Semantic search
| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 56963 | 13 February 2024 17:10:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a synopsis of information provided by the licensee via email: On February 13, 2024, a non-licensed supervisor violated the station's FFD policy. The employee's access at Sequoyah Nuclear Plant has been terminated. The NRC resident inspector has been notified. |
| ENS 56888 | 13 December 2023 09:45:00 | Sequoyah | NRC Region 2 | The following information is a synopsis of information provided by the licensee via phone and email: On December 11, 2023, Tennessee Valley Authority Sequoyah Nuclear Plant completed an internal 10 CFR Part 21 evaluation concerning Siemens 6.9kV, 1200A vacuum circuit breakers, Model No. 7-HKR-50-1200-130. Three separate breakers were found with issues including loose wires terminated incorrectly and the mechanism-operated control switch clevis pin missing a cotter key. Additionally, the mastic insulating pads were found defective on all three lower primaries by way of separation. The affected breakers were never installed in a safety related application. The NRC Resident Inspector will be notified. A written notification will be provided within 30 days. The manufacturer, Siemens, was notified of the defects. The only plant known to be affected at the time of the report is the Sequoyah Nuclear Plant. | |
| ENS 56845 | 9 November 2023 15:55:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via email: A controlled substance was found in the protected area. The NRC Resident Inspector has been notified. |
| ENS 56592 | 27 June 2023 11:52:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0831 (EDT) on June 27, 2023, Sequoyah Nuclear Plant reported an oil discharge into the plant intake located on the Tennessee River to the (Department of Transportation) National Response Center (report number 1371356). The source of oil was from a broken hydraulic hose from equipment in use on the intake. This oil spill is minor and did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56322 | 25 January 2023 13:22:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information is a synopsis of information provided by the licensee via fax and phone: On May 23, 2022, Framatome informed Tennessee Valley Authority (TVA) of a deviation of breakers purchased under contract. On January 23, 2023, TVA determined that a defect of the basic component could create a substantial safety hazard. Framatome Inc. identified a deviation in the Siemens medium voltage vacuum circuit breaker where a failure to electrically charge or electrically close could occur. Framatome Inc. identified this as a departure from the technical requirements included in the procurement document. It is noted that the ability to electrically trip the circuit breaker would not be affected by the condition. TVA was notified by Framatome under 10 CFR 21.21(b) to evaluate the application of the breaker for a substantial safety hazard. The TVA evaluation identified these breakers as intended for use in safety related Class 1E applications where a loss of the closure function would impact mitigation of design basis accidents and transients. During the Framatome dedication testing/inspection of Siemens medium voltage vacuum breakers, a hi-pot test failure on one circuit breaker was encountered. Troubleshooting and inspection found damage to charging motor wiring. It was determined that the cause of the damage was due to the manner in which control wiring was routed and connected to the internal bracket in close proximity to a bracket edge. This edge caused damage to wiring after significant number of cycles were applied to the breaker prior to dedication testing. TVA received nine medium voltage vacuum circuit breakers at an offsite warehouse facility. While located at that facility, TVA, with assistance from Framatome, examined the affected breakers for the wire routing condition. The wiring harnesses of certain breakers were corrected. Framatome is to examine medium voltage vacuum circuit breakers that may be purchased under this contract for the wiring condition and correct as necessary before delivery. The NRC Senior Resident Inspector has been notified. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i).
The following information is a synopsis of information provided by the licensee via phone: The Sequoyah site licensing manager requested via phone call to the HOO that the model number for the basic component with the defect be listed in the Part 21 event narrative in addition to the official Part 21 report. The component discussed is a Siemens 6.9kV, 1200A, 125VDC Vacuum Circuit Breaker, Model No.: 7-HKR-50-1200-130. Notified R2DO (Miller) and the Part 21 Reactors Group (Email). |
| ENS 55867 | 29 April 2022 07:04:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax: On 4/28/2022, at 2338 EDT, Sequoyah received an unexpected alarm for seismological recording initiated. At 2341 EDT, unexpected alarm 1/2 Safe Shutdown Earthquake response spectra exceeded was received. The National Earthquake Information Center was contacted to confirm there was no seismic activity, and this was also confirmed on the U.S. Geological Survey website. The alarms were determined to be invalid, and they occurred due to a failure in the seismic monitoring system. This failure results in loss of ability to assess the Emergency Action Level for Initiating Condition HU2 `Seismic event greater than Operating Basis Earthquake (OBE) levels' per procedure EPIP-1, `Emergency Plan Classification Matrix.' If an actual seismic event had occurred, HU2 could not be assessed. However, compensatory measures have been implemented and include assessing OBE criteria based on alternative criteria contained in procedure AOP-N.05, `Earthquake,' which provides conservative guidance when seismic instruments are unavailable. This is an eight-hour, non-emergency notification for an event resulting in a major loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified." The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The faulty detector was removed from service, so the remaining detector provides conservative detection as the only source to make-up the logic for a seismological alarm. |
| ENS 55866 | 29 April 2022 00:19:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via telephone: On 04/28/22, at 2355 EDT, with both Sequoyah Unit 1 and 2 in Mode-1, 100 percent, a Notice of Unusual Event was declared due to receiving two seismic alarms and security feeling ground movement. Additionally, security in a tower heard an explosion. Both units remain in Mode-1, 100 percent and they are investigating the validity of the seismic alarms before proceeding with the Abnormal Operating Procedure required shutdown. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee will notify the NRC Resident Inspector. The state of Tennessee and the Tennessee Valley Authority were notified. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS NRCC THD Desk(email), and DHS Nuclear SSA (email).
The following is a summary of information provided by the licensee via telephone: On 4/29/22, at 0406 EDT, Sequoyah Unit 1 and Unit 2 terminated the Notice of Unusual Event. The Civil Engineers determined that the alarms were due to a failed seismic indicator channel. Through interviews, only one security officer felt ground movement for a couple of seconds and heard a faint rumbling sound. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee will notify the NRC Resident Inspector. The state of Tennessee and the Tennessee Valley Authority were notified. Notified R2DO (Miller), NRR EO (Miller), and IR MOC (Gott) via email. Additionally, notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS NRCC THD Desk(email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via email: SQN (Sequoyah Nuclear Plant) is retracting the previous NOUE (Notice of Unusual Event) declaration made on 4/28/22 at 2355 (EDT) based on Emergency Action Level HU2 for a seismic event greater than Operating Basis Earthquake levels. Following the declaration of the NOUE, the station reviewed all available indications and determined that a seismic event had not occurred. The instrumentation failure was documented under Event Notification #55867. Notified R2DO (Miller), and IR MOC (Gott), NRR EO (Miller) via email. |
| ENS 55742 | 16 February 2022 17:01:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax or email: At 1128 EST on 2/16/2022, the SQN (Sequoyah Nuclear) Shift Manager was notified that TVA (Tennessee Valley Authority) attempted to notify Tennessee Emergency Management Agency (TEMA) regarding routine siren testing at 0750. TVA was unable to reach TEMA via telephone land line or the Emergency Communication and Notification System (ECNS). TEMA Watch Point staff were located at their back-up facility. TVA subsequently notified TEMA via cell phone that there were communication issues with the primary and backup notification methods. It was determined that the TEMA back-up facility was not able to receive incoming calls. At 0820, TEMA positioned personnel at their primary facility in order to respond to notifications. This restored primary and backup means of notifying the state because the primary facility was not affected by the communication issues. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Offsite Communications Capability because it affected TVA's ability to notify the State of TN. The licensee has notified the NRC Resident Inspector. |
| ENS 55421 | 20 August 2021 16:00:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 0905 EDT, it was discovered both trains of Auxiliary Building Gas Treatment System (ABGTS) were simultaneously INOPERABLE due to the auxiliary building secondary containment enclosure (ABSCE) being 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. ABSCE and ABGTS were returned to operable.
This is a retraction of the 8-hour Immediate notification (EN55421) made to the NRC by Sequoyah Nuclear Plant on August 20, 2021. Sequoyah is retracting this event notification based on the following: Regulatory Guidance in NUREG-1022, Revision 3, 'Event Reporting Guidelines 10 CFR 50.72 and 50.73', Sections 2.8 'Retraction and Cancellation of Event Reporting', and 4.2.3 'ENS Notification Retraction'. On August 20, 2021 personnel found door A-118 open. This door is part of the ABSCE. During the initial investigation, it was found that other personnel had the door open using Precaution A of 0-TI-SXX-000-016.0 which allows material access through ABSCE doors if the door is closed within three minutes. It was found that A-118 door had been open for greater than three minutes. With this door open the ABSCE was beyond its capability for ABGTS fan to maintain the required pressure during an Aux. Building Isolation. Thus, the site declared the ABSCE and both Trains of ABGTS inoperable per LCO 3.7.12 Conditions A, B and E. With the ABSCE being a single train system, this caused a condition that "could have prevented the fulfillment of the safety function" which requires an Immediate Notification to the NRC within eight hours under 10 CFR 50.72 (b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). This Immediate Notification was reported on August 20, 2021 at 1600 EDT. It was later determined that at 'Time of Discovery', although Door A-118 was open, it was not obstructed, the door was open by normal means, was capable of being closed and was now attended. The time requirement per 0-TI-SXX-000-016.0 for closure of an open ABSCE door is within three minutes of notification. Although the individual found holding the door was unaware of the requirement of 0-TI-SXX-000-016.0 to close the door, communications were established and the Main Control Room (MCR), upon discovery of the 'Open Door', could have directed closure starting at the Time of Discovery if required. Since the MCR was aware the door was open, had communications established with personnel at the door, the door was capable of closure and not restricted, the three minute closure requirement of 0-TI-SXX-000-016.0 was met. Subsequently, the door was closed within approximately two minutes of notification to close. The closure of the door with these procedural measures met confirmed the integrity of the ABSCE and therefore Operability of ABGTS. Based on the above critical thinking, entry into LCO 3.7.12 Condition A, B, and E was retracted on August 22, 2021 at 2044 EDT. With the LCO conditions retracted and the above determination that at the Time of Discovery safety function was maintained, the Immediate Notification per 10 CFR 50.72 (b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D) was not required. The issue of Past Operability remains for instances in time that the door did not have appropriate compensatory measures in place. Any further notification required for this event will be submitted as a Licensee Event Report. Notified R2DO (Miller) |
| ENS 55379 | 25 July 2021 16:00:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 1238 EDT on July 25, 2021, the Unit 2 Ice Bed became INOPERABLE due to SR (Surveillance Requirement) 3.6.12.1 exceeding its surveillance interval. LCO (Limiting Condition for Operation) 3.6.12 was declared not met as required by SR 3.0.1. SR 3.6.12.1 to verify maximum ice bed temperature is less than or equal to 27 degrees F could not be completed due to a failed temperature recorder. The results of the backup method of temperature verification were verified satisfactory at 1258 EDT and the LCO condition was then exited. The ice bed is a single train system which functions to control radiation release and mitigate the consequences of an accident by scrubbing radioactive iodine and providing a heat sink to limit containment pressure within design limits, therefore the requirements of 10 CFR 50.72 (b) (3) (v) (C) and (D) were met. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55272 | 24 May 2021 12:01:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | Unit 2 is not impacted and remains stable in Mode 1 at 100 percent power. 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. No relief valves opened. All Rods fully inserted. Decay heat is being removed by Auxiliary Feedwater via the steam dumps. The plant is in a normal post-trip electrical line-up. |
| ENS 55130 | 9 March 2021 11:58:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A). The event was an invalid actuation of the Unit 1 Containment Ventilation Isolation (CVI) system. On January 11, 2021 at 1152 Eastern Standard Time (EST) with Unit 1 at 100% power, Train 'A' of the CVI System actuated due to an invalid high radiation signal from 1-RM-90-130, Containment Purge Air Exhaust Monitor. The cause of the signal was determined to be a failed sample pump associated with the radiation monitor. 1-RM-90-130 was in service at the time of the invalid signal. The Train 'A' Containment Ventilation Isolation signal was a full actuation of that train and the system functioned as designed. Prior to and following the invalid high radiation alarms, all radiation monitors except 1-RM-90-130 were stable at their normal values; therefore, the CVI was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. The failed pump was replaced and returned to service. This event was entered into the corrective action program as CR 1663398. The NRC Resident Inspector was notified. |
| ENS 54977 | 1 November 2020 11:04:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 0556 EST on 11/01/2020, Sequoyah received unexpected alarms for seismological recording initiated and (Units) 1/2 Safe Shutdown Earthquake response spectra exceeded. No seismic event was felt on site, the National Earthquake Information Center was contacted to confirm there was no seismic activity, and this was also confirmed on the U.S. Geological Survey website. The alarms were determined to be invalid, and they occurred due to a failure in the seismic monitoring system. This failure results in loss of ability to assess the Emergency Action Level for Initiating Condition HU2 'Seismic event greater than Operating Basis Earthquake (OBE) levels' per procedure EPIP-1. If an actual seismic event occurred, HU2 could not be assessed. However, compensatory measures have been implemented and include assessing OBE criteria based on alternative criteria contained in procedure AOP-N.05 'Earthquake' which provides conservative guidance when seismic instruments are unavailable. This is an eight hour, non-emergency notification for an event resulting in a major loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 54918 | 28 September 2020 14:39:00 | Sequoyah | NRC Region 2 | On 9/28/20 at 1143 EDT, a notification to the National Response Center was made after discovery of a visible oil sheen on the waters of the U.S. (Sequoyah's side of the intake forebay skimmer wall at the Essential Raw Cooling Water (ERCW) building). The source was an oil bucket that overflowed with rain at the ERCW pumping station. Efforts are in progress to eliminate all other potential sources of oil at the station that could be released to the environment. Estimate of volume spilled is less than one quart. The following agencies have also been notified: - EPA Region 4 - Tennessee Emergency Management Agency (TEMA) - Tennessee Department of Environment and Conservation (TDEC). Cleanup is in progress. Measures to prevent recurrence are being taken. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. | |
| ENS 54800 | 24 July 2020 09:00:00 | Sequoyah | NRC Region 2 | At 0105 (EDT) on 7/24/20 it was discovered Unit 2 Ice Bed was INOPERABLE. Therefore, since this is a single train system the requirements of 50.72 (b)(3)(v)(C) and (D) have been met. This condition is being reported as an 8-hour non-emergency NRC Notification. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This condition put the unit in a 48-hour LCO. The old chillers were put into service to bring the temperature of the ICE bed down. At 0833 EDT, the technical specification limit was no longer exceeded and the unit exited the LCO. | |
| ENS 54708 | 13 May 2020 05:38:00 | Sequoyah | NRC Region 2 | At 0208 EDT on 05/13/2020, Sequoyah Unit 1 was at 100% power when an automatic reactor trip signal was received concurrent with a low steam line pressure safety injection signal. The low steam line pressure safety injection signal was actuated from the steam pressure rate of decrease feature. Main steam isolation valves (MSIVs) automatically closed as designed and steam generator pressures stabilized following the isolation. All other safety-related equipment operated as designed, with the exception of 1-FCV-61-122 Glycol inboard containment isolation valve which failed to automatically isolate on a Phase A containment isolation signal. The corresponding outboard containment isolation valve, 1-FCV-61-110, automatically isolated as designed which isolated penetration X-114. Safety injection was terminated at 0221 EDT 5/13/20, and Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure with auxiliary feedwater supplying the steam generators and decay heat removal via steam generator atmospheric relief valves. There is no indication of any primary to secondary leakage. The electrical alignment is normal with shutdown power supplied from off-site power. There is no current operational impact to Unit 2. There is no impact on public health or safety. Post safety injection actuation investigation is in progress. The NRC Resident Inspector has been notified. | |
| ENS 54487 | 22 January 2020 04:34:00 | Sequoyah | NRC Region 2 | EN Revision Imported Date : 2/21/2020 CONTAINMENT RELIEF VALVES INOPERABLE At 22:18 (EST) on 1/21/20, it was discovered that all Unit 1 containment vacuum relief isolation valves were closed and all vacuum relief lines were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The isolation valves were opened and the vacuum relief valves were restored to operable. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
At 1549 (EST), February 20, 2020, a completed engineering evaluation of the condition initially reported on January 22, 2020 determined that the inoperability of the Sequoyah Unit 1 Containment Vacuum Relief System affected the ability to protect containment against an external pressure event. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The condition was resolved when isolation valves were opened on January 21, 2020 and the vacuum relief lines were restored to an operable status. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B), "an unanalyzed condition that significantly degrades plant safety. Subsequent to the initial notification, continued evaluation of the reported condition has concluded that the isolation of the containment vacuum relief function did not prevent the fulfillment of a safety function that is needed to control the release of radioactive material; nor mitigate the consequences of an accident therefore this event is not reportable under 10 CFR 50.72(b)(3)(v), "Event or Condition that could have prevented fulfillment of a safety function. The NRC Resident has been notified. Notified R2DO (Musser) | |
| ENS 54446 | 16 December 2019 09:12:00 | Sequoyah | NRC Region 2 | At 0358 EST, on 12/16/2019, with Unit 1 in Mode 1 at 100 (percent) power and Unit 2 in Mode 1 at 47 (percent) power, a valid actuation of the Emergency Diesel Generators (EDG) occurred. The reason for the emergency diesel generator auto start was that the normal feeder breaker from the 1C 6.9KV Unit Board to the 1B-B 6.9KV Shutdown Board (SDBD) tripped due to the breaker's 51G relay actuating causing an under-voltage signal on the 1B-B 6.9KV Shutdown Board. All 4 Emergency Diesel Generators automatically started as designed when the 6.9KV Shutdown Board under-voltage signal was received. The 1B-B 6.9KV Shutdown Board was automatically energized from the 1B-B 6.9KV Diesel Generator. All required 6.9KV loads were sequenced back on to the 1B-B 6.9KV Shutdown Board as designed after the board was energized from its emergency diesel generator. The remainder of the electrical system is in normal alignment. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Emergency Diesel Generators. There was no impact to the health and safety of the public or plant personnel. The NRC Senior Resident has been notified. | |
| ENS 54438 | 12 December 2019 08:14:00 | Sequoyah | NRC Region 2 | EN Revision Imported Date : 3/4/2020 MANUAL REACTOR TRIP DUE TO A LOSS OF HEATER DRAIN TANK PUMP FLOW At 0432 EST, on 12/12/19, Sequoyah Unit 2 experienced a manual reactor trip. The trip was initiated due to a loss all number 3 Feedwater Heater Drain Tank pump flow; plant procedures directed a manual reactor trip if power is greater than 80 percent. The Auxiliary Feedwater System (AFW) automatically actuated as required when the expected post trip feedwater isolation actuation actuated. Reactor Coolant System (RCS) temperature is being maintained by the steam dump system with all 4 Reactor Coolant Pumps (RCPs) in service. All control and shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the transient. Unit 2 is currently stable at normal operating temperature and normal operating pressure in Mode 3. The electrical system is in a normal alignment. There was no impact on U1. There was no impact to the health and safety of the public or plant personnel. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10CFR50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification accordance with 10CFR50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. The NRC Resident Inspector was notified.
The following update to the EN submitted on 12/12/19 is being made to provide clarification on reporting criteria originally described in paragraph five of EN 54438: This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector was notified. Notifed R2DO (Davis). | |
| ENS 54380 | 12 November 2019 08:28:00 | Sequoyah | NRC Region 2 | On November 12, 2019, the Central Emergency Control Center (CECC) was removed from service for a planned facility upgrade project. The CECC is a common Emergency Operations Facility (EOF) for the TVA Nuclear sites (Browns Ferry / Sequoyah / Watts Bar). The duration of the upgrade project is approximately 75 days. If an emergency is declared requiring CECC activation during this period, an alternate CECC will be used. During this period, the alternate CECC will be staffed and activated using existing emergency procedures. This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the CECC will be unavailable for more than 72 hours. The Emergency Response Organization has been notified that the CECC will be unavailable during the upgrade project and to report to the alternate CECC in the event of an emergency. There is no impact on the health and safety of the public or plant employees. The NRC Resident Inspector has been notified.
The CECC facility upgrade project is sufficiently complete such that the CECC was returned to a functional status at 1400 EDT on January 31, 2020. The NRC Resident Inspector has been notified. Notified R2DO (Baptist). | |
| ENS 54276 | 13 September 2019 11:57:00 | Sequoyah | NRC 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.
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 54242 | 27 August 2019 02:34:00 | Sequoyah | NRC Region 2 | At 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 53999 | 14 April 2019 06:44:00 | Sequoyah | NRC 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.
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 53754 | 26 November 2018 08:31:00 | Sequoyah | NRC 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).
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).
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).
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).
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 53751 | 24 November 2018 21:27:00 | Sequoyah | NRC Region 2 | At 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 53680 | 21 October 2018 19:58:00 | Sequoyah | NRC Region 2 | This 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 53270 | 19 March 2018 02:27:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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 52769 | 24 May 2017 04:10:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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 52597 | 7 March 2017 14:39:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 52469 | 30 December 2016 16:37:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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 52426 | 13 December 2016 14:40:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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 52187 | 17 August 2016 17:46:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 52172 | 11 August 2016 16:35:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51994 | 8 June 2016 17:10:00 | Sequoyah | NRC Region 2 | Westinghouse 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 51935 | 16 May 2016 21:17:00 | Sequoyah | NRC Region 2 | Westinghouse 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 51900 | 3 May 2016 15:13:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51854 | 7 April 2016 18:01:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | This 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 51720 | 9 February 2016 17:25:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51654 | 14 January 2016 16:07:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | A 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 51559 | 23 November 2015 11:52:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51527 | 10 November 2015 20:13:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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 51451 | 5 October 2015 10:00:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | This 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 51392 | 14 September 2015 08:12:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51265 | 27 July 2015 13:44:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51259 | 24 July 2015 17:04:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 51117 | 4 June 2015 11:18:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | This 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 50878 | 11 March 2015 09:30:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At (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 50856 | 2 March 2015 10:22:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | At 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 50853 | 27 February 2015 17:04:00 | Sequoyah | NRC Region 2 | Westinghouse 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 50757 | 23 January 2015 15:09:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | On 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. |