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 Entered dateSiteRegionReactor typeEvent description
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.
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 5369727 October 2018 16:52:00Watts BarNRC Region 2On October 27, 2018, at 1533 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. (Main Steam Isolation Valves) MSIVs were required to be isolated due to cooldown. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident has been notified."
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 5355722 August 2018 11:23:00Watts BarNRC Region 2At 0943 EDT on August 22, 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 reactor automatically tripped due to a main turbine trip signal. 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."
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 5339811 May 2018 15:19:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 1011 EDT on May 11, 2018, Containment Shield Building Annulus differential pressure exceeded the required limit. The Shield Building was declared inoperable requiring entry into Technical Specification (TS) 3.6.15 Conditions A and B. The event was initiated by failure of the operating annulus vacuum fan. Main Control Room Operators manually started the stand-by annulus vacuum fan to recover pressure. Shield Building Annulus differential pressure was restored to the required value at 1016 EDT and TS 3.6.15 Condition A and B were exited on May 11, 2018 at 1016 EDT. The failure mechanism for the annulus vacuum fan is being investigated. The Containment Shield Building ensures the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analysis during a Loss of Coolant Accident (LOCA). The Emergency Gas Treatment System (EGTS) would have automatically started and performed its design function to maintain the Shield Building Annulus differential pressure within required limits. The event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident has been notified.
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 5334017 April 2018 12:02:00Browns FerryNRC Region 2GE-4At 0416 CDT on April 17, 2018, the High Pressure Coolant Injection System (HPCI) was isolated due to a water side leak from the gland seal condenser. Unit 1 HPCI remains inoperable pending repairs to the gland seal condenser. 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.
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 5330029 March 2018 22:28:00Browns FerryNRC Region 2GE-4At 1344 on March 29, 2018, it was determined (engineering evaluation) that an unanalyzed condition that significantly degraded plant safety previously existed. During a postulated control room abandonment due to a fire, and concurrent with a Loss of Offsite Power (LOOP), the required number of Emergency Equipment Cooling Water (EECW) pumps would not have been available from 10/28/2015 to 3/10/2018. On March 8, 2018, during relay functional testing it was discovered that the C3 Emergency Equipment Cooling Water (EECW) pump closing springs did not recharge with the breaker transfer switch in emergency. On August 23, 2012, a wire modification was performed that contained a drawing error resulting in wire placement on the incorrect connection points for the C3 EECW pump. On March 10, 2018, the C3 EECW pump breaker wiring was corrected and subsequent testing was completed satisfactorily. Prior to 10/28/2015, Brown's Ferry Nuclear Plant (BFN) adhered to Appendix R fire protection requirements which did not credit the C3 EECW pump for fire protection from the backup control location. On 10/28/2015, BFN transitioned to National Fire Protection Association (NFPA) 805 fire protection requirements which takes credit for the C3 EECW pump from the backup control location. This condition is being reported pursuant to 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'. This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(ii)(B). 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.
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 5328826 March 2018 11:51:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 1058 Eastern Daylight Time (EDT) on March 26, 2018, a Main Control Room (MCR) alarm was received for low control room positive pressure. At 1100 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 1100 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 1058 EDT to 1100 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). The NRC Resident Inspector has been notified.
ENS 5327019 March 2018 02:27:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 3/16/2018 at approximately 1630 EST, an industrial safety accident occurred at Sequoyah Nuclear Plant that involved an Arc Flash injury of two contract employees. While performing work near a non safety related 6.9kV electrical bus, an arc occurred injuring the two employees. Both personnel were transported to an offsite medical facility for treatment. Neither were contaminated. The cause of the arc flash is not understood at this time, an accident investigation has been initiated by TVA. The SQN (Sequoyah Nuclear) NRC Senior Resident Inspector has been notified. No safety related systems required to establish or maintain safe shutdown were affected. Both Unit 1 and 2 remain at 100 (percent) power. TVA has received and responded to media inquiries concerning this event. As a result, this event is considered reportable under 10CFR50.72(b)(2)(xi).
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 5326716 March 2018 22:04:00Browns FerryNRC Region 2

At 1604 (CDT) on March 16, 2018, Browns Ferry Nuclear Plant (BFN) Engineering reported an unanalyzed condition affecting the Residual Heat Removal (RHR) heat exchangers in a postulated fire event. It was discovered that the Residual Heat Removal Service Water (RHRSW) heat exchanger piping associated (with) the credited heat exchangers in the NFPA 805 Nuclear Safety Capability Analysis (NSCA) could experience water hammer damage. Fire damage to the cables for the RHRSW outlet motor operated valves could cause the valves to spuriously open and drain the RHRSW piping. Subsequent starting of the RHRSW pumps on the affected header could cause water hammer loads and damage the piping. Review of NFPA 805 analyses show the cables associated with this condition are routed in Fire Areas 01-03, 02-03, 02-04, 03-03, 16 and 23. There are 11 cases where the deterministically credited heat exchanger could be affected. Compensatory fire watch measures have been established. This event requires an 8 hour report in accordance with 10CFR50.72(b)(3)(ii)(B), 'Any event or condition that results in: (B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. CR 1139620 documents this condition in the Corrective Action Program. The NRC Resident Inspector has been notified.

  • * * RETRACTION AT 2215 EST ON 11/29/2018 FROM NEEL SHUKLA TO MARK ABRAMOVITZ * * *

NRC notification 53267 was made to ensure that the eight-hour non-emergency reporting requirements of 10 CFR 50.72 were met when the licensee discovered an unanalyzed condition with the potential to significantly degrade plant safety. On August 22, 2018, an independent analysis was completed which determined that the RHRSW system would remain functional during the postulated scenario. Based on this analysis, a revised functional evaluation was performed by BFN which determined that the condition did not constitute an unanalyzed condition that significantly degraded plant safety. For credited RHR heat exchangers for fire events in Fire Areas 01-03, 02-03, 02-04, 03-03, 16, and 23, the RHRSW piping will remain intact and the valves will operate manually after a water hammer event. This condition did not significantly degrade plant safety and is therefore not reportable under 10 CFR 50.72(a)(2)(ii)(B). On November 16, 2018, TVA canceled the 60 day report which had been submitted for this condition. TVA's evaluation of this event notification is documented in the corrective action program. The licensee has notified the NRC Resident Inspector. Notified the R2DO (Shaeffer).

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 5317317 January 2018 22:41:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 2002 EST on January 17, 2018, annulus differential pressure exceeded its pressure limit. At that time, the Shield Building was declared inoperable requiring entry into Technical Specification 3.6.15, Conditions A and B. Action was taken by field operators to swap annulus vacuum control dampers to restore annulus differential pressure. At 2024 EST, annulus differential pressure was restored to required limits, the Shield Building was declared operable, and LCO 3.6.15, Conditions A and B were exited. The temporary loss of the Shield Building resulted from a failure of the annulus vacuum control system to maintain the required differential pressure. Manual swap-over of pressure control to the backup damper restored differential pressure to required limits allowing exit from TS LCO 3.6.15 and restoration of the Shield Building safety function. 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). The Emergency Gas Treatment System (EGTS) would have automatically started and performed its design function to maintain annulus vacuum within required limits. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
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.