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 Entered dateSiteRegionReactor typeEvent description
ENS 4783425 February 2020 07:37:00Browns FerryNRC Region 2GE-4On March 14, 2012, it was determined that in the event of an Appendix R fire, fire damage to cables in certain fire areas could cause a Residual Heat Removal Service Water System (RHRSW) pump to spuriously start, overload EDG A and B, and render them inoperable during certain Appendix R fires. This was reported as an unanalyzed condition (Ref. EN #47764). An extent of condition analysis was completed on April 13, 2012. From this analysis it was determined that EDG A, D, 3EC, and 3ED could exceed the maximum rated loading due to the potential for an automatic or spurious start of RHRSW Pumps B3 and D3 that supply Emergency Equipment Cooling Water (EECW) to essential safety equipment. The following are the Fire Areas (FA) affected: EDG A in FA 21 EDG D for FA 2-3 and 9 EDG 3EC in FA 1-1, 1-3, and 20, and EDG 3ED in FA 1-1, 1-3, 1-4, and 20. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60 day written report IAW 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as PER 536176. 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 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 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 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 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 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 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 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.
ENS 5375022 November 2018 03:56:00Browns FerryNRC Region 2

EN Revision Text: HPCI UNEXPECTEDLY TRANSFERRED TO ALTERNATE SUCTION SOURCE DURING TESTING At 2125 (CST) on 11/21/2018, it was discovered that U1 High Pressure Coolant Injection System (HPCI) was inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 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. During performance of a routine surveillance, HPCI automatically transferred from its normal suction source to the alternate suction source. The control room operator then manually tripped the HPCI turbine. HPCI was already inoperable in accordance with Technical Specifications (TS) Limiting Condition for Operability (LCO) 3.5.1, ECCS Operating, Condition C during performance of the surveillance. However, this condition was not expected nor induced by the testing. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. CR 1469109 documents this condition in the Corrective Action Program.

  • * * RETRACTION ON 12/28/18 AT 1300 EST FROM MARK MOEBES TO JEFFREY WHITED * * *

ENS Event Number 53750, made on November 22, 2018, is being retracted. NRC notification 53750 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 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. During performance of a routine surveillance, the High Pressure Coolant Injection (HPCI) System automatically transferred from its normal suction source to the alternate suction source. As a result, Unit 1 HPCI was declared inoperable. On December 20, 2018, a Past Operability Evaluation was completed which determined that the HPCI System remained operable. The evaluation determined that the HPCI System could have performed its specified safety function of vessel injection throughout the time that the suction path was aligned to the torus. 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 1469109. The licensee has notified the NRC Resident Inspector. Notified R2DO (Desai).

ENS 5368021 October 2018 19:58:00SequoyahNRC Region 2This notification is being made due to the death of an employee on-site. A Security Officer was found unresponsive on the Turbine Building Moisture Separator Re-heater deck on the Unit 1 side. Upon arrival of Fire Operations and on-site medical the individual had suffered an apparent heart attack. Hamilton County Emergency Medical Services will be transferring the individual to the medical examiner's office. The on-site NRC Senior Resident Inspector has been notified. The licensee believes this event may receive media attention and a press release could be issued.
ENS 5367821 October 2018 06:19:00Browns FerryNRC Region 2

At 0200 Central Daylight Time on 10/21/2018, Browns Ferry Nuclear Plant Unit 3 commenced a reactor shutdown as required by the Technical Requirements Manual Limiting Condition for Operation 3.4.1 Coolant Chemistry Condition D due to conductivity greater than 10 micro mho/cm at 25 degrees Celsius. The required action for this condition is to immediately initiate an orderly shutdown and be in Mode 4 as rapidly as cooldown rate permits. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION AT 1719 EST ON 12/13/2018 FROM NEEL SHUKLA TO MARK ABRAMOVITZ * * *

ENS Event Number 53678, made on 10/21/18, is being retracted. NRC notification 53678 was made to ensure that the four-hour non-emergency reporting requirements of 10 CFR 50.72 were met when the licensee discovered a condition requiring shut down of a reactor. 10 CFR 50.72 requires a report in accordance with 50.72(b)(2)(i) for any Technical Specifications (TS) required reactor shutdown. NUREG-1022 only specifies TS applicability and makes no mention of a Technical Requirements Manual (TRM) required shutdown. Because the shutdown comes from the TRM and not the TS as discussed in 10 CFR 50.72 and NUREG-1022, an EN was not required. TVA's evaluation of this event notification is documented in the corrective action program. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ehrhardt).

ENS 5366111 October 2018 15:37:00Browns 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 16, 2018, at approximately 1736 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2B 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 with the exception of the Unit 1 Refuel Zone Supply Fan Outboard Isolation Damper, 1-FCO-64-5, that failed to indicate closed position. 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 (Motor Generator) Set trip was a failed (shorted) operating coil associated with the 480 VAC motor starter inside the control box. 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 1440047 and 1440050. The NRC Resident Inspector has been notified of this event.
ENS 5359311 September 2018 05:17:00Watts BarNRC Region 2At 0113 EDT on September 11, 2018, it was discovered both trains of CREVS (control room emergency ventilation system) were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The door to the main control room habitability zone from the turbine building was left open and unattended for about a minute, breaking the pressure boundary in the room, resulting in an alarm. The door was closed, clearing the alarm and the CREVS was considered operable.
ENS 5355822 August 2018 19:35:00Browns FerryNRC Region 2On 08/22/2018 at 1803 hours CDT, Browns Ferry Nuclear Plant declared an Unusual Event per EAL HU4, a fire potentially degrading the level of safety of the plant. At 1748 CDT Unit 1 received a call reporting smoke coming from the 480V Condensate Demineralizer Panel 3 in the Unit 3 turbine building elevation 557'. At approximately 1803 (CDT), the incident commander on the scene confirmed a fire inside the panel and all three units entered 0-AOI-26-1, Fire Response. The board was subsequently de-energized by operations personnel and the fire was extinguished at 1806 CDT. SM (Shift Manager) exited EAL HU4 and all three units exited 0-AOI-26-1 at 1840 CDT. Fire operations remain on scene to monitor. A team is being assembled for damage assessment and recovery. The fire did not affect any safety systems, no plant transients resulted, and no injuries were reported. This event is reportable within 1 hour IAW 10 CFR 50.72(a)(1)(i). The NRC Resident inspector has been notified. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email).
ENS 534979 July 2018 17:01:00Browns FerryNRC Region 2On 07/09/2018 at 1111 CDT, Browns Ferry Unit 1 Operators identified U1 High Pressure Cooling Injection system steam supply valves were isolated. After reviewing ICS (Integrated Computer System), Operations determined isolation occurred at 0958 CDT during performance of surveillance testing. The Browns Ferry Nuclear Plant Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable at 0958 CDT due to an inadvertent isolation that occurred during testing. During performance of surveillance procedure 1-SR-3.3.6.1.2(3B) HPCI System Steam Supply Low Pressure Functional test, an erroneous signal was induced causing actuation of primary containment isolation system group IV (i.e., HPCI Isolation). Technical Specification 3.5.1, ECCS-Operating, Condition C was entered as a result of the inoperable HPCI system. This constitutes an unplanned HPCI system inoperability and requires an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(v)(D). The erroneous signal was cleared and the HPCI isolation was reset. Upon reset of the isolation signal, the HPCI system was returned to available status. The HPCI system was unavailable for 2 hours and 55 minutes, however the HPCI system remains inoperable. There was no impact to the health and safety of the public or plant personnel as a result of this condition. The NRC Resident Inspector has been notified. A condition report has been entered into the Licensee's Corrective Action Program to capture this event.
ENS 5346722 June 2018 12:14:00Watts BarNRC Region 2At 0841 EDT on June 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 95% power. All control and shutdown bank rods inserted properly in response to the automatic 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. The reactor automatically tripped due to a main turbine trip. The turbine trip was caused by main generator electrical trip. An investigation is in progress. The automatic 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. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5346220 June 2018 11:39:00Browns FerryNRC Region 2On June 20, 2018 at 1003 CDT, the licensee declared a Notification of Unusual Event based on Emergency Action Level (EAL) 6.5.U, toxic gas release on site. The Notification of Unusual Event was terminated at 1025 CDT. The toxic gas release occurred when site personnel were filling a fire suppression carbon dioxide (CO2) tank outside the diesel generator building. The relief valve in the common diesel generator room for Unit 1 and 2 diesel generators inadvertently lifted causing a toxic gas environment by releasing CO2 into the room. The licensee terminated the tank fill stopping the release of CO2, and with the door to the room being opened, the gas cleared in about 20 minutes. The licensee has notified the NRC Resident Inspector. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email).
ENS 5335622 April 2018 04:28:00Watts BarNRC Region 2Westinghouse PWR 4-Loop

On April 22, 2018 at 0222 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 0227 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM TONY PATE TO HOWIE CROUCH ON 5/4/18 AT 1455 EDT * * *

This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 2 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector staff has been informed of this event retraction. Notified R2DO (Desai) of this retraction.

ENS 5335522 April 2018 02:34:00Watts BarNRC Region 2Westinghouse PWR 4-Loop

On April 21, 2018 at 2152 EDT, Watts Bar Nuclear Plant (WBN) Unit 1 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 2222 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM ANTHONY PATE TO DONALD NORWOOD AT 1310 EDT ON 5/9/2018 * * *

This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 1 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector has been informed of this event retraction. Notified R2DO (Ehrhardt).

ENS 5334920 April 2018 00:55:00Watts BarNRC Region 2Westinghouse PWR 4-LoopOn April 19, 2018 at 1944 EDT, Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows current acceptance criteria for gas accumulation in the WBN Unit 1 and Unit 2 Safety Injection System (SIS) and Residual Heat Removal System (RHRS) discharge piping may be non-conservative. The surveillances that check void values and allow venting of the systems are to be performed utilizing conservative criteria at more frequent intervals to ensure gas void volumes remain under acceptable limits. Additional analysis is being performed to determine final actions. The NRC Resident Inspector has been notified.
ENS 5332712 April 2018 12:14:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 0920 EDT on April 12, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic 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. The cause of the automatic reactor trip is being investigated. The automatic 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. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5329126 March 2018 20:07:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 1839 Eastern Daylight Time (EDT) on March 26, 2018, a Main Control Room (MCR) alarm was received for low control room positive pressure. At 1840 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 and Condition B entered. At 1840 EDT on March 26, 2018, the alarm cleared, CREVS was declared operable and LCO (Limiting Condition for Operation) 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 1839 EDT to 1840 EDT, WBN (Watts Bar Nuclear) 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). A watch has been posted at the door to prevent recurrence. The NRC Resident Inspector has been notified.
ENS 5326918 March 2018 16:16:00Browns FerryNRC Region 2GE-4At 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. All safety system operated as expected. At no time was public health and safety at risk. This event is reportable within 4 hours 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.' It is also reportable within 8 hours 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) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 532477 March 2018 12:25:00Browns FerryNRC Region 2GE-4

The licensee declared an Unusual Event based on Emergency Action Level (EAL) 6.7.U and entry into the site Security Plan. All required actions or compensatory measures have been completed. The Notice of Unusual Event was terminated at 1142 CST. There was no impact to the operation of any of the units at the Browns Ferry site. The licensee has notified the NRC Senior Resident Inspector. See EN #53248. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email).

  • * * UPDATE AT 1816 EST ON 03/07/2018 FROM DAVID RENN TO JEFF HERRERA * * *

The licensee provided additional information regarding the event. Notified the R2DO (Musser), IRD MOC (Gott), NRR EO (Miller).

ENS 531964 February 2018 12:00:00Watts BarNRC Region 2Westinghouse PWR 4-Loop

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.

  • * * RETRACTION AT 1336 EST ON 03/29/2018 FROM TONY PATE TO TOM KENDZIA * * *

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 5316210 January 2018 13:53:00Browns FerryNRC Region 2GE-4At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours 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 preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 5313220 December 2017 18:18:00Watts BarNRC Region 2Westinghouse PWR 4-LoopOn December 20, 2017, at 1040 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) 1B-B 6.9kV Shutdown Board (SDBD) normal feeder breaker opened. The loss of voltage to the 1B-B SDBD resulted in the start of the 1B-B Motor Driven Auxiliary Feedwater (MDAFW) pump, the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) pump, and the start of all four Emergency Diesel Generators (EDGs). Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Following initial investigation, the 1B-B 6.9 kV SDBD was transferred to its alternate offsite power source, Common Station Service Transformer (CSST) C at 1217 EST. At 1230 EST, the 1B-B 6.9 kV SDBD alternate feeder breaker opened. The loss of voltage to the 1B-B SDBD did not result in the restart of the 1B MDAFW pump, the Unit 1 TDAFW pump, or EDGs; this equipment remained running from the earlier event. Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Restoration of normal offsite power to the 1B-B SDBD was completed at 1654. Other than several common Unit Technical Specifications having not been met, Unit 2 was not operationally impacted by the transfer of the 1B-B Shutdown Board to onsite power and remains in Mode 1 at 100% power. This report is made per 10 CFR 50.72(b)(3)(iv)(A). NRC Resident Inspector has been notified. The licensee investigation continues for the cause of the event.
ENS 5309226 November 2017 16:16:00Watts BarNRC Region 2Westinghouse PWR 4-LoopOn November 26, 2017, at 1225 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) Unit 2 experienced an unplanned ECCS discharge to the Unit 2 Reactor Coolant System (RCS) while de-pressurized, in Mode 5, with the Pressurizer vented to the Pressurizer Relief Tank. ECCS injection via the Boron Injection flow path occurred during planned Safety Injection system Engineered Safety Features Actuation System (ESFAS) testing. The Boron Injection flow path should have been isolated and should not have resulted in any injection flow to the Unit 2 RCS. Since the injection was not a part of the pre-planned test this is reportable under 10 CFR 50.72(b)(2)(iv), System Actuation. All other systems responded as expected in accordance with the ESFAS testing procedure. The unintended ECCS injection flow was isolated and flow through the Boron Injection path was verified to be stopped at 1232 EST. The Unit 2 Pressurizer level and pressure remained below any limits and no safety limits were challenged. NRC Resident Inspector has been notified.
ENS 5307014 November 2017 15:13:00Browns FerryNRC Region 2GE-4This 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 September 15, 2017, during a TVA (Tennessee Valley Authority) review of Operations logs, it was determined that a reportable condition occurred in January 2017 but no NRC report had been made. On January 10, 2017, at 0300 Central Standard Time (CST), Browns Ferry Nuclear Plant, Unit 3, received Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals. The Group 2, 3, 6, and 8 isolations caused the initiation of all three trains of the Standby Gas Treatment (SBGT) system and Control Room Emergency Ventilation (CREV) subsystem 'A.' At 0311 CST, Operations personnel discovered that the 3A1 RPS circuit protector had tripped on undervoltage. 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 Drywall 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 Drywall Pressure. At the time of the event, these conditions did not exist; therefore the actuation of the PCIS was invalid. All affected equipment responded as designed. This condition was the result of an undervoltage condition on the 3A1 circuit protector. During trouble shooting, the undervoltage setpoints were found to be 116 VAC and 115 VAC, when the normal as left acceptance band is 109.7 VAC to 111.3 VAC. The 3A RPS protective relays had been previously replaced in September 2016. The most likely cause of the undervoltage condition in these relays is infant mortality. The NRC Resident Inspector has been notified of this event.