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 Entered dateSiteRegionReactor typeEvent description
ENS 5399914 April 2019 06:44:00SequoyahNRC Region 2At 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."
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 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 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 5276924 May 2017 04:10:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn May 23, 2017 at 2330, while transferring 2A-A 6.9 kV Shutdown Board from its alternate power source to its normal power source in support of outage testing, a failure occurred which resulted in the loss of the Shutdown Board, emergency start of all 4 Emergency Diesel Generators (EDGs), and required the manual emergency stop of 2A-A EDG. During transfer of the 2A-A 6.9kV Shutdown Board, the hand switch for the normal feeder breaker on the shutdown board was being maintained in the 'CLOSE' position while the alternate feeder breaker hand switch was placed in 'TRIP.' As expected, the alternate feeder breaker opened and the normal feeder breaker closed. However, the upstream supply breaker to the normal feeder breaker immediately tripped due to an overcurrent relay actuation on a single phase. As a result, the 2A-A 6.9 kV Shutdown Board deenergized, initiating a blackout signal which started all 4 of the station's EDGs. During board stripping (opening of all feeder and load breakers, to prepare the board for automatic reenergization from the EDG), the normal feeder breaker to the Shutdown Board failed to trip. This failure to trip prevented the emergency feeder breaker in the output of 2A-A EDG from closing, in accordance with interlock logic. As a result, 2A-A 6.9 kV Shutdown Board remained deenergized which prevented the cooling water supply valve for the EDG from opening due to loss of motive power. This lack of cooling caused operators to emergency stop the 2A-A EDG. Power was restored to the Shutdown Board on May 24, 2017 at 0037. Unit 1 is currently stable in Mode 1, at 100% power and Unit 2 is stable in Mode 5 with RCS at 164 F and 340 psig. The cause of the breaker trip on overcurrent and the failure of the normal feeder to trip on load shedding are under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified.
ENS 525977 March 2017 14:39:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 0830 (EST) on March 7, 2017, operators discovered that on March 3, 2017 at 2046 a door was blocked open creating a breach of the auxiliary building secondary containment enclosure (ABSCE) boundary that exceeded the allowed ABSCE breach margin. As a result, both Unit 1 and Unit 2 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. The condition has been corrected and ABGTS was restored to operable as of 0949 March 7, 2017. 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 adsorbed 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 a 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.
ENS 5246930 December 2016 16:37:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 12/30/16 at 1302 EST, Unit 1 began withdrawing control bank rods for an approach to criticality following a refueling outage. At 1305 EST, operators observed that control rod H-6 did not withdraw. Operators entered the applicable Abnormal Operating Procedure. Operators tripped the reactor as required by plant procedures and entered applicable Emergency Operating Procedures. The act of manually tripping the reactor generated a valid trip signal in the plant Reactor Protection System. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater was already supplying the Steam Generators; a Feedwater Isolation occurred due to plant conditions. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators, and decay heat removal via the steam dumps. Method of decay heat removal is Steam Generators via the steam dumps. Current reactor coolant system conditions: Temperature at 548 degrees F and stable, pressure 2235 psig and stable. All control and shutdown banks are inserted. Electrical alignment is normal, supplied by offsite power. No impact to Unit 2, Unit 2 is in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector.
ENS 5242613 December 2016 14:40:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 12/13/16 at 1410 (EST), the following voluntary communication was made to the State of Tennessee in accordance with Tennessee Valley Authority's (TVA) guidance for communicating inadvertent radiological spills/leaks that are below regulatory reporting requirements to outside agencies and in alignment with NEI 07-07, 'Industry Ground Water Protection Initiative'. On 12/12/16, Sequoyah Nuclear Plant determined that a spill of greater than 100 gallons (approximately 3000 gallons) of condensate storage tank water with tritium levels of 1560 pCi/L (picocuries per liter) was spilled to a yard drain. The spill occurred on 12/5/16, during the filling of the Unit 1 #4 steam generator when a hose connection on a temporary fill skid failed. No elevated tritium levels have been detected at the Sequoyah Yard Drainage Pond before or after the event. This is reported in accordance with 10CFR50.72(b)(2)(xi), the required reportable threshold for tritium is 20,000 pCi/L. The licensee will notify the NRC Resident Inspector.
ENS 5218717 August 2016 17:46:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1722 (EDT) on 8/17/16, a Past Operability Evaluation (POE) determined the configuration of the Emergency Gas Treatment System (EGTS) flow controllers that existed prior to 0420 on 8/6/16 constituted an Unanalyzed Condition due to not meeting single failure criteria. This POE examined the condition where EGTS may auto-swap from the flow control path in A-Auto to the Standby flow control path upon the start of a Design Basis Event (DBE). The intended design of the EGTS swap over flow control path in Auto to Standby was to detect and respond to an actual failure of the A-Auto flow control path. The unnecessary auto-swap to Standby could prevent the EGTS train configured in Auto from performing its required safety function during a DBE. The POE performed a detailed calculation to determine the release effects due to the failure of the redundant trains of EGTS controllers. These calculations concluded that failure of both trains of EGTS controllers would not result in exceeding the 10CFR100 limits, however this condition was unanalyzed and failed to meet single failure criteria. This condition is reportable under 10CFR50.72(b)(3)(ii)(B), Unanalyzed Condition due to a system required to meet the single failure criterion does not do so. This condition had no impact to the health and safety of the public. The NRC Resident Inspector has been notified.
ENS 5217211 August 2016 16:35:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1015 (EDT) on August 11, 2016, it was discovered that a Fire Protection damper associated with the Control Room Emergency Ventilation System had closed unexpectedly due to component failure. The closure rendered both trains of the Control Room Emergency Ventilation System (CREVS) inoperable requiring both Unit 1 and 2 to enter Technical Specification Limiting Condition of Operation (LCO) 3.7.10 Condition G. Condition G requires immediate entry into LCO 3.0.3. At 1159 on August 11, 2016, actions were taken to block the deficient damper in the open position restoring both trains of CREVS to a fully operable condition and allowing exit of LCO 3.0.3 and 3.7.10 Condition G. The purpose of CREVS is to provide a protected environment from which occupants can control the (respective) unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. In the event of a design basis accident, emergency ventilation components realign to supply filtered air and to pressurize the Control Room Envelop (CRE). While the damper was closed both trains of CREVS were incapable of supplying the Relay Room as well as the Technical Support Center and its associated support spaces. These locations constitute part of the CRE, therefore both trains of CREVS were inoperable. Both trains of CREVS being inoperable affected the habitability of the TSC where the assessment capability of the facility for all emergencies was adversely effected. The NRC Resident Inspector has been notified.
ENS 519948 June 2016 17:10:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

At 1526 Eastern Daylight Time on 6/8/2016, a determination was made involving the potential impact of a tornado on the Emergency Diesel Generators (EDGs). The EDGs are required to be operable to provide power to ensure that acceptable fuel design limits, reactor coolant system pressure boundary limits, and containment integrity are not exceeded during abnormal transients. Further, the EDGs are designed with a crankcase pressure trip (setpoint = 1 inch water), which is bypassed during an emergency start. Engineering has determined that a tornado could potentially cause actuation of the crankcase pressure trip due to a low barometric condition. If an emergency start signal has NOT previously occurred, then during a tornado, actuation of the crankcase pressure trip would energize the shutdown relay causing an EDG lockout condition. The EDG lockout condition prevents subsequent EDG starts (normal or emergency) until operators manually reset the lockout condition locally at the EDG. This condition could potentially affect all four EDGs simultaneously. The EDGs are operable but degraded. All EDGs have successfully passed their required surveillances within the appropriate frequency. No severe weather warnings or watches are forecast in the local areas, which could challenge the crankcase pressure trip.

This condition places both units in an unanalyzed condition that potentially significantly degrades plant safety, 10 CFR 50.72 (b)(3)(ii)(B). A compensatory measure has been established, that upon notification of a Tornado Warning, the EDGs would be 'emergency started' and run during the time the Tornado Warning was in effect. This action bypasses the crankcase pressure trip function and allows the EDGs to perform their required safety function. The NRC Senior Resident Inspector has been notified.

ENS 5193516 May 2016 21:17:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

On May 16, 2016 at 2105, Sequoyah Nuclear Power Plant identified a nonconforming condition involving the Emergency Diesel Generator (EDG) fire dampers installed in Units 1 and 2. Specifically, it has been identified that if a tornado causes a differential pressure across the east and west sides of the EDG Building, this could create a high airflow rate through the EDG Building ventilation path. The fire dampers for each EDG bay (required to isolate the space for CO2 fire suppression per SQN Fire Protection Report) have not been analyzed to withstand high air flows resulting from a tornado and could possibly fail in a way that impedes airflow for EDG cooling. This is an unanalyzed condition that could prevent all EDGs from supplying electrical power as designed during a tornado or other similar weather events. All 4 EDGs are required to be operable by both units' Technical Specifications to provide electrical power to safe shutdown/safety related equipment following accident conditions coincident with a loss of offsite power. The Current Licensing Basis (CLB) requires that tornado effects be considered in the design of safety related SSCs (Systems, Structures, and Components), and it cannot be demonstrated at this time that the described SSCs will withstand the design basis tornado. It has been determined that the CLB may not adequately address possible design basis tornado scenarios.

The EDGs are located inside the power plant structure and are currently capable of performing their safety function. The occurrence of such an event is highly unlikely and there is no imminent concern regarding severe weather involving tornadoes. Compensatory measures have been developed to address the associated nonconformance. The condition described above is being reported as an unanalyzed condition that significantly degrades plant safety per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.

ENS 519003 May 2016 15:13:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 0833 (EDT) on May 3, 2016, security received an alarm on a Main Control Room (MCR) door. At 0847 (EDT), security notified the MCR staff of the door alarm and that the door was incapable of closure. At this time, both control room ventilation filtration trains (CREVS) were declared inoperable in accordance with Technical Specification 3.7.10, Condition B, due to the inoperability of the Control Room Envelope (CRE). Attempts to close the door were made, and it was identified that a screw had become wedged at the base of the door preventing it from latching. At 0855 (EDT), the screw was removed and the door verified to close as designed. CREVS was determined to be 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. In addition to an intact CRE boundary maintaining CRE occupant dose from a large radioactive release below the calculated dose in the licensing basis consequence analysis for DBAs, it also ensures the occupants are protected from hazardous chemicals and smoke. This condition had no impact to the health and safety of the public. The NRC Resident Inspector has been notified.
ENS 518547 April 2016 18:01:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThis notification is being made as the result of the review of an occurrence on March 30, 2016 at 2220 (EDT) that resulted when a major portion of the site high pressure fire protection (HPFP) system, including fire suppression capabilities for the Main Turbine Building, Auxiliary Building, Control Building, Diesel Generator Buildings, and both Unit 1 and Unit 2 Containments were isolated without having the required compensatory suppression systems established. Upon discovery of the non-functional HPFP system, compensatory fire watches were established and an alternate means to provide water to the HPFP system was aligned. A review of the Sequoyah Nuclear Plant (SQN) Safe Shutdown Analysis identified this loss of fire suppression may not have ensured the required equipment remained available under certain postulated fire scenarios. The analysis determined that the effects of a postulated fire in specific fire areas could have prevented critical systems or components from performing their intended functions, potentially resulting in the inability to achieve and maintain safe shutdown. Analysis identified areas which credit the availability of fire suppression to assure that the safe shutdown capability could have been achieved, the site did not have fire suppression for approximately 45 hours. No actual fire occurred or existed during the time the fire suppression system was not functional. Installed fire detection equipment and communication to the Main Control Room remained available. The condition has been corrected and the HPFP system is functional. At the time of the non-functional HPFP system, it was not recognized that an unanalyzed condition that could have significantly degraded plant safety existed. The condition placed both Unit 1 and Unit 2 in an unanalyzed condition that significantly degraded plant safety and is reportable under 10 CFR 50.72(b)(3)(ii)(B). This 8-hour non emergency notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(B). The condition has been entered into the licensee's corrective action program (CR 1155763) and a License Event Report will be submitted. The NRC Resident Inspector has been notified of this condition. The original clearance that created this event was satisfactory as written, however, one of the valves was leaking and the clearance boundaries were expanded. The clearance was issued at 1411 EDT on 3/29/2016.
ENS 517209 February 2016 17:25:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1415 EST on 02/09/2016, Sequoyah Unit 1 was at 0 percent power (mode 3, 526F, 2235 psig) when a low steam line pressure Safety Injection actuated from Loop 2 Steam Generator. Prior to this event, the Loop 2 Main Steam Isolation Valve bypass was opened at 1413 EST for main steam line warm up in preparation for unit startup. Loop 2 Main Steam Isolation Valve bypass closed automatically following low steam line pressure Safety Injection. Following the Safety Injection, all safety-related equipment operated as designed. Current Reactor Coolant System temperature and pressure - Unit 1 is currently being maintained in Mode 3 at approximately 517 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via steam generator atmospheric relief valves. There is no indication of any primary to secondary leakage. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. The cause of the Safety Injection actuation is under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. Due to RCS pressure, the only system that injected into the RCS was the charging system. The AFW system initiated to feed the steam generators and the Emergency Diesel Generators started but did not load.
ENS 5165414 January 2016 16:07:00SequoyahNRC Region 2Westinghouse PWR 4-LoopA non-licensed supervisory employee tested positive for alcohol during a random fitness for duty test. The individual's plant access has been denied. The NRC Resident Inspector has been notified.
ENS 5155923 November 2015 11:52:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 0820 EST on 11/23/2015, Sequoyah Unit 1 was at 100% power when operators identified the Loop #3 Main Steam Isolation Valve (MSIV) came off its full open seat. This was evidenced by no OPEN indication on the main control board, dual indication on the post accident monitoring panel, and a change in both Tavg and steam pressure. Operators were dispatched locally to the MSIV and to the battery board room to investigate if a cause could be identified for the MSIV movement. The field investigation identified no issues. The operating crew manually tripped the reactor at 0844 EST due to an increasing Tavg-Tref mismatch. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the feedwater isolation signal. Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure, approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. There is no indication of any primary to secondary leakage. All control rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2 as it continues through the refueling outage with the core off-loaded. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. All control rods fully inserted during the reactor trip. The atmospheric steam dumps did operate during the transient and then shut. After the trip, the MSIV re-opened.
ENS 5152710 November 2015 20:13:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 11/10/2015 at 1502 (EST), Unit 2 MCR (main control room) was notified by workers in containment that 2 ice suits had been dropped into the Unit 2 Containment Reactor Cavity Equipment Pit. Based upon size and location of the dropped suits, Unit 2 entered LCO 3.6.15 (Containment Recirculation Drains) Condition B and LCO 3.0.3 for refueling canal drains being inoperable. The two refueling canal drains and the ice condenser drains function with the ice bed, Containment Spray System and ECCS to limit the pressure and temperature that could be expected following a DBA (Design Basis Accident). Following performance of a Safety Function Determination it was determined that, during the short duration when both coats were in the process of being retrieved, they could have potentially clogged the drains and prevented the fulfillment of safety functions if there was a DBA. Both suits were retrieved from the equipment pit by 1556 (EST) and all LCO conditions were exited. The licensee notified the NRC Resident Inspector.
ENS 514515 October 2015 10:00:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThis 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the Unit 2, Train B Containment Ventilation Isolation (CVI) at Sequoyah Nuclear Plant. At 1919 EDT on August 7, 2015, during planned performance of a Unit 2 containment vent, the Train B CVI actuated due to an invalid Hi Rad signal from 2-RM-90-131, Containment Vent Radiation Monitor. In addition to the Train B CVI alarm, unexpected alarms were received for 2-RM-90-106, Lower Containment Radiation Monitor and 2-RM-90-112, Upper Containment Radiation Monitor instrument malfunctions as they isolated for the CVI and 2-RM-90-131 Hi Rad alarm. Prior to the invalid Hi Rad alarm, all radiation monitors were stable at their normal values. All required automatic actuations occurred as designed. Upon investigation, the cause of the invalid Hi Rad alarm was due to an exposed shield wire at the 2-RM-90-131 detector. Preventative maintenance had been performed the week prior to the CVI and it is believed the damage occurred at that time. Control Room Operators performed Annunciator Response actions and verified by diverse indications that the subject condition was an invalid Hi Rad signal. There were no indications of degraded reactor coolant system parameters or fuel failure. Applicable Technical Specification (TS) Limiting Condition for Operations (LCOs) were entered and the radiation monitors declared inoperable. No Emergency Response criteria were applicable with the subject radiation monitors inoperable. Radiological surveys performed in the vicinity of 2-RM-90-131 verified no abnormal radiological conditions. Radiation Monitor 2-RM-90-131 was removed from service, the shield wire was repaired and returned to service with no issues. Radiation Monitors 2-RM-90-106 and 2-RM-90-112 were tested and returned to service. The applicable TS LCOs were exited. At the time of the event, plant conditions for a Hi Rad alarm did not exist; therefore, the CVI was invalid. The NRC Resident Inspector was notified.
ENS 5139214 September 2015 08:12:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 0426 EDT on 9/14/2015, Sequoyah Unit 1 was at 100% power when the Vital Instrument Power Board (VIPB) 1-II deenergized. A manual reactor trip was initiated in accordance with the Abnormal Operating Procedure for the loss of VIPB 1-II. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP (Normal Operating Temperature/Normal Operating Pressure), approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 547 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary to secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100 (percent). There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector.
ENS 5126527 July 2015 13:44:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1043 EDT on 7/27/2015, Sequoyah Unit 1 was at 82% power and continuing to perform a startup when the reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3, at NOT/NOP (normal operating temperature and normal operating pressure), approximately 545 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. The immediate cause of the trip was an electrically-induced turbine trip. Due to fluctuating voltage the main generator voltage regulator was taken to manual; immediately after this the unit tripped. Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 545 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector.
ENS 5125924 July 2015 17:04:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1351 EDT on 7/24/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. The immediate cause of the trip was an electrically-induced turbine trip. There was no associated work in progress related to this and all systems were normally aligned. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. The 2B-B Emergency Diesel Generator is currently in service for the performance of an unrelated surveillance. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. The cause of the main generator lockout is under investigation.
ENS 511174 June 2015 11:18:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThis 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the Train B Phase A Containment Isolation at Sequoyah Nuclear Plant. At 1320 EDT on April 12, 2015, during planned performance of the Containment Isolation Train-A, Phase A Isolation Testing and Emergency Gas Treatment System (EGTS) Cleanup System Test, the main control room received several Train-B annunciators. Upon investigation, it was determined that an invalid signal to the Train-B Solid State Protective System (SSPS) actuated the Train B, Phase A Containment Isolation. The invalid isolation signal was the result of a human performance error during the performance of the Phase A Isolation Test surveillance procedure. Operations personnel responded to the SSPS initiation, testing was aborted, ensured that all equipment operated as designed and restored affected systems in accordance with plant procedures. Approval to restart testing was obtained. All prerequisites were met and testing of the SSPS Train-A, Phase A Isolation was completed satisfactorily. As part of the prerequisite test alignment of the Train-A, Phase A, Unit 2 had entered a planned 7 day action for EGTS being inoperable. During the test when the Train-B of Phase A actuated, the suction dampers for Unit 1 supply to EGTS were closed per plant procedures. This prevented Train-B EGTS from aligning to Unit 1 and allowed Train-B of EGTS to remain operable for Unit 2. An SSPS Phase A signal can be generated automatically by a Safety Injection Signal (SIS) or manually. At the time of the event, plant conditions for an SIS did not exist; therefore, the Phase A actuation was invalid. The licensee notified the NRC Resident Inspector.
ENS 5087811 March 2015 09:30:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt (0621 EDT) on 3/11/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a Power Range Nuclear Instrumentation negative rate signal, caused by a malfunction in the rod control system. There was no associated work in progress related to this and all systems were normally aligned. Current Temperature and Pressure - temperature is 547 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector.
ENS 508562 March 2015 10:22:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 06:45 EST on 3/2/2015, Sequoyah Unit 2 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 2 is currently being maintained in Mode 3 at NOT/NOP (normal operating temperature and pressure), approximately 548 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a main generator C phase differential relay actuation. There was no associated work in progress related to this and all systems were normally aligned. It is currently not understood why the relay actuated. Current Temperature and Pressure - temperature is 548 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. The 2B Emergency Diesel Generator is currently unavailable for planned maintenance and will be returned to service prior to unit restart. There is no operational impact to Unit 1. Unit 1 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector" and the state. No primary or secondary safety valves lifted during this event.
ENS 5085327 February 2015 17:04:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

On February 27, 2015, from 1305 to 1323 (EST), a loss of the ENS (Emergency Notification System) communication line, the HPN (Health Physics Network) line, and the other communication systems that the MCR (Main Control Room) operators use as part of the emergency plan occurred at TVA Sequoyah Units 1 and 2. A loss of these systems was experienced when transferring from a backup generator to an uninterruptible power supply (UPS) during fueling of the backup generator. Earlier in the day the normal power supply to the two (2) plant communications buildings was deenergized due to an unrelated electrical issue and backup diesel generators were supplying power to the buildings. While refueling one of the backup generators, the power supply was transferred to the UPS for personal safety reasons, however, the load on the UPS was greater than its capacity and its output (breaker) tripped.

This loss of communications was corrected in approximately 18 minutes once the fuel tank was filled and the generator was started. Phone line capability has been restored and verified to be functional. Sequoyah is currently investigating why the load was greater than the UPS capacity. The radiological emergency plan (REP) was reviewed and other acceptable methods for communications were available. There were no personnel injuries as a result of this event and no impact on plant operations. Both units remain at 100% power. The NRC Resident Inspector has been notified. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the loss of communications capability which includes events that would significantly impair the ability of the licensee to implement the functions of its emergency plans if an emergency were to occur.

ENS 5075723 January 2015 15:09:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 1/23/2015 at 1150 EST, an oil sheen was verified to be in the waters of the U.S. (Sequoyah's Intake Forebay). At 1221 (EST), a notification to the National Response Center (report #1106407) was made as a result of the oil sheen (< 1 pint total). The source of the oil is from temporary connections in support of maintenance. Cleanup efforts are in progress. The following additional agencies have also been notified: EPA Region 4, Tennessee Department of Environment and Conservation (TDEC), and Tennessee Emergency Management Agency (TEMA). The licensee also notified the Hamilton County Emergency Management Agency and the NRC Resident Inspector.
ENS 507176 January 2015 15:31:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

On January 6, 2015, at approximately 0001 EST, a partially empty can of beer was found by a security officer. Site Security took possession of the can and it was removed from the protected area. An investigation has been initiated by Site Security. This condition has been documented in TVA's corrective action program. This report is submitted pursuant to 10 CFR 26.719 (b)(1) based on the presence of alcohol within the protected area. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM ZACHARY KITTS TO JEFFREY HERRERA AT 1503 EST ON 1/15/15 * * *

Additional information provided by the Licensee of the origination of the beer can. The licensee notified the NRC Resident Inspector. Notified the R2DO (Musser).

ENS 5069619 December 2014 14:57:00SequoyahNRC Region 2Westinghouse PWR 4-LoopA contract supervisory employee had a confirmed positive test for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5058130 October 2014 12:48:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn October 30, 2014, at 1100 EDT, TVA conducted a briefing for government officials and other stakeholders regarding the decision to accelerate the Boone Reservoir annual drawdown after discovery of a sink hole near the base of the embankment and a small amount of water and sediment found seeping from the river below the dam. TVA is continuously monitoring the dam and conducting an investigation to determine the source of the water seepage. The dam is located upstream of all three TVA nuclear sites. There are currently no nuclear plant operability or safety issues, and TVA is assessing the impacts on the plant licensing bases The licensee has notified the NRC Resident Inspector.
ENS 5024430 June 2014 23:11:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

On June 27, 2014, TVA identified in a reanalyzed hydrologic analysis for Sequoyah Nuclear Plant (SQN) a deviation from the current hydrologic analysis. The flooding analysis in Section 2.4.3 of the SQN UFSAR assumes that the Watts Bar West Saddle Dike fails completely and instantaneously at approximately 1.5 feet of overtopping during a Peak Maximum Flood (PMF). This assumption exists in the original design basis analysis and the revised analysis which supports SQN-TS-12-02, "Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis". The results of recent studies of the West Saddle Dike, conducted as part of the Fukushima Order 2.1 flooding review, indicate that the complete and instantaneous failure of the Watts Bar West Saddle Dike may not be a valid assumption. If the dike does not fail, analyses performed using the codes and methods consistent with those used in original plant design show that the east floodwall of the Watts Bar Dam would overtop. As a result of this overtopping, the east floodwall is assumed to fail. Based on this assumption and analysis, failure of the east floodwall of the Watts Bar Dam would result in an increase in the flood level at the SQN Plant Site. The current licensing basis PMF level for SQN is 719.6 feet as stated in Section 2.4.2.2 of the SQN UFSAR. In addition, it should be noted that by letter dated August 10, 2012, as supplemented by letters dated April 5, 2013 and January 16, 2014, TVA proposed a revised PMF level of 722.0 feet. Introducing non failure of the Watts Bar West Saddle Dike indicated a potential increase of approximately 1.5 feet over the revised PMF level. TVA performed additional analysis using current industry standard for flooding analysis. Specifically, TVA modeled the condition using the United States Army Corps of Engineers Hydrologic Engineering Center River Analysis System (HEC-RAS) tool. TVA's analysis of the condition using HEC-RAS determined that all required safety equipment for SQN would not be impacted and are considered operable based on a Prompt Determination of Operability completed on June 30, 2014. This report addresses a condition as described in 10 CFR 50.72 (b)(3)(ii)(B). TVA is making this report consistent with the guidance of NUREG-1022 regarding the application of engineering judgment to the evaluation of reportability of an unanalyzed condition. The NRC Resident Inspector has been notified of this condition.

  • * * RETRACTION AT 1441 EDT ON 8/21/2014 FROM MATT LEENERTS TO MARK ABRAMOVITZ * * *

On June 30, 2014, SQN reported (Event 50244) that during a re-analysis conducted as part of the Fukushima Order 2.1 flooding review, a probable maximum flood (PMF) design assumption that the Watts Bar Dam west saddle dike fails completely and instantaneously at approximately 1.5 feet of overtopping, was determined to be a non-conservative flood model assumption (i.e., invalid). As a result, TVA postulated that Watts Bar Dam's east floodwall would fail, increasing the site flood level at Sequoyah Nuclear Plant (SQN) by 1.5 feet; a condition that was beyond the current licensing basis. Through subsequent analysis, TVA has demonstrated that although the west saddle dike may not completely and instantaneously fail during a PMF (as previously assumed), the consequential increase in reservoir levels does not result in a failure of the Watts Bar Dam east floodwall and would not result in an increase in the flood level at SQN. Therefore, the previously reported 10 CFR 50.72(b)(3)(ii)(B) event is being retracted. The NRC Resident Inspector has been informed of this event retraction. Notified the R2DO (Hickey).

ENS 4973817 January 2014 14:54:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn January 17, 2014, at 1400 EST Sequoyah (SQN) notified the Tennessee Department of Environment and Conservation (TDEC) of a condition that was identified on December 4, 2013, while conducting core samples. The core drilling contractor detected a petroleum odor, notified the SQN Environmental Group who observed the core drilling and confirmed a slight petroleum odor from a core boring in which water was detected at approximately 41 feet below the ground surface. At that time, as a result of the type of core drilling that was conducted, TVA could not verify if there were petroleum products in the groundwater. SQN obtained geoprobe boring samples to determine the extent and the depth of the petroleum. Petroleum product was detected in the soil between 30 and 35 feet in depth. This condition has been determined to be associated with a legacy diesel fuel oil leak at SQN in the 1990's and was previously reported. SQN previously completed a remediation project from 1992 - 2010, with oversight from TDEC. The petroleum product appears to be isolated at approximately 30 to 35 feet in depth, is in a perched water table not utilized for drinking water, and appears to be isolated to the diesel building area. To verify the petroleum product is isolated; TVA is conducting additional geoprobe boring to help define the limited area impacted by the legacy diesel fuel oil leak. The Sequoyah NRC Resident Inspectors have been notified.
ENS 496903 January 2014 22:07:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

At 1500 EST on 01/03/2014, TVA determined that during certain conditions, Service Air usage (air used for non-safety related tools/equipment) could result in introducing air into the Auxiliary Building Secondary Containment Enclosure that could, in worst-case conditions, exceed the margin required to maintain the Auxiliary Building Gas Treatment System (ABGTS) operable for Sequoyah Units 1 and 2. ABGTS is required to be operable for both units by Technical Specifications. This is an unanalyzed condition that could prevent both trains of ABGTS from performing (their) safety function(s). Service air has been isolated to the Auxiliary Building and is under administrative controls until further analysis (is) complete. This is additional information discovered during follow-up evaluation regarding the issue identified in LER 50-327/2013-004. Further analysis will be performed to determine safety significance. There is 1600 scfm margin in the ABSGTS. The Service Air compressors have an 1850 scfm capacity. The licensee informed the NRC Resident Inspector.

  • * * RETRACTION FROM BRUCE BUCH TO DANIEL MILLS AT 1428 EST ON 1/30/2014 * * *

Sequoyah Nuclear Plant, Units 1 and 2, are retracting the 8 hour non-emergency notification January 3, 2014 at 2207 EST (EN# 49690). The notification on January 3, 2014, reported under certain conditions, service air usage could result in the Auxiliary Building Secondary Containment Enclosure (ABSCE), in worst case conditions, exceeding the margin required to maintain the Auxiliary Building Gas Treatment System (ABGTS) operable and prevent both trains of ABGTS from performing its safety function(s). Subsequent engineering analysis concluded acceptable margin was available. Both trains of ABGTS would have remained operable and capable of performing its design function(s) at all times. The engineering analysis results are captured in the licensee's corrective action program. Based on the new analysis, the condition reported in EN #49690 did not result in an unanalyzed condition that significantly degraded plant safety. This event report is being retracted. The NRC Resident Inspector has been briefed on the analysis results and informed of this retraction. Notified R2DO (McCoy).

ENS 4947627 October 2013 22:29:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

At 1730 EDT on 10/27/2013, SQN (Sequoyah Nuclear) discovered that Unit 1 containment penetration X-108 had a maintenance flange installed with a service air connection attached. The service air connection was connected to a temporary air compressor supplying air to maintenance loads inside Unit 1 containment. Contrary to the requirements of the breaching permit, personnel were not stationed at the penetration to isolate the service air connection in the event of the air line rupturing inside Unit 1 containment or upon initiation of an auxiliary building isolation signal. Since the Unit 1 containment is open to the auxiliary building as part of outage activities, if the service air line had ruptured, the additional air into the Unit 1 containment could have exceeded the capacity of the Auxiliary Building Gas Treatment System (ABGTS) and potentially have impacted the ability of the ABGTS to perform its design safety function. This resulted in both trains of the ABGTS being declared inoperable requiring Unit 2 to enter the action of LCO 3.0.3. The service air line was isolated immediately and Unit 2 exited the action of LCO 3.0.3 at 1732 EDT. At the time of the event, Unit 1 was defueled and did not require ABGTS to be operable. Unit 1 subsequently entered Mode 6 at 1904 EDT on 10/27/2013 and is currently conducting refueling operations. Unit 2 remains in Mode 1, 100% power and stable. There were no actual operational impacts to either unit. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM BRUCE BUCH TO DANIEL MILLS AT 1428 EST ON 1/30/2014 * * *

Sequoyah Nuclear Plant, Units 1 and 2, are retracting the 8 hour non-emergency notification made on October 27, 2013 at 2229 EDT (EN# 49476). The notification on October 27, 2013, reported that if the service air line (associated with penetration X-108) had ruptured, the additional air into the Unit 1 containment could have exceeded the capacity of the Auxiliary Building Gas Treatment System (ABGTS) and potentially have impacted both trains of ABGTS from performing its safety function(s). Subsequent engineering analysis concluded acceptable margin was available. Both trains of ABGTS would have remained operable and capable of performing its design function(s) at all times. The engineering analysis results are captured in the licensee's corrective action program. Based on the new analysis, the condition reported in EN #49476 did not result in a potential uncontrolled radioactive release. This event report is being retracted. The NRC Resident Inspector has been briefed on the analysis results and informed of this retraction. Notified R2DO (McCoy).

ENS 4877824 February 2013 15:20:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn February 24, 2013 at 1205 (EST) with reactor power at 25% and the turbine offline, a manual reactor trip for Sequoyah Unit 2 was initiated due to loss of condenser vacuum indication causing closure of condenser steam dumps, opening of the Steam Generator Atmospheric Relief Valves, and lowering hotwell level resulting in imminent loss of hotwell pumps. The cause of the event was determined to be a faulty test connection on B Condenser vacuum pressure switch. During the event, steam pressure rose to the setpoint for the first Steam Generator code safety valve (1064 psig). (The safety valve opened, then reseated). Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. The reactor trip was uncomplicated. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (Normal Pressure and Temperature), with auxiliary feedwater supplying the steam generators and maintaining level at approximately 33% narrow range. Method of decay heat removal is via atmospheric reliefs to the atmosphere. Current RCS conditions: temperature (is) 547 degrees F and stable. Pressure (is) 2235 psig and stable. (There is) no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal and supplied from offsite power. (There is) no impact to Unit 1. Unit 1 is operating at 100% power / Mode 1. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart is 02/25/2013. (The licensee plans a press release.) The licensee notified the NRC Resident Inspector.
ENS 487256 February 2013 17:10:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn July 28, 2009, TVA identified, in the Corrective Action Program, the potential to overtop and fail earthen embankments at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams. This condition could have resulted in an increase in the probable maximum flood (PMF) level at Sequoyah Nuclear (SQN) Units 1 & 2. TVA initiated immediate actions to address the condition by conducting additional analyses and developing contingent actions. Additional actions were developed including the installation of modular flood barriers (which were) completed in December 2009. The barriers increase the effective height of the affected embankments preventing their overtopping and failure. The increase in PMF could have affected plant equipment including the emergency diesel generator system and the essential raw cooling water system. Additional details regarding the modular flood barriers and the results of TVA's subsequent hydrologic analyses for SQN were discussed in public meetings between TVA and the NRC staff on July 7, 2010 and May 31, 2012, and provided in TVA letters to the NRC dated August 10, 2012, October 30, 2012, and January 18, 2013. This report addresses a condition as described in 10 CFR 50.72 (b)(3)(ii)(B). Affected safety-related equipment is currently operable. The NRC Resident Inspector has been notified of this condition. See related event notifications from Watts Bar (EN #48723) and Browns Ferry (EN #48724).
ENS 4858412 December 2012 23:25:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1914 EST on 12/12/12, TVA determined that Sequoyah Unit 1 and 2 were at risk of flooding into the ERCW (Emergency Raw Cooling Water) Station Building during a design basis flood due to conduit penetrations not being filled with material required to make the building water tight. The lack of a barrier would allow flood waters to enter the ERCW building at a rate greater than the sump pumps can remove creating a condition that could result in the ERCW pumps being unavailable to perform their design function during a flood event above plant grade. This condition places both units in an unanalyzed condition that significantly degrades plant safety (10 CFR 50.72 (b)(3)(ii)(B)), and could prevent the fulfillment of the safety related function of ERCW needed to shutdown the reactor and maintain it in a safe shutdown condition (10CFR 50.72 (b)(3)(v)(A)). Compensatory actions are being established to be capable of removing or limiting water that could leak into the building during the event. The required safety related equipment is currently operable. There are no indications of conditions that might result in a flood in the near term. The NRC Resident Inspector has been notified of this condition.
ENS 483835 October 2012 11:19:00SequoyahNRC Region 2Westinghouse PWR 4-LoopA non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified.
ENS 4819816 August 2012 22:09:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1926 EDT on 8/16/2012, Unit 2 reactor automatically tripped on single loop loss of flow, following #4 RCP trip. Cause of RCP trip is under investigation. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (normal operating pressure/normal operating temperature), with auxiliary feedwater supplying the steam generator. Method of decay heat removal is via steam dumps to condenser. Current RCS conditions: Temp = 547 degrees F and stable, Pressure = 2235 psig and stable. No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from offsite power. No impact to Unit 1: Unit 1 is operating at 100% power / Mode 1. The licensee notified the NRC Resident Inspector.
ENS 480746 July 2012 13:04:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1012 EDT on July 6, 2012, Tennessee Valley Authority (TVA) notified Tennessee Emergency Management Agency (TEMA) due to a loss of twenty-one (21) offsite sirens. Siren outages are believed to be from power supply failures caused by high winds resulting in downed trees last night. Efforts are in place to restore offsite sirens to service. A loss of 21 of the 108 offsite sirens does not constitute a significant degradation in the Alert and Notification System. (A loss of 32 sirens constitutes a significant degradation.) Both Sequoyah units remain at 100 percent power. NRC Resident Inspector has been notified.
ENS 4795924 May 2012 17:31:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 5/24/2012 at 1541 EDT, a notification to the National Response Center was made after the discovery of a visible oil sheen on waters of the U.S. (the Tennessee River side of Sequoyah's intake forebay skimmer wall). The source of the oil appears to be a tipped or overflowing catchpan located in the Essential Raw Cooling Water (ERCW) pumping station. All catchpans in the pumping station have been emptied to eliminate them as immediate potential source of oil released to the environment. The following agencies have also been notified: EPA Region 4, and the Tennessee Emergency Management Agency (TEMA). The Tennessee Department of Environment and Conservation (TDEC) will be notified. Cleanup is in progress. Measures to prevent recurrence are being taken. The NRC Resident Inspector will be notified.
ENS 4792214 May 2012 13:45:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn 05/14/2012 at 1024 EDT, a notification to the National Response Center was made as a result of an oil sheen on waters of the U.S. (Sequoyah's Diffuser pond). The discharge from an onsite pond was isolated stopping what appears to be the source of all to the diffuser pond. The source of leakage is still under investigation. The following additional agencies have also been notified: EPA Region 4, Tennessee Department of Environment and Conservation (TDEC), and Tennessee Emergency Management Agency (TEMA). Oil is maintained onsite, (and) no oil has been discharged to the river. Cleanup is in progress. The licensee notified the NRC Resident Inspector. Notified DOE (Turner), EPA (Post), USDA (Beverly), HHS (Emerson), and FEMA (Casto).
ENS 4766012 February 2012 09:26:00SequoyahNRC Region 2Westinghouse PWR 4-Loop

At 0756 (EST) on 2/12/12, an electrical event occurred in the 161 kV switchyard that resulted in bus differential relay operating, and opening all PCBs (Protective Circuit Breakers) within 161 kV bus 2-2. Shortly after, a call was received from site security regarding an explosion in the switchyard. At 0820, a report was received that visible damage was present on an insulator between a PCB and a Motor Operated Disconnect (MOD) in the 161 kV switchyard, which is where the fault occurred. Follow up reports indicated an oil leak from a current transformer. Two other PCBs were identified to have visible damage. After the event, while opening MODS for PCBs that tripped open, the 'A' phase for one MOD did not open. No safety related equipment required to establish or maintain safe shutdown was affected. All the Diesel Generators remain in operable standby conditions. Unit 1 remained steady at 100% with no transients. Unit 2 had a 2B Condenser Circ Water pump trip off. Crew responded to the pump trip and Unit 2 remains at 100%. One offsite power source is inoperable and second offsite power source has been verified operable. Grid status is Yellow. There were no personnel injuries reported due to the fault and resulting explosion. The loss of the offsite power source resulted in the licensee entering a 72-hr. LCO. The licensee is investigating both the cause of loss of offsite power and the circulating water pump trip. The licensee has notified the State of Tennessee and the NRC Resident Inspector.

  • * * UPDATE FROM KEVIN WILKES TO HOWIE CROUCH AT 1056 EST ON 2/12/12 * * *

At 1019 EST, the licensee terminated their Notification of Unusual Event. The basis for termination was that the fault was cleared and the conditions for EAL entry no longer exists. The licensee has made state and local notifications and has notified the NRC Resident Inspector. Notified R2DO (Calle), IRD (Kozal), NRR (Evans), DHS (Beach), FEMA (Fuller), and NICC (Berger)

ENS 4753419 December 2011 17:25:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 1644 EST, on 12/19/11, TVA's Sequoyah Nuclear Plant made voluntary offsite notifications to the Tennessee Radiological Health Department Director and the Tennessee Department of Environmental Conservation - Senior Director Water Programs, to inform them of the following: On October 31, 2011, Sequoyah proactively installed two new groundwater monitoring wells in an area known to have contained previously reported releases of tritium in an effort to further characterize and validate the scope of the plume. These releases were reported in 2006 as part of NEI's (Nuclear Energy Institute) groundwater initiative. On December 16, 2011, elevated levels of tritium were identified in water samples taken from one new onsite monitoring well. The tritium levels were confirmed to be greater than 20,000 pCi/L which is the threshold for drinking water. No groundwater monitoring wells are used for drinking or irrigation purposes onsite. The highest level sampled was 22,760 pCi/L. Samples taken in the discharge channel located 30 yards from this groundwater monitoring well confirmed no detectable tritium. Refueling Water Storage Tank levels are being monitored and no active leak is in progress. Samples of adjacent wells have been taken and confirmed no unexpected changes in tritium levels of these wells. Additionally, Sequoyah has sampled at the station discharge to the Tennessee River and confirmed no detectable levels in any sample. In a conservative decision making process and in accordance with the groundwater protection initiative established by the nuclear industry, Sequoyah is voluntarily communicating sample results likely attributed to a previously reported tritium spill. The plant is continuing to review the sample results to confirm this is related to the historical tritium plume. The plant will take appropriate actions as outlined in the Groundwater Tritium Action Plan, which has been initiated to address this issue. Sequoyah Nuclear Plant has had an extensive groundwater monitoring program in place since 2008. The environmental sampling program consists of sixteen groundwater wells which are periodically sampled in accordance with industry standards. The NRC Resident Inspector has been notified. The following agencies have been, or will be updated: Hamilton County, State of Tennessee, Nuclear Energy Institute and American Nuclear Insurers. Tennessee Valley Authority plans a media notification for this issue.
ENS 4726113 September 2011 16:18:00SequoyahNRC Region 2Westinghouse PWR 4-LoopA non-licensed employee supervisor had a confirmed positive drug test during random testing. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector.
ENS 4725612 September 2011 13:54:00SequoyahNRC Region 2Westinghouse PWR 4-LoopOn August 25, 2011, it was discovered that a hazardous waste drum containing x-ray fixer waste was not shipped to the permitted Hazardous Waste Storage Facility within the timeframe required by the Tennessee Department of Environment and Conservation (TDEC) regulations. The drum was shipped to the permitted Hazardous Waste Storage Facility on September 01, 2011. The written notification to TDEC was signed on September 12, 2011. The licensee informed the NRC Resident Inspector.
ENS 472498 September 2011 10:34:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThis report is a 60-day telephone notification in lieu of a written licensee event report being made under 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1). The event was an invalid actuation of a Unit 2 Containment Ventilation Isolation (CVI). At the time of the event Unit 1 was at 100% power and Unit 2 was at 85% power. At 1520 EDT on 5/5/2011, an 'A' train CVI signal was inadvertently initiated during a surveillance test for containment purge air exhaust radiation monitor 2-RM-90-130. The inadvertent CVI signal was initiated due to incorrect connection of test equipment. The signal caused the 'A' train containment upper and lower compartment radiation monitor isolation valves to close. Unit 2 entered Technical Specification Limiting Condition for Operation (LCO) 3.3.3.1 Action 27 and LCO 3.4.6.1 Action b, due to the isolation of lower compartment radiation monitor 2-RM-90-106. The inadvertent CVI signal was also received by the containment vent system, but the containment vent system was not in service and no valves were actuated. The radiation monitoring (system 90) and the containment vent (system 30) systems received a complete 'A' train CVI signal. The 'A' train radiation monitoring isolation valves closed as designed. The containment vent system was not in service, and since the valves were already closed, no valves were actuated. Actual plant conditions did not exist that required a CVI signal. Therefore, this actuation was invalid. The delay in reporting this event was due to an initial interpretation that the event did not result in an actuation of the systems listed in paragraph 10CFR50.73(a)(2)(iv)(B), because only one system was in service which was affected by the actuation. Subsequent discussions noted that while only one system was in service, both systems received the CVI signal. The event is reported as a 60-day telephone notification in lieu of a written licensee event report being made under 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1). The date, when the final determination of the invalid system actuation was made, was not provided. The licensee notified the NRC Resident Inspector.
ENS 4716919 August 2011 01:10:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 2250 EDT on 8/18/2011, Unit 1 Reactor/Turbine automatically tripped on RCP (Reactor Coolant Pump) Busses UV (Under-Voltage) trip. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. No primary PORVs and/or Safety Valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F (degrees) and 2233 psig, with Auxiliary Feedwater supplying the Steam Generators. At the time of the trip, a 50G (instantaneous overcurrent ground) relay flag was found dropped on the '1A' 6.9 KV unit board. Subsequently, the '1A' 6.9 KV start bus was found to have transferred to its alternate supply, 'B' CSST (Common Station Service Transformer). 1A condenser circulating water pump motor trip out was also received in the MCR (Main Control Room). The method of decay heat removal is via steam dumps to the condenser with MSIVs open. The current temperature and pressure is stable. There is no indications of any primary/secondary leakage. All control rods inserted. The electrical alignment is normal with the exception of the above mentioned items, supplied from off-site power. There is no impact to Unit 2. Unit 2 is operating at 100% power/ Mode 1. The licensee notified the NRC Resident Inspector.
ENS 4708121 July 2011 00:53:00SequoyahNRC Region 2Westinghouse PWR 4-LoopAt 2129 EST on 7/20/2011, Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. Primary PORVs and/or safety valves lifted and reseated as indicated by tailpipe temperatures and PRT pressure. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with Auxiliary Feedwater supplying the steam generators. At the time of the trip, maintenance was in progress on Preferred Inverter #1. AOP-P.09 'Loss of 120VAC Preferred Power' was used to restore power to #1 Preferred board after the trip. Method of decay heat removal is via steam dumps to the condenser with MSIVs open. Current temperature and pressure: Temperature - 548 degrees Fahrenheit and stable, Pressure 2235 - psig and stable No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from off-site power. No impact on Unit 2. Unit 2 is operating at 100% power/Mode 1 The NRC Resident Inspector has been informed. The licensee notified the State of Tennessee.