03-09-2016 | On November 23, 2015, at 0844 Eastern Standard Time, Sequoyah Nuclear Plant ( SQN) Unit 1 reactor was manually tripped due to plant parameters indicating that the Loop 3 Main Steam Isolation Valve ( MSIV) had started drifting in the closed direction. Prior to the reactor trip, the open light indication on the main control board for the Loop 3 MSIV was noted to be extinguished. The light bulb was replaced with no change in indication. At the same time, the Post Accident Monitoring panel indicator for the Loop 3 MSIV displayed full open; however, within two to three minutes, the panel provided dual indication. Subsequently, Operators noted that the reactor coolant system temperature and Loop 3 Steam Generator ( SG) pressure were both rising, and the Loop 3 SG flow was lowering. These indications confirmed the Loop 3 MSIV was drifting closed. Following the reactor trip, all plant safety systems operated as designed, all control rods fully inserted, and auxiliary feedwater automatically initiated from the feedwater isolation signal, as expected. Troubleshooting identified a loose termination associated with the Loop 3 MSIV handswitch that would result in a slow loss of air pressure and cause the MSIV to slowly drift in the closed direction. The direct cause was determined to be a loose electrical connection on the MSIV handswitch. The root cause was determined to be inadequate work practices during replacement of the MSIV handswitch in 1994 that resulted in the loose electrical connection. The corrective action to prevent recurrence is revision of the work control planning procedure to ensure specific connection fastener torque values are utilized during work order planning. SQN Unit 2 was unaffected by this event. |
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Category:Letter
MONTHYEARML24032A0202024-01-31031 January 2024 NPDES Biocide/Corrosion Treatment Plan Annual Report, Cy 2023 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 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 ML24018A0142024-01-17017 January 2024 Engine Systems, Inc., Report No. 10CFR21-0137, Rev. 1, 56913-EN 56913 ML24011A3182024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), October 2023 ML24011A3172024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), September 2023 ML24011A3202024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), December 2023 ML24011A3162024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), August 2023 ML24011A3192024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), November 2023 IR 05000327/20234422024-01-11011 January 2024 95001 Supplemental Inspection Report 05000327/2023442 and 05000328/2023442 and Follow-Up Assessment Letter ML24010A2132024-01-10010 January 2024 CFR 21.21 Final Report Regarding Siemens Medium Voltage Circuit Breakers ML24018A0952024-01-0404 January 2024 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance Report No. 10CFR21-0137, Rev. 0 ML24004A0332024-01-0303 January 2024 Interim Report of a Deviation or Failure to Comply Crompton Instruments Type 077 Ammeter ML24004A0402024-01-0303 January 2024 Response to NRCs November 8, 2023, Request for Additional Information - Related to Independent Spent Fuel Storage Installation CNL-23-068, 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 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) ML23346A1222023-12-12012 December 2023 Annual Non-Radiological Environmental Operating Report - 2023 IR 05000327/20234202023-11-28028 November 2023 Security Baseline Inspection Report 05000327/2023420 and 05000328/2023420 CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23324A4362023-11-0909 November 2023 Exam Corporate Notification Letter Aka 210-day Letter ML23307A0822023-11-0808 November 2023 Request for Additional Information August 4, 2022, Exemption Request for Deviating from the Conditions of Certificate of Compliance No. 1032, Amendment No. 3, Related to Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation IR 05000327/20230032023-11-0303 November 2023 Integrated Inspection Report 05000327/2023003 and 05000328/2023003 ML23306A1592023-11-0202 November 2023 Enforcement Action EA-22-129 Inspection Readiness Notification ML23292A0792023-10-19019 October 2023 Tennessee Valley Authority - Emergency Plan Implementing Procedure Revision, Includes EPIP-5, Revision 58, General Emergency IR 05000327/20230112023-10-16016 October 2023 Triennial Fire Protection Inspection Report 05000327/2023011 and 05000328/2023011 ML23285A0882023-10-12012 October 2023 Submittal of Sequoyah Nuclear Plant, Units 1 and 2, Submittal of Updated Final Safety Analysis Report Amendment 31 ML23284A4252023-10-11011 October 2023 10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; Commitment Summary Report; and Update to the Fire Protection Report ML23283A2792023-10-10010 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Requirements Manual ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23277A0462023-10-0404 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification Bases ML23275A0272023-09-29029 September 2023 Submittal of Discharge Monitoring Report (DMR) Quality Assurance Study 43 Final Report 2023 ML23271A1662023-09-28028 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision IR 05000327/20234032023-09-14014 September 2023 Cyber Security Inspection Report 05000327/2023403 and 05000328/2023403 (Cover Letter) ML23257A0062023-09-14014 September 2023 Enforcement Action EA-22-129 Inspection Postponement Request ML23254A2192023-09-11011 September 2023 Emergency Plan Implementing Procedure Revisions ML23254A0652023-09-0707 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000327/20230052023-08-29029 August 2023 Updated Inspection Plan for Sequoyah Nuclear Plant, Units 1 and 2 - Report 05000327/2023005 and 05000328/2023005 ML23233A0122023-08-17017 August 2023 Unit 1 Cycle 25 Refueling Outage - 90-Day Inservice Inspection Summary Report - Supplement ML23233A0142023-08-15015 August 2023 Discharge Monitoring Report (Dmr), July 2023 ML23215A1212023-08-0303 August 2023 301 Exam Administrative Items (2B) Normal Release ML23215A1572023-08-0303 August 2023 Enforcement Action EA-22-129 Inspection Readiness Notification CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information 2024-01-04
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000327/LER-2021-001, & LER 2021-001-00 for Sequoyah Nuclear Plant, Units 1 & 2, Ice Bed Inoperable Due to Exceeding Surveillance Requirement Frequency2021-09-22022 September 2021 & LER 2021-001-00 for Sequoyah Nuclear Plant, Units 1 & 2, Ice Bed Inoperable Due to Exceeding Surveillance Requirement Frequency 05000327/LER-2017-0022017-07-14014 July 2017 Automatic Actuation of Emergency Diesel Generators Due to Loss of Power to 6.9kV Shutdown Board, LER 17-002-00 for Sequoyah, Unit 1, Regarding Automatic Actuation of Emergency Diesel Generators Due to Loss of Power to 6.9kV Shutdown Board 05000327/LER-2017-0012017-04-26026 April 2017 Breached Door Renders Both Trains of the Auxiliary Building Gas Treatment System Inoperable, LER 17-001-00 for Sequoyah, Unit 1, Regarding Breached Door Renders Both Trains of the Auxiliary Building Gas Treatment System Inoperable 05000327/LER-2016-0012016-04-11011 April 2016 Automatic Safety Injection due to Low Steam Line Pressure on Loop 2 Main Steam, LER 16-001-00 for Sequoyah, Unit 1, Regarding Automatic Safety Injection due to Low Steam Line Pressure on Loop 2 Main Steam 05000327/LER-2015-0042016-01-22022 January 2016 Manual Reactor Trip due to Main Steam Isolation Valve Drifting in the Closed Direction, LER 15-004-00 for Sequoyah, Unit 1, Regarding Manual Reactor Trip Due to Main Steam Isolation Valve Drifting in the Closed Direction 05000328/LER-2015-0022016-01-0606 January 2016 Unanalyzed Condition Due To Inoperable Containment Recirculation Drains, LER 15-002-00 for Sequoyah, Unit 2, Regarding Unanalyzed Condition Due to Inoperable Containment Recirculation Drains ML11276A1492011-09-30030 September 2011 Revised Submittal Schedule for Supplemental Report for License Event Report 327/2011-003, Unit I Reactor Trip as a Result of Turbine Control Card Failure ML11256A0162011-09-0606 September 2011 Withdrawal of License Event Report 327/2011-002, Feedwater Regulator Valve Inoperable ML0516403152005-06-0707 June 2005 LER 50-001-00 for Sequoyah, Unit 1 Re Automatic Reactor Trip Following Loss of Turbine Auto Stop Oil (ASO) Pressure ML0325804392003-09-0202 September 2003 LER 03-S01-00, Sequoyah Units 1 & 2, 30-Day Report Re Uncontrolled Security Weapon in the Protected Area 2021-09-22
[Table view] |
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I. Plant Operating Conditions Before the Event
At the time of the event, Sequoyah Nuclear Plant (SQN) Unit 1 reactor was operating at 100 percent rated thermal power (RTP). The condition described in this LER did not impact SQN Unit 2.
II. Description of Events
A. Event:
On November 23, 2015, at 0844 Eastern Standard Time (EST), SQN Unit 1 reactor was manually tripped due to plant parameters indicating that Loop 3 Main Steam Isolation Valve (MSIV) [EIIS Code SB] [EIIS Code ISV] had started drifting in the closed direction. Prior to the reactor trip, the open light indication [EIIS Code IL], on the main control room (MCR) panel for the MSIV was noted to be extinguished. The light bulb was replaced with no change in indication. At the same time, the Post Accident Monitoring (PAM) indicator for the MSIV displayed full open; however, within two to three minutes dual indication (mid-position) was provided. Subsequently, operators noted that the reactor coolant system (RCS) [EIIS Code AB] temperature and Loop 3 Steam Generator (SG) [EIIS Code SG] pressure were both slowly rising, and the Loop 3 SG flow was slowly lowering. These indications confirmed the Loop 3 MSIV was slowly drifting closed. Operators placed the handswitch [EIIS Code HS] for the MSIV in the open position for approximately 5 seconds. This resulted in no apparent affect. Operators manually tripped the reactor per procedure.
After the reactor trip, it was noted that all three lights on the MCR panel for the MSIV (closed, 10 percent closed, and open) illuminated followed by an immediate return to full open indication.
Additionally, PAM indication confirmed the MSIV was full open.
Troubleshooting identified a loose nut on a termination for the handswitch associated with the Loop 3 MSIV. The loose nut on the terminal could cause intermittent power through the circuit, which could cause flickering indicator lights and intermittent power to the solenoid. The loss of a single source of power to the solenoid could cut off the air supply to the MSIV, but not completely open the vent. This could result in a slow loss of air pressure and cause the Loop 3 MSIV to slowly drift in the closed position.
All plant safety related equipment operated as designed, all control rods fully inserted, and auxiliary feedwater (AFVV) [EIIS Code BA] automatically initiated from the feedwater isolation signal, as expected. No complications were experienced during the reactor trip.
This event is' reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event that resulted in a manual or automatic actuation of the Reactor Protection System and the Auxiliary Feedwater System.
B. Status of structures, components, or systems that were inoperable at the start of the event and contributed to the event:
There were no inoperable structures, components, or systems that contributed to this event.
C. Dates and approximate times of occurrences:
On November 23, 2015, at 0815 EST, operators noted the open light indicator on the MCR panel for the Loop 3 MSIV was extinguished while the PAM panel indicator for the MSIV indicated the valve was full open. Within minutes, the PAM panel indicated the MSIV was in mid-position. Operators noted the RCS temperature and Loop 3 SG pressure were both slowly rising, and the Loop 3 SG flow was slowly lowering. These indications confirmed the Loop 3 MSIV was slowly drifting closed. At 0844, the Unit 1 reactor was manually tripped.
November 23, 2015 at I All three light indicators for the MSIV on the MCR panel were 10845 EST illuminated followed by an immediate return to only full open indication. Coincidently, the indicator for the MSIV on the PAM panel indicated the valve was full open.
D. Manufacturer and model number of each component that failed during the event:
There were no components that failed during this event.
E. Other systems or secondary functions affected:
There were no other systems or functions affected by this event.
F. Method of discovery of each component or system failure or procedural error:
Operators observed open light indication for the Loop 3 MSIV on the MCR panel was extinguished while PAM indication initially showed full open. Approximately two to three minutes later, the PAM panel displayed dual indication. Subsequently, operators noted that the RCS temperature and Loop 3 SG pressure were both slowly rising, and the Loop 3 SG flow was slowly lowering. These indications confirmed the Loop 3 MSIV was slowly drifting closed.
G. The failure mode, mechanism, and effect of each failed component, if known:
There were no failed components associated with this event.
H. Operator actions:
After the Loop 3 MSIV was verified to be drifting closed by diverse indications, the operators established trigger values for Loop 3 SG pressure and RCS Tave-Tref mismatch. Once the Loop 3 MSIV showed dual indication on the PAM instrumentation, operators briefed for a potential manual reactor trip. After it was apparent that the Loop 3 MSIV was continuing to close, the operators made the decision to manually trip the reactor. Following the reactor trip, operators entered Emergency Procedure E-0, "Reactor Trip or Safety Injection," and then transitioned from E-0 to Emergency Subprocedure ES-0.1, "Reactor Trip Response." No human performance issues were identified.
I. Automatically and manually initiated safety system responses:
All plant safety related equipment operated as designed, all control rods fully inserted, and AFW automatically initiated from the feedwater isolation signal, as expected.
III. Cause of the event
A. The cause of each component or system failure or personnel error, if known:
The direct cause of the MSIV drifting in the closed direction was a loose connection (terminal lug and nut assembly) on the MSIV handswitch located in the MCR.
B. The cause(s) and circumstances for each human performance related root cause:
The root cause for this event was determined to be inadequate work practices during MSIV handswitch replacement in 1994. In 1994 during replacement of the handswitch, technicians utilized less than adequate work practices and human performance tools (i.e., fastener tightness, situational awareness, self-check, verification and procedure use) resulting in the assembly of the handswitch with a loose connection.
The root cause analysis is documented in Condition Report 1107656.
IV. Analysis of the event:
Prior to the event, SQN Unit 1 was operating at approximately 100 percent RTP with the RCS pressure and temperature near the nominal value of approximately 2235 pounds per square inch gauge (psig) and approximately 578 degrees Fahrenheit. Both the motor driven and the turbine driven AFW pumps and steam dump valves and the atmospheric relief valves were available.
The plant transient response including reactor power, RCS pressure, RCS temperature, pressurizer level, RCS secondary side pressure, and AFW flow remained within technical specification limits and were bounded by the Updated Final Safety Analysis Report (UFSAR) analysis. Containment pressure, temperature, and radiation levels were unaffected by this transient. SG level changes experienced during this event were bounded by UFSAR analysis. The plant responded as expected for the conditions of the trip.
V. Assessment of Safety Consequences
There were no safety consequences as a result of the event. All safety systems functioned as designed and no complications were experienced. Subsequent investigation determined that the Loop 3 MSIV remained capable of closing during the event and able to perform its safety function.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event:
There were no components that failed during this event. There were no other components that could have performed the same function as the Loop 3 MSIV.
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:
This event did not occur when the reactor was shut down. Safety-related systems that were needed to shut down the reactor, maintain safe shutdown conditions, remove residual heat or mitigate the consequences of an accident remained available throughout the event.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from discovery of the failure until the train was returned to service:
There was no failure that rendered a train of a safety system inoperable during this event.
VI. Corrective Actions
Corrective Actions are being managed by TVA's corrective action program under Condition Report 1107656.
A. Immediate Corrective Actions:
Troubleshooting of the Loop 3 MSIV handswitch was conducted. The cause of the intermittent electrical signal to the MSIV handswitch was identified and corrected.
B. Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future:
The corrective action to prevent recurrence is revision of the work control planning procedure to ensure specific connection fastener torque values are utilized during work order planning.
VII. Additional Information
A. Previous similar events at the same plant:
A review of previous reportable events for the past three years at SQN identified standards for multi-wire terminations and verifications associated with work performed in the mid-1990s.
B. Additional Information:
None.
C. Safety System Functional Failure Consideration:
This event did not result in a safety system functional failure.
D. Scrams with Complications Consideration:
This event did not result in an unplanned scram with complications.
VIII. Commitments:
None.