leads to a Condition Prohibited by the Technical Specifications LER 17-003-00 for Watts Bar, Unit 1, Regarding Inadequate Operability Determination Leads to a Condition Prohibited by the Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-014 | Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function LER 17-014-00 for Watts Bar, Unit 1, Regarding Main Control Room Boundary Door Left Open Leading to a Loss of Safety Function | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000391/LER-2017-004 | Manual Reactor Trip Due to Inoperable Rod Position Indication LER 17-004-00 for Watts Bar Nuclear Plant, Unit 2 Regarding Manual Reactor Trip Due to Inoperable Rod Position Indication | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-2017-004 | Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup LER 17-004-01 for Watts Bar, Unit 1, Regarding Manual Reactor Trips Due to Failed Reactor Coolant Pump Power Transfer During Plant Startup | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000391/LER-2017-005 | Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error LER 17-005-00 for Watts Bar, Unit 2, Regarding Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-2017-005 | Isolation of the 1 B-B Safety Injection Pump Leads to Condition Prohibited by Technical Specifications LER 17-005-00 for Watts Bar re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-015 | Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications LER 17-015-00 for Watts Bar, Units 1 and 2, Regarding Failure to Enter Limiting Condition of Operation Action Statement Results in a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000390/LER-2017-006 | Structural Degradation of 161 kV Line Pole Leads to a Condition Prohibited by Technical Specifications LER 17-006-00 for Watts Bar, Unit 1, Regarding Structural Degradation of 161 kV Line Pole Leads to a Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000391/LER-2017-006 | Manual Reactor Trip in Response to Indication of Multiple Dropped Control Rods | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000390/LER-2017-007 | Multiple Unreported Potential Loss of Safety Function Events Associated with Inoperable Single Train Systems Due to Misinterpretation of Reporting Guidance LER 17-007-00 for Watts Bar, Unit 1, 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)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000390/LER-2017-008 | Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation LER 17-008-00 for Watts Bar, Unit 1, Regarding Shield Building Inoperability and Potential Loss of Safety Function Resulting from Spurious Equipment Operation | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000390/LER-2017-009 | Unanalyzed Condition Related to Dual Unit Operation of the Essential Raw Cooling Water System During a Design Basis Accident | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition |
|