|Start date||Reporting criterion||Title||Event description||System||LER|
|ENS 54487||22 January 2020 03:18:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition|
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 52187||17 August 2016 21:22:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition Due to Not Meeting Single Failure Criteria||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.||Emergency Gas Treatment System|
|ENS 51994||8 June 2016 19:26:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition That Could Potentially Degrade Plant Safety|
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.
|Reactor Coolant System|
Emergency Diesel Generator
|ENS 51935||17 May 2016 01:05:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition That Significantly Degrades Plant Safety|
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.
|Emergency Diesel Generator|
|ENS 51854||7 April 2016 19:07:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition - High Pressure Fire Suppression Isolated from Containment||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.||Main Turbine|
|ENS 51527||10 November 2015 20:02:00||10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat|
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
|Unanalyzed Condition Due to Debris Dropped Into Reactor Cavity Equipment Pit||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 50244||1 July 2014 02:46:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition That Could Have Resulted in an Increased Maximum Flood Level|
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 184.108.40.206 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.
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 49690||3 January 2014 20:00:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Abgts Potentially Inoperable for Both Units Under Worst-Case Conditions|
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.
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).
Auxiliary Building Gas Treatment System
|ENS 48584||13 December 2012 00:14:00||10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor|
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
|Risk of Possible Flooding to Ercw Building During Design Basis Flood||At 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 48725||28 July 2009 05:00:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition That Could Have Resulted in an Increased Maximum Flood Level||On 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).||Emergency Diesel Generator|
|ENS 44814||28 January 2009 19:25:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition Related to a Postulated Appendix R Fire Scenario||A postulated Appendix R fire scenario has been identified which could result in the plant being in an unanalyzed condition that may degrade plant safety. The unanalyzed condition involves a potential loss of power to the credited train of fire safe shutdown equipment for postulated fires in two areas of the Auxiliary Building. Safety-related loads on the 6.9KV Shutdown Boards (Centrifugal Charging Pumps, Safety Injection Pumps, Containment Spray Pumps, Motor-Driven Auxiliary Feedwater Pumps, and Residual Heat Removal Pumps) are equipped with local control switches to allow starting or stopping the pumps from the Auxiliary Building. The DC control circuit wiring for these local switches is routed in the same area as the power cables for pump motors. If the fire damages the control circuit cables, the control circuit fuses may fail, resulting in loss of trip capability for the associated 6.9KV breaker. If the breaker is already closed (due to a spurious signal or a previous valid start signal), then the breaker could be disabled in the closed position. If the fire damages the power cables resulting in a phase-to-phase fault with the load breaker trip capability disabled, then the Shutdown Board feeder breaker is designed to trip open on overcurrent to provide backup protection. This could result in the shutdown board being de-energized. In two areas in the Auxiliary Building (elevation 669 and 690 common areas), the above scenario could result in a condition which is outside the fire safe shutdown analysis due to loss of power to the credited train of fire safe shutdown equipment (e.g. Centrifugal Charging Pump, Essential Raw Cooling Water Pump, and motor-operated valves). As a result of this condition, Sequoyah has entered the Fire Protection Report Limiting Condition for Operation 3.7.12 for inoperable fire barriers. In accordance with this LCO action, the operability of fire detectors in the affected areas has been verified and an hourly fire watch has been established in the affected areas. This issue has been entered into the corrective action program. A permanent resolution is being evaluated. The NRC Resident Inspector has been notified.||Auxiliary Feedwater|
Residual Heat Removal