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 Entered dateSiteRegionReactor typeEvent description
ENS 549325 October 2020 14:25:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 6, 2020, at approximately 1749 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced a loss of Reactor Protection System (RPS) Bus 2A. Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolated in response to this event. The PCIS isolations caused the initiation of Standby Gas Treatment (SBGT) trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem A. Unit 2 declared RCS leakage detection instrumentation inoperable and entered TS LCO 3.4.5 condition A, B, and D with required action D.1 to enter LCO 3.0.3 immediately. Unit 2 entered TS LCO 3.0.3 with required actions to be in Mode 2 within 10 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours. Upon investigation, it was discovered that an age-related overheating condition resulted in the failure of the 2A RPS Motor Generator (MG) set, causing the feeder beaker from the 2A 480v Remote Motor-Operated Valve distribution board to trip. On August 6, 2020, at approximately 1808 CDT, Operations personnel commenced restoration of Unit 2 to normal after transferring 2A RPS to its alternate power supply. The 2A RPS MG Set drive motor was replaced on August 24, 2020. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. Plant conditions which initiate PCIS Group 3 actuations are RV Low Water Level or Reactor Water Cleanup Area High Temperature. Plant conditions which initiate PCIS Group 6 actuations are RV Low Water Level, High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation. Plant conditions which initiate PCIS Group 8 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. All affected safety systems responded as expected. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1628707. The NRC Resident Inspector has been notified of this event.
ENS 549315 October 2020 13:51:00Browns FerryNRC Region 2This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of an emergency service water system component that does not normally run and which provides an ultimate heat sink. On August 6, 2020, at approximately 0128 CDT, the A3 Emergency Equipment Cooling Water (EECW) pump received an auto-start signal while performing Post-Maintenance Testing (PMT) on the 3C Core Spray pump. Normally, the involved EECW pump would be started prior to testing to prevent an auto-start; however, in this case the pump was not running prior to the test. When the 3C Core Spray pump breaker was closed while in the test position, an unanticipated actuation of the A3 EECW pump occurred. Work was stopped and the workers reported to the Control Room to evaluate the condition. Based on a review of this event, individuals involved were coached on understanding system response prior to performing work. The A3 EECW pump responded in accordance with the plant design. No other plant equipment was affected during this event. There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution. Reference corrective action document CR 1628479. The NRC Resident Inspector has been notified of this event.
ENS 5491828 September 2020 14:39:00SequoyahNRC Region 2

On 9/28/20 at 1143 EDT, a notification to the National Response Center was made after discovery of a visible oil sheen on the waters of the U.S. (Sequoyah's side of the intake forebay skimmer wall at the Essential Raw Cooling Water (ERCW) building). The source was an oil bucket that overflowed with rain at the ERCW pumping station. Efforts are in progress to eliminate all other potential sources of oil at the station that could be released to the environment. Estimate of volume spilled is less than one quart. The following agencies have also been notified: - EPA Region 4 - Tennessee Emergency Management Agency (TEMA) - Tennessee Department of Environment and Conservation (TDEC). Cleanup is in progress. Measures to prevent recurrence are being taken.

This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified.

ENS 5487031 August 2020 15:30:00Watts BarNRC Region 2This 60-day telephone notification is being submitted in accordance with paragraphs 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) to report two invalid Containment Ventilation Isolation (CVI) actuations at Watts Bar Nuclear Plant (WBN) Unit 1. On July 23, 2020, at 0956 Eastern Daylight Time (EDT), the Train A CVI actuated due to an invalid high radiation signal from 1-RM-90-130, Containment Purge Air Exhaust Monitor. Upon investigation, the high radiation signal was caused by a failed power supply. Corrective action included replacing the power supply, 1-RM-90-130 detector, and restoring the system to service. On August 7, 2020, at 2017 EDT, the Train A CVI actuated due to an invalid high radiation signal from 1-RM-90-130, Containment Purge Air Exhaust Monitor. Upon investigation, a small tear was identified in the foil covering the scintillation detector. This defect caused erratic indication and the system actuation. The foil was replaced and the system was restored to service. Prior to and following the invalid high radiation alarms, all radiation monitors except 1-RM-90-130 were stable at their normal values; therefore, the CVI was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. These events were entered into the corrective action program as CR 1625135 and CR 1628904. The NRC Resident Inspector was notified.
ENS 5480024 July 2020 09:00:00SequoyahNRC Region 2At 0105 (EDT) on 7/24/20 it was discovered Unit 2 Ice Bed was INOPERABLE. Therefore, since this is a single train system the requirements of 50.72 (b)(3)(v)(C) and (D) have been met. This condition is being reported as an 8-hour non-emergency NRC Notification. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This condition put the unit in a 48-hour LCO. The old chillers were put into service to bring the temperature of the ICE bed down. At 0833 EDT, the technical specification limit was no longer exceeded and the unit exited the LCO.
ENS 5479521 July 2020 08:58:00Browns FerryNRC Region 2The following was received from TVA - Brown's Ferry at 0858, 21 July 20. On July 21, 2020, at 0435 hours Central Daylight Time, Browns Ferry Unit 1 inserted a manual reactor scram due to degrading main condenser vacuum from marine biofouling at the intake structure. Browns Ferry Unit 2 is in Mode 4 and Browns Ferry Unit 3 is at approximately 76% rated thermal power and stable. Primary Containment Isolation Systems received an actuation signal for groups 2, 3, 6, 8 on reactor water level below +2". All Primary Containment Isolation System groups that received an actuation signal performed as designed. Additionally, all other systems functioned as designed. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) - Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event also requires an 8 hour report per 10CFR50.72(b )(3)(iv)(A), Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation and (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The NRC Resident Inspector has been notified." The plant is stable in Mode 3 and will remain shutdown until marine growth clogging the intake structure abates.
ENS 5479420 July 2020 18:11:00Browns FerryNRC Region 2

On July 20, 2020, at 1325 hours Central Daylight Time, Brown's Ferry Unit 2 inserted a manual reactor scram due to degrading main condenser vacuum from marine biofouling at the intake structure. Brown's Ferry Unit 1 performed a down power to 43% and Unit 3 down powered to 76%. Conditions are stable on both Unit 1 and 3 following unit down power.

Primary Containment Isolations Systems received an actuation signal for groups 2, 3, 6, and 8 on reactor water level below +2". All Primary Containment Isolations System groups that received an actuation signal performed as designed. Additionally, all other systems functioned as designed. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) - Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

This event also requires an 8 hour report per 10CFR 50.72(b)(3)(iv)(A), "Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation and (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The NRC Resident Inspector has been notified. All control rods fully inserted and decay heat is being removed via normal feedwater and condenser.

ENS 5478115 July 2020 14:58:00Watts BarNRC Region 2At 0835 EDT on July 15, 2020, it was discovered that the main control room (MCR) envelope was inoperable due to a MCR door being found ajar; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The door was closed, restoring the MCR envelope to operable at 0839 EDT. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5475322 June 2020 13:45:00Watts BarNRC Region 2

At 1304 EDT on June 22, 2020, Watts Bar Nuclear Plant (WBN) Units 1 and 2 initiated voluntary communication to the State of Tennessee and local officials as part of the Nuclear Energy Institute (NEI) Groundwater Protection Initiative (GPI), after receiving analysis results for one on-site monitoring well that indicated tritium activity above the GPI voluntary communication threshold. WBN identified and corrected the cause. This condition did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi).

There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

ENS 5472020 May 2020 12:07:00Watts BarNRC Region 2At 0521 EDT on May 20, 2020, with Unit 2 in Mode 3 at 0 percent power and Unit 1 defueled, an actuation of the Emergency Diesel Generator (EDG) System occurred while transferring the 1B-B 6.9kV Shutdown Board (SDBD) from the maintenance feed to its normal power supply. The reason for the 1B-B 6.9kV SDBD failing to transfer to the normal power supply is under investigation. The EDGs automatically started as designed when a valid actuation signal was received. The event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the EDGs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5471819 May 2020 19:57:00Watts BarNRC Region 2At 1329 EDT on May 19, 2020, it was discovered that the main control room (MCR) envelope was inoperable due to a control room envelope (CRE) door being found ajar; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The CRE door was closed, restoring the MCR envelope to operable at 1331 EDT. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5470913 May 2020 13:10:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On March 16, 2020, at approximately 0102 CDT, Browns Ferry Nuclear Plant (BFN), Unit 3 received motor trip-out alarms and diagnosed Group 2 and 3 Primary Containment Isolation System (PCIS) Isolations, 3C Residual Heat Removal (RHR) Pump tripping and Reactor Water Cleanup (RWCU) system isolating. All affected safety systems responded as expected. BFN, Unit 3, was nearing the end of the U3R19 refueling outage at the time of the event, and was still dependent on the Shutdown Cooling (SDC) system. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. At the time of the event, these conditions did not exist: therefore, the PCIS actuation was invalid. The event was determined to have been caused by clearance restoration activities in an unprotected control panel. A fuse re-installation inadvertently created a fault condition between two different plant 120 VAC power sources when the fuse holder's lower spring clip contacted a different fuse. This was a result of age-related degradation of the fuse holder, its close proximity to other fuses, and the lack of insulating isolation barriers between fuses. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1594925. The NRC Resident Inspector has been notified of this event.
ENS 5470813 May 2020 05:38:00SequoyahNRC Region 2At 0208 EDT on 05/13/2020, Sequoyah Unit 1 was at 100% power when an automatic reactor trip signal was received concurrent with a low steam line pressure safety injection signal. The low steam line pressure safety injection signal was actuated from the steam pressure rate of decrease feature. Main steam isolation valves (MSIVs) automatically closed as designed and steam generator pressures stabilized following the isolation. All other safety-related equipment operated as designed, with the exception of 1-FCV-61-122 Glycol inboard containment isolation valve which failed to automatically isolate on a Phase A containment isolation signal. The corresponding outboard containment isolation valve, 1-FCV-61-110, automatically isolated as designed which isolated penetration X-114. Safety injection was terminated at 0221 EDT 5/13/20, and Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure with auxiliary feedwater supplying the steam generators and decay heat removal via steam generator atmospheric relief valves. There is no indication of any primary to secondary leakage. The electrical alignment is normal with shutdown power supplied from off-site power. There is no current operational impact to Unit 2. There is no impact on public health or safety. Post safety injection actuation investigation is in progress. The NRC Resident Inspector has been notified.
ENS 546976 May 2020 14:11:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On March 8, 2020, at approximately 2021 CDT, Browns Ferry Nuclear Plant Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and the initiation of Standby Gas Treatment Trains A and B, and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The RPS MG Set trip was believed to have been caused by an intermittent short across a spike suppressor, which led to a loss of generator output signal to a voltage regulator. The affected components have been replaced. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1593265. The NRC Resident Inspector has been notified of this event.
ENS 5453119 February 2020 10:20:00Watts BarNRC Region 2

EN Revision Text: NOTIFICATION OF UNUSUAL EVENT DUE TO FIRE IN CONTROL BUILDING At 0957 EST on February 19, 2020, a Notification of Unusual Event (NOUE) has been determined to be present at the Watts Bar plant Unit 1 under criteria HU4 for a fire potentially degrading the safety of the plant (fire for more than 15 minutes). The NRC Senior Resident Inspector has been notified for this event. Notified DHS SWO, FEMA Operations Center, CISA IOCC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 02/19/2020 AT 1151 EST FROM ANDREW WALDMANN TO DONALD NORWOOD * * *

The fire was declared extinguished at 1033 EST. The NOUE was terminated at 1126 EST. The investigation into the cause of the fire is in progress. Notified R2DO (Musser), NRR EO (Miller), and IRD MOC (Kennedy). Additionally, notified DHS SWO, FEMA Operations Center, CISA IOCC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * RETRACTION ON 2/20/2020 AT 1453 EST FROM MICHAEL BUTHEY TO RICHARD L. SMITH * * *

Watts Bar Nuclear Plant (WBN) is retracting Event Notice 54531 (NOUE notification) based on the following additional information. WBN reported a condition that was determined to meet the definition of a FIRE in the plant Emergency Preparedness Implementing Procedures (EPIP) based on indications available to the decision-maker at the time the declaration was made. A fire, without observation of flame, is considered present if large quantities of smoke and heat are observed. Moderate quantities of smoke were observed coming from an electrical cabinet not required to support safe plant operation. Once Fire Brigade personnel were able to access the affected room, no evidence of flame or significant heat was observed. Plant personnel ultimately determined that an overheated electrical component (transformer) resulted in the smoke. As such, the actual conditions did not meet the EPIP definition of a fire. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Musser), NRR EO (Miller), and IRD MOC (Kennedy).

ENS 5453219 February 2020 10:20:00Watts BarNRC Region 2At 0936 EST on February 19, 2020, the Watts Bar Unit 1 reactor was manually tripped while operating at 100 percent power in response to loss of control of water level for steam generator number 3. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. There is no impact to Unit 2. The manual actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50. 72(b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event.
ENS 5448722 January 2020 04:34:00SequoyahNRC Region 2

EN Revision Text: CONTAINMENT RELIEF VALVES INOPERABLE At 22:18 (EST) on 1/21/20, it was discovered that all Unit 1 containment vacuum relief isolation valves were closed and all vacuum relief lines were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The isolation valves were opened and the vacuum relief valves were restored to operable.

There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 02/20/2020 AT 1626 EST FROM FRANK SCHULTE TO BRIAN P. SMITH * * *

At 1549 (EST), February 20, 2020, a completed engineering evaluation of the condition initially reported on January 22, 2020 determined that the inoperability of the Sequoyah Unit 1 Containment Vacuum Relief System affected the ability to protect containment against an external pressure event. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The condition was resolved when isolation valves were opened on January 21, 2020 and the vacuum relief lines were restored to an operable status. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B), "an unanalyzed condition that significantly degrades plant safety. Subsequent to the initial notification, continued evaluation of the reported condition has concluded that the isolation of the containment vacuum relief function did not prevent the fulfillment of a safety function that is needed to control the release of radioactive material; nor mitigate the consequences of an accident therefore this event is not reportable under 10 CFR 50.72(b)(3)(v), "Event or Condition that could have prevented fulfillment of a safety function. The NRC Resident has been notified. Notified R2DO (Musser)

ENS 5444616 December 2019 09:12:00SequoyahNRC Region 2

At 0358 EST, on 12/16/2019, with Unit 1 in Mode 1 at 100 (percent) power and Unit 2 in Mode 1 at 47 (percent) power, a valid actuation of the Emergency Diesel Generators (EDG) occurred. The reason for the emergency diesel generator auto start was that the normal feeder breaker from the 1C 6.9KV Unit Board to the 1B-B 6.9KV Shutdown Board (SDBD) tripped due to the breaker's 51G relay actuating causing an under-voltage signal on the 1B-B 6.9KV Shutdown Board. All 4 Emergency Diesel Generators automatically started as designed when the 6.9KV Shutdown Board under-voltage signal was received.

The 1B-B 6.9KV Shutdown Board was automatically energized from the 1B-B 6.9KV Diesel Generator. All required 6.9KV loads were sequenced back on to the 1B-B 6.9KV Shutdown Board as designed after the board was energized from its emergency diesel generator. The remainder of the electrical system is in normal alignment.

This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Emergency Diesel Generators. There was no impact to the health and safety of the public or plant personnel. The NRC Senior Resident has been notified.

ENS 5443812 December 2019 08:14:00SequoyahNRC Region 2

At 0432 EST, on 12/12/19, Sequoyah Unit 2 experienced a manual reactor trip. The trip was initiated due to a loss all number 3 Feedwater Heater Drain Tank pump flow; plant procedures directed a manual reactor trip if power is greater than 80 percent.

The Auxiliary Feedwater System (AFW) automatically actuated as required when the expected post trip feedwater isolation actuation actuated. Reactor Coolant System (RCS) temperature is being maintained by the steam dump system with all 4 Reactor Coolant Pumps (RCPs) in service. All control and shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the transient. Unit 2 is currently stable at normal operating temperature and normal operating pressure in Mode 3. The electrical system is in a normal alignment. There was no impact on U1. There was no impact to the health and safety of the public or plant personnel. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10CFR50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification accordance with 10CFR50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. The NRC Resident Inspector was notified.

ENS 5439116 November 2019 03:02:00Watts BarNRC Region 2At 2353 EST on November 15, 2019, a Main Control Room (MCR) alarm was received for low control room positive pressure. At 2355 EST, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification 3.7.10 Control Room Emergency Ventilation System (CREVS) was declared not met for both trains. Watts Bar Unit 1 and Unit 2 entered Condition B. At 2355 EST on November 15, 2019, the alarm cleared, CREVS was declared operable and LCO 3.7.10 Condition B was exited for both units. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. See similar EN #54390. The licensee has taken compensatory measures while investigating the cause.
ENS 5439016 November 2019 03:02:00Watts BarNRC Region 2At 2234 Eastern Standard Time (EST) on November 15, 2019, a Main Control Room (MCR) alarm was received for low control room positive pressure. At 2236 EST, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification 3.7.10 Control Room Emergency Ventilation System (CREVS) was declared not met for both trains. Watts Bar Unit 1 and Unit 2 entered Condition B. At 2236 EST on November 15, 2019, the alarm cleared, CREVS was declared operable and LCO 3.7.10 Condition B was exited for both units. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5438212 November 2019 10:32:00Browns FerryNRC Region 2

On November 12, 2019, the Central Emergency Control Center (CECC) was removed from service for a planned facility upgrade project. The CECC is a common Emergency Operations Facility (EOF) for the TVA Nuclear sites (Browns Ferry / Sequoyah / Watts Bar). The duration of the upgrade project is approximately 75 days. If an emergency is declared requiring CECC activation during this period, an alternate CECC will be used. During this period, the alternate CECC will be staffed and activated using existing emergency procedures. This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the CECC will be unavailable for more than 72 hours. The Emergency Response Organization has been notified that the CECC will be unavailable during the upgrade project and to report to the alternate CECC in the event of an emergency. There is no impact on the health and safety of the public or plant employees. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 1316 EST ON 11/14/19 FROM BARUCH CALKIN TO JEFF HERRERA * * *

The event information was updated to indicate that the event occurred at 0700 EST. The NRC Resident Inspector has been notified. Notified the R2DO(Musser).

  • * * UPDATE FROM ALAN PRUCHA TO KERBY SCALES AT 1526 EST ON 1/31/2020 * * *

The CECC facility upgrade project is sufficiently complete such that the CECC was returned to a functional status at 1350 CST on January 31, 2020. The NRC Resident Inspector has been notified. Notified R2DO (Baptist).

ENS 5438012 November 2019 08:28:00SequoyahNRC Region 2

On November 12, 2019, the Central Emergency Control Center (CECC) was removed from service for a planned facility upgrade project. The CECC is a common Emergency Operations Facility (EOF) for the TVA Nuclear sites (Browns Ferry / Sequoyah / Watts Bar). The duration of the upgrade project is approximately 75 days. If an emergency is declared requiring CECC activation during this period, an alternate CECC will be used. During this period, the alternate CECC will be staffed and activated using existing emergency procedures. This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the CECC will be unavailable for more than 72 hours. The Emergency Response Organization has been notified that the CECC will be unavailable during the upgrade project and to report to the alternate CECC in the event of an emergency. There is no impact on the health and safety of the public or plant employees. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM BRYAN KLEIN TO DONALD NORWOOD AT 1450 EST ON 1/31/2020 * * *

The CECC facility upgrade project is sufficiently complete such that the CECC was returned to a functional status at 1400 EDT on January 31, 2020. The NRC Resident Inspector has been notified. Notified R2DO (Baptist).

ENS 5437912 November 2019 07:57:00Watts BarNRC Region 2

On November 12, 2019, the Central Emergency Control Center (CECC) was removed from service for a planned facility upgrade project. The CECC is a common Emergency Operations Facility (EOF) for the TVA Nuclear sites (Browns Ferry / Sequoyah / Watts Bar). The duration of the upgrade project is approximately 75 days. If an emergency is declared requiring CECC activation during this period, an alternate CECC will be used. During this period, the alternate CECC will be staffed and activated using existing emergency procedures. This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the CECC will be unavailable for more than 72 hours. The Emergency Response Organization has been notified that the CECC will be unavailable during the upgrade project and to report to the alternate CECC in the event of an emergency. There is no impact on the health and safety of the public or plant employees. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM JUSTIN GALLAGHER TO KERBY SCALES AT 1620 EST ON 1/31/2020 * * *

The CECC facility upgrade project is sufficiently complete such that the CECC was returned to a functional status at 1400 EST on January 31, 2020. The NRC Resident Inspector has been notified. Notified R2DO (Baptist)

ENS 5434121 October 2019 15:52:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On December 29, 2018, at approximately 0220 Central Standard Time (CST), Browns Ferry Nuclear Plant (BFN), Unit 3 experienced an unexpected loss of power to the 3A Reactor Protection System (RPS) Bus due to the trip of the 3A RPS motor generator (MG) set. This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. This event is being reported as a late 60 day non-emergency notification. This missed notification was identified on August 23, 2019. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the trip of the RPS MG Set was a failure of the motor winding insulation of all three phases. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1478564 and 1543534. The NRC Resident Inspector has been notified of this event.
ENS 5433216 October 2019 10:22:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 20, 2019, at approximately 1133 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS MG Set trip was dirty potentiometer windings on an Over Voltage Relay. The dirt prevented the potentiometer's wiper from contacting its windings, resulting in erratic setpoint values. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1542603, 1542608, and 1542569. The NRC Resident Inspector has been notified of this event.
ENS 543021 October 2019 07:05:00Browns FerryNRC Region 2On 10/1/2019, at 0307 CDT, Unit 2 was conducting a normal reactor startup and received a valid Reactor Protection System (RPS) scram. The reactor was critical in MODE 2 at the Point of Adding Heat. Operators began withdrawing Source Range Monitor (SRM) Instrumentation per procedure. When the operator depressed the SRM Drive Out pushbutton to withdraw the last two SRMs (C and D), an unexpected full Reactor Scram was received. Annunciator indication in the Main Control Room indicated a Neutron Monitoring Scram. The Intermediate Range Monitors (IRM) D, E, F, H and G all indicated Upscale High High. There were no Emergency Core Cooling System (ECCS) or Containment Isolation System actuations. All other systems functioned as designed. The cause of the Reactor Scram is still under investigation. This event requires a 4-hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' The NRC Resident Inspector has been notified.
ENS 5430030 September 2019 10:28:00Browns FerryNRC Region 2This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On July 31, 2019, at approximately 1650 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 1 experienced a Primary Containment Isolation System (PCIS) Group 6 isolation during performance of surveillance procedure 1-SR-3.3.6.2.3(A), Reactor/Refueling Zone Ventilation Radiation Monitor 1-RM-90-140/142 Calibration and Functional Test. The Group 6 isolation caused the initiation of Standby Gas Treatment (SBGT) Trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem B. Unit 1 H2O2 Analyzer and Drywell Radiation Monitor CAM, 1-RM-90-256, were declared Inoperable and Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.5 Condition B was entered. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 6 actuations are Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. This condition was the result of two cleared fuses in the alarm logic. The apparent cause is a ground fault on the A6 Open Drain Input/Output Module. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Acton Program as Condition Report 1537358. The NRC Resident Inspector has been notified of this event.
ENS 5427813 September 2019 12:23:00Browns FerryNRC Region 2

EN Revision Text: EMERGENCY OPERATING FACILITY UNAVAILABLE DUE TO ACCESS ISSUES This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. A condition impacting access to the Emergency Operating Facility, Central Emergency Control Center (CECC), located in the TVA Chattanooga Office Complex occurred on September 13, 2019 at 0427 CDT. Fire suppression capabilities for the TVA Chattanooga Office Complex are currently impacted by a water main failure rendering access to the facility unsafe for personnel. If an emergency is declared requiring CECC activation during this period, other emergency response centers will be activated and staffed using existing emergency planning procedures and have the capability to perform the functions normally performed by the CECC. This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the condition affects the functionality of an emergency response facility. The condition does not affect the health and safety of the public. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 9/19/2019 AT 1334 EDT FROM NEEL SHUKLA TO ANDREW WAUGH * * *

Water lines impacting the Chattanooga Office Complex were repaired, and as of time 0734 EDT on 9/16/19, the CECC was returned to a functional status. The NRC Resident Inspector has been notified. Notified R2DO (Ehrhardt).

ENS 5427713 September 2019 12:21:00Watts BarNRC Region 2

EN Revision Text: EMERGENCY OPERATING FACILITY UNAVAILABLE DUE TO ACCESS ISSUES This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. A condition impacting access to the Emergency Operating Facility, Central Emergency Control Center (CECC), located in the TVA Chattanooga Office Complex occurred on September 13, 2019 at 0527 EDT. Fire suppression capabilities for the TVA Chattanooga Office Complex are currently impacted by a water main failure rendering access to the facility unsafe for personnel. If an emergency is declared requiring CECC activation during this period, other emergency response centers will be activated and staffed using existing emergency planning procedures and have the capability to perform the functions normally performed by the CECC. This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the condition affects the functionality of an emergency response facility. The condition does not affect the health and safety of the public. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 9/19/2019 AT 1341 EDT FROM DEAN BAKER TO ANDREW WAUGH * * *

Water lines impacting the Chattanooga Office Complex were repaired, and as of time 0734 EDT on 9/16/19, the CECC was returned to a functional status. The NRC Resident Inspector has been informed of this event update. Notified R2DO (Ehrhardt).

ENS 5427613 September 2019 11:57:00SequoyahNRC Region 2

EN Revision Text: EMERGENCY OPERATING FACILITY UNAVAILABLE DUE TO ACCESS ISSUES This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. A condition impacting access to the Emergency Operating Facility, Central Emergency Control Center (CECC), located in the TVA Chattanooga Office Complex occurred on September 13, 2019 at 0527 EDT. Fire suppression capabilities for the TVA Chattanooga Office Complex are currently impacted by a water main failure rendering access to the facility unsafe for personnel. If an emergency is declared requiring CECC activation during this period, other emergency response centers will be activated and staffed using existing emergency planning procedures and have the capability to perform the functions normally performed by the CECC. This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the condition affects the functionality of an emergency response facility. The condition does not affect the health and safety of the public. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 09/16/2019 AT 1148 EDT FROM SCOTT THOMAS TO BRIAN LIN VIA PHONE * * *

Water lines impacting the Chattanooga Office Complex were repaired, and as of time 0734 EDT on 9/16/19, the CECC was returned to a functional status. The NRC Resident Inspector has been informed of this event update. Notified R2DO (Ehrhardt).

ENS 5426611 September 2019 03:10:00Browns FerryNRC Region 2A lightning strike occurred at approximately 1502 CDT on 09/10/2019, and a resulting power surge damaged some of the security door card reader system equipment. However, this did not affect access to plant areas for personnel who were already within protected area. At 1830 on 09/10/2019, it was discovered that some of the oncoming night shift personnel could not access particular areas that required the use of security card readers. Extent of condition check at 1934 on 09/10/2019 determined that access to 1A and 3A Electric Board Rooms, which contain remote shutdown panels and Fire Safe Shutdown equipment. was prohibited for the night shift personnel. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B) - Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Access was restored to all plant areas at 2106 on 9/10/2019. No plant events occurred during the time frame that the 1A & 3A Electric Board Rooms inaccessible that would have required access to these areas. The NRC Resident Inspector has been notified.
ENS 542521 September 2019 00:09:00Watts BarNRC Region 2While operating at 100 percent power, the Watts Bar Unit 1 reactor was manually tripped at 2055 EDT on August 31, 2019 due to loss of steam generator #2 level control. The trip was not complex. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems, including Auxiliary Feedwater, actuated as designed. Operations responded and stabilized the plant. Decay heat is being removed by the Auxiliary Feedwater and the Steam Dump System. Unit 2 is not affected. The cause of the loss of steam generator water level control is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The automatic actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified.
ENS 5424227 August 2019 02:34:00SequoyahNRC Region 2At 0109 EDT, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a dropped rod causing a negative rate trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 2 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5422014 August 2019 20:00:00Watts BarNRC Region 2This 60-day telephone notification is being submitted in accordance with paragraphs 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) to report an invalid Containment Ventilation Isolation (CVI) actuation at Watts Bar Nuclear Plant (WBN) Unit 2. On July 26, 2019, at 1003 Eastern Daylight Time (EDT), the Train A CVI actuated due to an invalid High Radiation signal from 2-RM-90-130, Containment Purge Air Exhaust Monitor. Prior to and following the invalid High Radiation alarm, all radiation monitors except 2-RM-90-130 were stable at their normal values. All required automatic actuations occurred as designed. Upon investigation, the cause of the invalid High Radiation alarm was due to a failed ratemeter for 2-RM-90-130. Control room operators performed appropriate checks and confirmed that the subject indication was an invalid high radiation signal. The ratemeter for 2-RM-90-130 was replaced and the monitor returned to service. At the time of the event, plant conditions for a High Radiation alarm did not exist; therefore, the CVI was invalid. The NRC Resident Inspector was notified.
ENS 5416212 July 2019 22:50:00Browns FerryNRC Region 2At 1640 CDT on 7/12/19, Unit 1 High Pressure Coolant Injection (HPCI) received an invalid auto isolation signal which closed the HPCI steam supply valves rendering HPCI inoperable. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D), as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The isolation occurred while performing a calibration and functional check of a level switch for the Unit 1 Core Spray system. Continuity was checked across the incorrect set of contacts which completed the circuit in logic bus 'A' for the auto isolation signal in the HPCI system. There was no impact to the safety of the public or plant personnel during the time HPCI system was isolated. HPCI was returned to operable at 2110 CDT on 7/12/19. CR 1532094 documents this condition in the Corrective Action Program. The licensee has notified the NRC Resident Inspector
ENS 5412720 June 2019 17:59:00Watts BarNRC Region 2At 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.
ENS 5411212 June 2019 11:55:00Watts BarNRC Region 2At 0849 (EDT), a significant air leak on an inline air filter was identified. At 0908, the leak on the filter was isolated. A subsequent review of this situation determined that this air leak impacted operation of the A Train of the Control Room Emergency Air Temperature Control System (CREATCS) which is required to be operable in accordance with Technical Specification 3.7.10. At the time of this event, the B Train of CREATCS was out of service for planned maintenance. With both trains of CREATCS out of service, both Watts Bar Units entered a condition that could have prevented fulfillment of a safety function. This condition was terminated when the leaking air filter was isolated. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). NRC Resident Inspector has been notified.
ENS 5407722 May 2019 05:45:00Watts BarNRC Region 2On May 22, 2019, at 0233 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a failure of the #2 Main Feedwater Regulating Valve during power ascension following a refueling outage. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Dumps. Unit 2 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact to WBN Unit 1. The NRC Senior Resident has been notified.
ENS 5402123 April 2019 09:44:00Watts BarNRC Region 2At 0232 EDT on April 23, 2019, a Main Control Room (MCR) alarm was received for low control room positive pressure. At 0233 EDT, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification 3.7.10 Control Room Emergency Ventilation System (CREVS) was declared not met for both trains. Watts Bar Unit 1 entered Condition B. Watts Bar Unit 2 was not performing movement of irradiated fuel assemblies and did not meet the APPLICABILITY for CREVS per LCO 3.7.10. At 0233 EDT on April 23, 2019, the alarm cleared, CREVS was declared operable and LCO 3.7.10 Condition B was exited. The safety function of the CRE boundary is to ensure the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of Design Basis Accident (DBA) consequences to CRE occupants. From 0232 EDT to 0233 EDT, (Watts Bar Nuclear) WBN was unable to validate that CREVS could fulfill its required Safety Function. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
ENS 5399914 April 2019 06:44:00SequoyahNRC Region 2

EN Revision Text: AUTOMATIC REACTOR TRIP DUE TO MAIN FEEDWATER PUMP TRIP At 0320 EDT, April 14, 2019, Sequoyah Unit 1 experienced an automatic reactor trip. The event was initiated by the trip of the 1A main feedwater pump. During the automatic unit runback, an automatic reactor trip was initiated due to low-low level in Steam Generator number 3. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. During this operational cycle, one control Rod Position Indicator (RPI) for core position E-5 in shutdown bank 'A' has been inoperable, and the appropriate Condition and Required Actions of (Technical Specification Limiting Condition of Operation) 3.1.7 were complied with. Due to this inoperable RPI, the associated shutdown rod is conservatively assumed to be full out and untrippable. Consequently, boration was required to establish adequate shutdown margin. All other Control and Shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. There was no impact on Unit 2. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.

  • * * UPDATE ON 8/6/19 AT 12:20 EDT FROM KEVIN MICHAEL TO KERBY SCALES * * *

The licensee provided an update to paragraph 2. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. All Control and Shutdown rods fully inserted, except E-5 which was previously identified and conservatively assumed to be in a full out position. Applicable TS actions were performed to maintain shutdown margin. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. Notified the R2DO (Gerald McCoy)

ENS 5395926 March 2019 16:08:00Browns FerryNRC Region 2On 3/26/2019 at 1030 CDT Engineering evaluation determined that Traversing lncore Probe (TIP) System test results related to Leak Rate Testing of 2-CKV-76-653, TIP Purge Header Check Valve, during the Unit 2 Refueling Outage resulted in a reportable condition. On 3/24/2019 at 1558 CDT, Leak Rate Testing identified a (local leak rate test) LLRT failure of 2-CKV-76-653. The gross leakage Leak Rate value exceeded the Technical Specification allowable value for Type C valves of less than 0.6 (allowable leakage) La. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The short-term corrective actions include repairing the valve such that it passes the test. The valve needs to be repaired before the unit can change modes.
ENS 5394217 March 2019 14:10:00Browns FerryNRC Region 2

EN Revision Text: HIGH PRESSURE COOLANT INJECTION SYSTEM DECLARED INOPERABLE At 0735 CDT on March 17, 2019, the High Pressure Coolant Injection (HPCI) system was isolated due to a water-side leak from the HPCI Gland Seal Condenser. Unit 3 declared the HPCI system Inoperable and entered Technical Specification LCO 3.5.1 Condition C with required actions to verify the Reactor Core Isolation Cooling system is Operable, and to restore the HPCI system to Operable status within 14 days. All other Unit 3 Emergency Core Cooling Systems (ECCS) remain Operable. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(V)(D), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(V)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this event.

  • * * RETRACTION FROM WESLEY CONKLE TO HOWIE CROUCH ON 4/23/19 AT 1549 EDT * * *

ENS Event Number 53942, made on March 17, 2019, is being retracted. NRC Notification 53942 was made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 (b)(3)(v)(D) were met when the licensee discovered an event, that at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. At 0735 CDT, on March 17, 2019, during the performance of a routine surveillance, a momentary pressure transient of 844 psig from the Feedwater system was introduced into the High Pressure Coolant Injection (HPCI) system discharge and suction piping that ruptured the seal on the gland seal condenser and flooded the U3 HPCI Room. Unit 3 HPCI was declared inoperable due to isolation of the waterside of the HPCl system. On April 11, 2019, a Past Operability Evaluation was completed which determined that the HPCI System remained operable. The evaluation of the potential pressure transient and room flooding concluded that the HPCI System could have performed its specified safety function of vessel injection throughout the time that the gland seal was ruptured. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(D). TVA's evaluation of this event is documented in the Corrective Action Program in Condition Report 149973. The licensee has notified the NRC Resident Inspector. Notified R2DO (Ehrhardt).

ENS 5392310 March 2019 04:38:00Browns FerryNRC Region 2At 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes. Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A). 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser.
ENS 5392210 March 2019 00:48:00Browns FerryNRC Region 2

At 0012 EST on 3/10/2019, Browns Ferry Unit-3 declared an Unusual Event due to a spurious trip of the generator breaker, resulting in a loss of AC power to the 4 kV shutdown boards greater than 15 minutes. All diesel generators started and loaded to supply onsite power. The reactor auto-scrammed, with all rods fully inserting. The Main Steam Isolation Valves opened and shutdown cooling was being conducted via the condenser. The licensee will exit the emergency declaration once offsite power is restored. There is no estimated restart date. Browns Ferry Unit 1 remains in Mode-1 (100%), Unit 2 remains in Mode-5 for a refueling outage. The NRC Resident Inspector has been notified. This event is related to EN 53923. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 3/10/19 AT 1419 EDT FROM JOHN HOLLIDAY TO BETHANY CECERE * * *

At 1310 CDT, Browns Ferry Unit-3 exited the Unusual Event when 161 kV lines were made available. The licensee is executing procedures for securing the diesel generators while alternate offsite power methods are utilized. Switchyard damage evaluation is in progress. The licensee will notify the NRC Resident Inspector. Notified R2DO (Desai), R2RA (Haney), DNRR (Nieh), NRR EO (Miller), and IRD (Grant). Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

ENS 539188 March 2019 15:51:00Browns FerryNRC Region 2Browns Ferry Nuclear Plant (BFN) is notifying state and local agencies of the presence of an oil sheen in the cold water channel. Water from the cold water channel was running into a tunnel that connects to the waters of the US. BFN Procedure RWI-007, Spill Prevention Control and Countermeasure Plan requires the National Response Center as well as other state and local agencies be notified of any oil sheen on the water. This oil spill is reportable to the EPA (National Response Center) under 40 CFR 112. The notification was made to the National Response Center at 1113 CST under notification number 1239580. The Alabama Emergency Management Agency (AEMA) and Alabama Department of Environmental Management (ADEM) were notified at 1120 CST. This event is reportable as a 4-hour Non-Emergency Notification report in accordance with 10 CFR 50.72(b)(2)(xi) 'Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the NRC Resident Inspector. The oil is believed to come from the number one cooling tower basin due to heavy rainfall.
ENS 5384022 January 2019 09:41:00Watts BarNRC Region 2This 60-day telephone notification is being submitted in accordance with paragraphs 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) to report an invalid Containment Ventilation Isolation (CVI) actuation at Watts Bar Nuclear Plant (WBN) Unit 1. On December 2, 2018 at 0028 Eastern Standard Time (EST), the Train A CVI actuated due to an invalid High Radiation signal from 1-RM-90-130, Containment Purge Exhaust Radiation Monitor. In addition to the Train A CVI, instrument malfunction alarms were received for 1-RM-90-106, Lower Containment Radiation Monitor and 1-RM-90-112, Upper Containment Radiation Monitor as the associated valves isolated for the CVI. A common instrument malfunction alarm was also received for 1-RM-90-130 and 1-RM-90-131, Containment Purge Exhaust Radiation Monitors. Prior to and following the invalid High Radiation alarm, all radiation monitors except 1-RM-90-130 were stable at their normal values. All required automatic actuations occurred as designed. Upon investigation, the cause of the invalid High Radiation alarm was due to a failed ratemeter for 1-RM-90-130. Control room operators performed appropriate checks and confirmed that the subject indication was an invalid high radiation signal. The ratemeter for 1-RM-90-130 was replaced and the monitor returned to service. At the time of the event, plant conditions for a High Radiation alarm did not exist; therefore, the CVI was invalid. The NRC Resident Inspector was notified.
ENS 5380121 December 2018 00:02:00Watts BarNRC Region 2At 1642 Eastern Standard Time (EST) on December 20, 2018, it was determined that both trains of Containment Air Return Fan (CARF) were simultaneously INOPERABLE from 0817 (EST) to 1129 (EST) on November 20, 2018. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
ENS 5375426 November 2018 08:31:00SequoyahNRC Region 2

At 0816 EST, a Notification of Unusual Event was declared for Unit 2 under Emergency Action Level H.U.4 for excessive smoke in the lower level of containment with a heat signal. Onsite fire brigade is responding to the event. A command post is established. Offsite support is requested by the fire brigade. No flames have been observed as of this report. The NRC Resident Inspector and State and Local government agencies will be notified. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 11/26/18 AT 1036 EST FROM BILL HARRIS TO JEFFREY WHITED * * *

At 1036 EST, Sequoyah Nuclear Station Unit 2 terminated the Notice of Unusual Event. The licensee determined that the source of the smoke in containment was oil on the pressurizer beneath the insulation, that heated up during plant heatup. The licensee did not see visible flame during the event. The licensee is still working to determine if there was any damage to the pressurizer. The licensee will notify the NRC Resident Inspector. Notified R2DO (Rose), R2RA (Haney), NRR (Nieh), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 11/26/18 AT 1337 EST FROM STEPHEN FRIESE TO KARL DIEDERICH * * *

Following declaration of the Notification of Unusual Event, TVA media relations communicated with the local media regarding the event. The licensee has notified the NRC Resident Inspector. Notified R2DO (Rose).

  • * * UPDATE ON 11/26/18 AT 1551 EST FROM STEPHEN FRIESE TO DONG PARK * * *

At 1036 EDT, Sequoyah Nuclear Plant (SQN) terminated the Notification Of Unusual Event (NOUE) due to initial report of heat and smoke in Unit 2 Lower Containment. At 1000 EDT, it was determined that no fire had occurred. Due to difficulty of access to some of the areas being searched, the source could not be identified prior to 1000 EDT. No visible flame (heat or light) was observed. The source of the smoke was determined to be residual oil from a hydraulic tool oil in contact with pressurizer piping. The pressurizer piping was being heated up to support Unit 2 start-up following U2R22 refueling outage. Once the residual oil dissipated, the smoke stopped. It has been concluded that no fire or emergency condition existed. Unit 2 is currently in Mode 5, maintaining reactor coolant temperature 160F-170F and pressure 325psig-350psig with 2A Residual Heat Removal (RHR) system in service in accordance with U2R22 refueling outage plan. The licensee has notified the NRC Resident Inspector. Notified R2DO (Rose).

  • * * RETRACTION ON 11/29/2018 AT 1358 EST FROM FRANCIS DECAMBRA TO ANDREW WAUGH * * *

Sequoyah Nuclear Plant (SQN) is retracting this notification based on the following additional information not available at the time of the notification: Following a full Reactor Building inspection, it was concluded that a fire did not exist. The source of the smoke originally reported was later determined to be residual oil from a hydraulic tool in contact with pressurizer piping. Once the residual oil dissipated, the smoke stopped. The source of heat originally reported was normal heated conditions associated with the pressurizer commensurate with plant conditions. SQN reported initially based on the available information at the time and to ensure timeliness with emergency declaration and reporting notification requirements. The licensee has notified the NRC Resident Inspector. Notified R2DO (Shaeffer).

ENS 5375124 November 2018 21:27:00SequoyahNRC Region 2At 1420 (EST) on November 24, 2018, operators discovered that a door was blocked open creating a breach of the auxiliary building secondary containment enclosure (ABSCE) boundary that exceeded the allowed ABSCE breach margin (of three minutes). As a result, Unit 1 entered Technical Specification Limiting Condition of Operation (LCO) 3.7.12 Condition B for two trains of Auxiliary Building Gas Treatment System (ABGTS) inoperable due to an inoperable ABSCE boundary in MODE 1, 2, 3, or 4, and both Units entered Condition E for one required ABGTS train inoperable with fuel stored in the spent fuel pool. In MODES 1, 2, 3, and 4, the analysis of the loss of coolant accident (LOCA) assumes that radioactive materials leaked from the Emergency Core Cooling System are filtered and absorbed by the ABGTS. For the fuel handling accident, the analysis assumes that the ABSCE boundary is capable of being established to ensure releases from the auxiliary and containment buildings are consistent with the dose consequence analysis. The event is reportable in accordance with 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to: (C) control the release of radioactive material and (D) mitigate the consequences of an accident. No actual LOCA or fuel handling accident occurred while both trains of ABGTS were inoperable. The condition had no impact on the health and safety of the public. The NRC Resident Inspector has been notified. This situation occurred because of maintenance activities. A breeching permit had been initiated however, the required personnel to ensure the door could be closed within the required three minutes were not assigned. The door was closed approximately 15 minutes after the situation was noticed.