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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54127 | 20 June 2019 17:40:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Inoperability of Containment Shield Building | At 1340 EDT on June 20, 2019, a breach in excess of allowable margin in the Unit 2 Shield Building annulus was identified. T.S. LCO 3.6.15, Condition A was entered. The breach is expected to be repaired within the 24 hours allowed LCO time. No other equipment issues were identified. The Shield Building ensures that the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analysis during a Loss of Coolant Accident (LOCA). This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C). NRC Resident Inspector has been notified. The breach consists of a tear in a flexible boot seal for a penetration associated with the suction path for gas treatment fans. There is no release of radioactive material associated with this event. | Shield Building | |
| ENS 53398 | 11 May 2018 14:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Shield Building Inoperable | At 1011 EDT on May 11, 2018, Containment Shield Building Annulus differential pressure exceeded the required limit. The Shield Building was declared inoperable requiring entry into Technical Specification (TS) 3.6.15 Conditions A and B. The event was initiated by failure of the operating annulus vacuum fan. Main Control Room Operators manually started the stand-by annulus vacuum fan to recover pressure. Shield Building Annulus differential pressure was restored to the required value at 1016 EDT and TS 3.6.15 Condition A and B were exited on May 11, 2018 at 1016 EDT. The failure mechanism for the annulus vacuum fan is being investigated. The Containment Shield Building ensures the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analysis during a Loss of Coolant Accident (LOCA). The Emergency Gas Treatment System (EGTS) would have automatically started and performed its design function to maintain the Shield Building Annulus differential pressure within required limits. The event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident has been notified. | Shield Building Emergency Gas Treatment System | |
| ENS 53196 | 4 February 2018 09:45:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Failure of Containment Penetration Thermal Relief Check Valves to Meet Surveillance Acceptance Criteria | At 0445 (EST) on February 4, 2018, Watts Bar Unit 1 entered Technical Specification 3.6.1 condition A and 3.6.3 condition A.1 and A.2 due to inoperable containment penetration thermal relief check valves 1-CKV-31-3407 and 1-CKV-31-3421 associated with one train of the Containment Incore Instrument Room Chiller system. During surveillance testing, the thermal relief check valves failed to open and pass flow as required by acceptance criteria. The two penetrations were subsequently drained and isolated in accordance with the surveillance procedure to remove any thermal expansion concerns. Technical Specification 3.6.1 was exited February 4, 2018 at 0512 once the two penetrations were drained and isolated. The purpose of the thermal relief check valves is to allow flow from an isolated penetration back into the upstream containment piping to prevent over-pressurization due to thermal expansion. Over-pressurization of an isolated containment penetration could potentially cause the penetration or both of the isolation valves to fail and provide a direct flow path to the environment from the potentially contaminated containment atmosphere under certain Design Basis Accidents. Therefore, failure of the thermal relief check valves to open could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C). NRC Resident Inspector has been notified.
The purpose of this notification is to retract ENS notification 53196 made on 2/4/2018 for Watts Bar Nuclear Plant. The previous notification reported a surveillance failure of two containment penetration thermal relief check valves that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. After Engineering evaluation, it has been determined there is reasonable assurance the two thermal relief check valves (1-CKV-31-3407 and 1-CKV-31-3421) were capable of performing their specified safety function to isolate containment and act as a thermal relief device during a design basis accident. The basis of the evaluation included: 1. No maintenance activities or interactions with the check valves had occurred since last tested. 2. All surveillance testing for the valves was within required frequency. 3. The opening force for a new check valve of the same size and similar to 1-CKV-31-3407 and 1-CKV-31-3421 is 0.38 pounds. Engineering analysis has determined the minimum failure pressure of the piping systems associated with the containment penetration in question is 450 psig. If it is assumed the force applied on the check valve seat reaches 450 psig, the force applied on the seat would reach 111 pounds or 300 times the force required to open a new, clean check valve. Based on engineering judgement of previous operating experience where the pressure required to open the same stuck check valve was within a safety factor of 6 to potential equipment damage, the thermal relief check valves would have opened prior to equipment damage and thus the identified condition would not have resulted in adversely affecting the containment isolation boundary. Entry into Technical Specification (TS) 3.6.1 condition A on 2/4/2018 at 0445 has been retracted. Although not a loss of safety function, the containment penetrations associated with 1-CKV-31-3407 and 1-CKV-31-3421 remain inoperable and are being tracked by TS 3.6.3 condition A.1 and A.2. The NRC Resident Inspector has been notified. Notified the R2DO (Rose). | ||
| ENS 52912 | 17 August 2017 16:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9Kv Shutdown Board | On August 17, 2017, at 1205 Eastern Daylight Time (EDT), the Watts Bar Nuclear Plant (WBN) lost power to the 1B-B 6.9kV Shutdown Board. The loss of power to this safety related bus resulted in an automatic start of the Unit 1 Turbine Driven Auxiliary Feedwater Pump (TDAFWP). The motor driven AFW pumps did not start and were not expected to start for this event. No other system actuations occurred as a result of loss of power to the 1B-B 6.9kV Shutdown Board. Power to the 1B-B Shutdown Board was restored at 1505 EDT on August 17, 2017. During the loss of power to the 1B-B Shutdown Board, a reduction in containment and control rod drive mechanism cooling occurred. At 1233 EDT, lower containment average temperature exceeded the procedural limitations specified, and Technical Specification 3.6.5 Condition A was entered for containment average air temperature not within limits. Lower containment average temperature was restored to within limits at 1525 EDT on August 17, 2017. This is reportable as a potential loss of safety function. Other than several common Unit Technical Specifications having not been met, Unit 2 was not operationally impacted by the loss of power to the 1B-B Shutdown Board and remains in Mode 1 at 100% power. The NRC Resident Inspector has been notified. The loss of power occurred during restoration from Emergency Diesel Generator maintenance. | Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 52618 | 9 March 2017 06:20:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Airlock Function Lost Due to Equalizing Valve Not Closing | On March 9, 2017 at 0120 EST, the equalizing valve for the Watts Bar Nuclear Plant (WBN) Unit 2 upper containment airlock inboard door was found not closed while the outboard airlock door was open. This created a containment bypass with leakage potentially greater than allowed by the design. The operator immediately identified, after opening the outer door of the elevation 757 Air Lock, the pressure equalizing valve for the inner door was not fully closed. The outer door was promptly shut to isolate the airlock. The inner door was then cycled which closed the equalizing valve. The total time that a containment bypass was present is estimated at five minutes. This condition was initially recognized as being potentially reportable. However, the assessment performed at the time, based on operator control of the outer airlock door, concluded there was a reasonable assurance of meeting the safety function. A subsequent independent review of this condition identified that this should have been reported under 10 CFR 50.72(b)(3)(v)(C) within eight hours of the event. The NRC Resident Inspector has been informed of this event. The licensee is planning on replacing the equalizing valve in the near future. | 05000391/LER-2017-001 | |
| ENS 51795 | 17 March 2016 05:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Loss of Emergency and Auxiliary Gas Treatment Systems | On March 17, 2016, at 0115 (EDT), Watts Bar Unit 1 (WBN1) entered Technical Specification Limiting Condition of Operation (TS LCO) 3.0.3 due to the inoperability of both trains of the Emergency Gas Treatment System (EGTS). TS LCO 3.7.12 Condition B was also entered at this time due to the inoperability of both trains of the Auxiliary Gas Treatment System (ABGTS). The train B EGTS and train B ABGTS had been removed from service for scheduled maintenance, when at 0115, the train A Auxiliary Air Compressor became inoperable. On March 17, 2016, at 0133, the train A Auxiliary Air Compressor was declared OPERABLE, and TS LCO 3.0.3 and 3.7.12 Condition B were exited. The auxiliary air system supports the EGTS by providing a safety grade air supply. When train A auxiliary air became inoperable, the supported train A EGTS and ABGTS became inoperable, creating a condition where both trains of EGTS and ABGTS were unavailable. In the event of an accident, the EGTS establishes a negative pressure in the annulus between the shield building and the steel containment vessel and the ABGTS establishes a negative pressure in the Auxiliary Building Secondary Containment Enclosure (ABSCE). Filters in these system mitigate the release of radioactive contaminants to the environment. WBN1 remained in Mode 1 at 100% power and no safety functions were required during the event. This event is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract event report no. 51795 made on 3/17/16 at 0802 (EDT). Previously, Tennessee Valley Authority (TVA) reported a loss of the Emergency and Auxiliary Building Gas Treatment Systems (EGTS/ABGTS) at Watts Bar Nuclear Plant Unit 1 (WBN1). Both trains of EGTS and ABGTS were declared INOPERABLE when the train A auxiliary air system cooling water supply bypass valve was isolated, prior to completing the requisite post maintenance testing following repairs to the normal cooling water supply solenoid valve. Upon recognition, WBN1 operations personnel declared the train A auxiliary air system INOPERABLE, resulting in inoperability of Train A EGTS and ABGTS and forcing entry into TS LCO (Limiting Condition for Operation) 3.0.3 (from TS LCO 3.6.9 EGTS) and 3.7.12 Condition B for ABGTS. At the time the condition was recognized, train B EGTS and train B ABGTS were INOPERABLE for scheduled maintenance. Subsequently, TVA completed the post maintenance testing of the train A auxiliary air system ERCW (Emergency Raw Cooling Water) normal supply solenoid valve and determined that the valve, while not fully qualified at the time, was in fact operable and capable of performing its safety function. Therefore, entry into TS LCO 3.0.3 and 3.7.12 Condition B was not necessary and the event is no longer reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. Notified the R2DO (McCoy). | Secondary containment Auxiliary Building Gas Treatment System Shield Building Emergency Gas Treatment System | |
| ENS 49541 | 14 November 2013 21:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Postulated Fire Induced Failure of Centrifugal Charging Pumps | During analysis of Watts Bar Nuclear (WBN) Unit 2 fire protection features, it was revealed that a potential fire induced failure of centrifugal charging pumps could occur in Unit 1. Specifically, a potential fire induced failure of both Unit 1 Chemical and Volume Control System centrifugal charging pumps (CCPs) (1-PMP-62-108-A and 1-PMP-62-104-B) could occur due a fire in either auxiliary building room 737.0-A1 (general area for elevation 737.0) or 757.0-A2 (6.9 kV and Shutdown Board Room A). It is postulated that a fire in these rooms could cause a spurious closure of the CCP suction valve (1-LCV-62-133-B) from the volume control tank (VCT) (1-TANK-62-129) and could disable the control circuit which opens the flow from the refueling water storage tank (RWST) suction valve (1-LCV-62-135-A). The fire safe shutdown analysis (Fire Protection Report, Part VI) currently addresses this occurrence via the performance of a prompt main control room operator action to open the RWST suction path. However, this procedurally directed action may require several minutes to complete and due to the potentially short duration (possibly as short as a few seconds) for CCP survivability without suction flow, the action has now been determined to be unacceptable. As a result, the loss of charging flow could result in a loss of injection to the reactor coolant pump (RCP) seals which could subsequently lead to a RCP seal failure and a small break loss of coolant event. WBN engineering is continuing to validate whether the CCP minimum flow recirculation would protect the pumps with both suction paths (VCT and RWST) isolated and with the reactor at normal operating pressure. WBN has established compensatory measures to ensure that a fire in affected rooms will not cause a spurious closure of the CCP suctions valves. The licensee has notified the NRC Resident Inspector. | ||
| ENS 48995 | 3 May 2013 05:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Temporary Loss of Both Trains of Emergency Gas Treatment System | On May 3, 2013, at 0111 (EDT), Technical Specification Limiting Condition of Operation (TS LCO) 3.0.3 was entered due to the loss of both trains of Emergency Gas Treatment System (EGTS). The Train B EGTS had been removed from service for scheduled maintenance and at 0111 the Train A auxiliary air dryer stopped functioning. On May 3, 2013, at 0155, Train B EGTS was restored to service and declared Operable, and TS LCO 3.0.3 was exited. The auxiliary air system is required to support multiple safety related systems. The auxiliary air system is the safety grade air supply for EGTS. As Train A auxiliary air was no longer Operable, and the B train EGTS system was inoperable, the safety function supported by EGTS was not available. The EGTS establishes a negative pressure in the annulus between the shield building and the steel containment vessel. Filters in the system then control the release of radioactive contaminants to the environment. Watts Bar Unit 1 remained in Mode 1 at 100% power. No reactivity was added to the plant. This event is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. | Shield Building Emergency Gas Treatment System | |
| ENS 45170 | 27 June 2009 17:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Auxiliary Building Gas Treatment System Rendered Inoperable | On June 27, 2009, Plant Operations was starting up the U2 (Unit 2) Auxiliary Building General Ventilation System upon completion of U2 damper maintenance. During startup of the U2 Auxiliary Building Ventilation System, the operator started the U2 General Exhaust fan followed by the startup of the U2 General Supply fan. During the time span of less than 1 minute, when the U2 General Exhaust fan was running but before the U2 General Supply fan was running, the differential pressure condition across a set of Auxiliary Building Secondary Containment Enclosure (ABSCE) doors caused the doors to come off of the tracks and therefore created an ABSCE breach. A breach of this magnitude renders both trains of Auxiliary Building Gas Treatment System (ABGTS) inoperable. Upon identification of the failed doors at 1320 EDT on June 27, 2009 the plant entered TS Condition 3.7.12.B for 2 trains of ABGTS being inoperable. Repairs to re-establish the integrity of the ABSCE were immediately initiated and TS Condition 3.7.12.B was exited at 1649 EDT, approximately 3 and 1/2 hours later. This event is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the safety function of structures or systems that are needed to : (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector. | Secondary containment Auxiliary Building Gas Treatment System | |
| ENS 45110 | 27 May 2009 23:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Both Trains of Auxiliary Building Gas Treatment System Rendered Inoperable | On May 27, 2009, Plant Operations was shutting down the Auxiliary Building General Ventilation system (ABGTS) as a compensatory measure in response to an historical deficiency that was identified in the surveillance procedure for conducting TS SR 3.7.12.4. The building's general ventilation was being shut down to place the plant in a known tested configuration to ensure continued safe operation while plans were being developed to properly perform SR 3.7.12.4. While shutting down the building's ventilation system, the operator shut down the building's supply fans followed by the shut down of building's exhaust fans. This sequence resulted in an unacceptable differential pressure condition across a set of Auxiliary Building Secondary Containment Enclosure (ABSCE) doors. Upon identification of the failed doors at 1905 EDT on May 27, the plant entered TS Condition 3.7.12.B for 2 trains of ABGTS being inoperable. Repairs to re-establish the integrity of the ABSCE were immediately initiated and TS Condition 3.7.12.B was exited at 2239 EDT, approximately 3 and 1/2 hours later. The reportability of this condition was not recognized at the time of this event. Based upon a review subsequent to the event on June 1, 2009, this event was determined to be reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the safety function of structures or systems that are needed to: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector. | Secondary containment Auxiliary Building Gas Treatment System |