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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56910 | 28 December 2023 16:29:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition | The following information was provided by the licensee via email: Plant alignment caused an unanalyzed condition regarding unit 1 and unit 2 Appendix R procedures. (Watts Bar Nuclear) (WBN) unit 1 and unit 2 Appendix R procedures require manual operator action times including (volume control tank) (VCT) isolation. They are calculated with an assumed hydrogen cover gas constant at 20 psig. This is to preclude hydrogen ingestion into the charging pumps with an operator action time of 70 minutes. Due to recent lower hydrogen concentration in the (reactor coolant system) (RCS), (unit 2) VCT hydrogen regulator set point was increased to 28 psig. This increased pressure set point invalidated the initial assumptions made in the Appendix R calculations for manual operator action times. WBN unit 1 VCT hydrogen regulator was also verified high out of band at 22 psig. WBN has restored unit 1 and unit 2 VCT hydrogen regulators to the required specification. The NRC Resident Inspector has been notified of this condition. | ||
| ENS 56541 | 25 May 2023 17:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition of Emergency Diesel Generator | The following information was provided by the licensee via phone and email: At 1345 EDT on May 25, 2023, it was determined that a fire barrier for area 737-A1B was not installed, and would render the 2A Emergency Diesel Generator (EDG) not operable in the event of a fire on the Unit 2 side of elevation 737 in the Auxiliary Building. The 2A EDG is the credited power source for fire safe shutdown for a fire located in this area. Without the credited source of power, this places WBN U2 (Watts Bar Nuclear Unit 2) in an unanalyzed condition. A fire watch has been established in the area until the issue is resolved. Therefore, this event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 53801 | 20 December 2018 05:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Containment Air Return Fan System Inoperable | At 1642 Eastern Standard Time (EST) on December 20, 2018, it was determined that both trains of Containment Air Return Fan (CARF) were simultaneously INOPERABLE from 0817 (EST) to 1129 (EST) on November 20, 2018. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. | ||
| ENS 53349 | 19 April 2018 23:44:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Related to Emergency Core Cooling Gas Accumulation Acceptance Criteria | On April 19, 2018 at 1944 EDT, Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows current acceptance criteria for gas accumulation in the WBN Unit 1 and Unit 2 Safety Injection System (SIS) and Residual Heat Removal System (RHRS) discharge piping may be non-conservative. The surveillances that check void values and allow venting of the systems are to be performed utilizing conservative criteria at more frequent intervals to ensure gas void volumes remain under acceptable limits. Additional analysis is being performed to determine final actions. The NRC Resident Inspector has been notified. | Residual Heat Removal | |
| ENS 52850 | 12 July 2017 16:38:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Related to Gdc-5 (General Design Criterion-5) for Dual Unit Operation | On July 12, 2017, at 1238 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows adequate Essential Raw Cooling Water (ERCW) flow may not be in place during dual unit limiting design basis conditions of one unit in Hot Shutdown on Residual Heat Removal (RHR) cooling when the other unit experiences a Loss of Coolant Accident (LOCA). Based on preliminary analysis, during a Unit 1 LOCA, Unit 1 receives adequate flow when following existing procedural guidance; however, Unit 2 may not receive adequate flow to meet cool-down requirements with design basis maximum temperatures. During a Unit 2 LOCA, however, current procedural guidance is not adequate to ensure the proper system alignment to establish correct ERCW Component Cooling Water (CCS) Heat Exchanger A and B flow rates for either unit's cool down requirements. Unit 2 has been shutdown for an extended period of time such that the flow delivered by ERCW is adequate to serve both Unit 1 in a LOCA and Unit 2 in less than Mode 3. The NRC Resident Inspector has been notified. | Residual Heat Removal | |
| ENS 51995 | 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 | 05000390/LER-2016-010 |
| ENS 50960 | 6 April 2015 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Volume Control Tank Isolation Time Analysis Error | On April 6, 2015, Watts Bar determined that the current Unit 1 Fire Protection Report (FPR) analysis for 10 CFR 50, Appendix R contained a non-conservative time for isolation of Volume Control Tank (VCT) following a postulated fire in room 737.0-A1A. Due to multiple fire-induced failures, Component Cooling System (CCS) cooling of letdown flow to the VCT will be lost in conjunction with increased Reactor Coolant System (RCS) injection flow through Reactor Coolant Pump (RCP) seals and Boron Injection Tank (BIT) flowpaths. As a result, VCT temperature and pressure would increase, which could cause RCP seal damage and loss of RCS inventory, and net positive suction head (NPSH) to the Centrifugal Charging Pumps (CCPs) could be lost. In this postulated fire scenario, Thermal Barrier Cooling is also not available due to fire-induced failures. The current FPR analysis assumes a time of 70 minutes for closure of VCT outlet isolation valves. Preliminary analysis performed by TVA showed that VCT outlet isolation is required in approximately 4 minutes. As a result, Watts Bar Unit 1 is in an unanalyzed condition. This condition significantly degrades plant safety because operator action to isolate the VCT in the event of a postulated fire in room 737.0-A1A would not have been performed in time to prevent RCP seal damage and loss of RCS inventory and eventual loss of NPSH to the CCPs. Watts Bar is utilizing previously established fire watches in the affected fire areas as a compensatory measure. This issue is being reported under 10 CFR 50.72(b)(3)(ii)(B), 'unanalyzed condition that significantly degrades plant safety.' The Watts Bar NRC Resident Inspector has been notified. | Reactor Coolant System | |
| ENS 50887 | 13 March 2015 15:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition | On March 13, 2015, Watts Bar determined that information received from Westinghouse represented an unanalyzed condition. The information challenged the current Unit 1 Fire Protection Report (FPR) analysis for 10 CFR 50 Appendix R. The current FPR analysis used a duration of 120 seconds before a low pressurizer pressure safety injection (SI) signal would be received due to a postulated failed open pressurizer power operated relief valve (PORV) from an Appendix R fire outside of the control building. During discussions with Westinghouse, Watts Bar determined that the 120 second value was non-conservative and not intended for this application. Recent information from Westinghouse indicates that a Sl signal would be received in approximately 34 seconds. This information places Watts Bar Unit 1 in an unanalyzed condition because the original analysis assumed that a Sl signal would not be received from fire-induced failures in the associated fire areas. While the existing Watts Bar Unit 1 Appendix R procedures direct operators to isolate the pressurizer PORV, the procedures did not have steps to mitigate and terminate the Sl. Without mitigating actions, the Sl would likely challenge the reactor coolant system boundary due to water relief through the remaining pressurizer PORV and/or safety valves. As a compensatory measure, Watts Bar has established administrative equipment controls to preclude the consequences of a spuriously opening pressurizer PORV. In addition, a compensatory measure in the form of fire watches has been established in areas where a fire could result in a spurious opening of a pressurizer PORV. This issue is being reported under 10 CFR 50.72(b)(3)(ii)(B), 'unanalyzed condition that significantly degrades plant safety.' The Watts Bar Resident Inspector has been notified. | Reactor Coolant System | |
| ENS 50596 | 5 November 2014 18:10:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Regarding Appendix R Procedures | On November 5, 2014, during a review of recommended procedure changes, TVA determined procedures for Appendix R fires did not include all the required operator manual actions to address inadvertent opening of the pressurizer spray valves. Failure to secure the reactor coolant pumps or auxiliary spray would invalidate Appendix R assumptions for not overfilling the pressurizer during an Appendix R event. Failure to take all the required actions would place WBN (Watts Bar Nuclear) Unit 1 in an unanalyzed condition. The three affected procedures have been revised on 11/5/2014, to correct the condition. The NRC Resident Inspector has been notified of this condition. | ||
| ENS 50540 | 15 October 2014 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Appendix R Unanalyzed Condition Affecting the Turbine Driven Aux Feedwater Pump | WBN U1 (Watts Bar Nuclear Unit-1) Appendix R (APP R) procedures do not include all the required operator manual actions to ensure manual control of the Unit 1 Turbine Driven Auxiliary Feed (TDAFW) Pump during a fire in room 713-A1A or 737-A1A. The current APP R procedure includes actions to transfer the Unit 1 TDAFW Pump to local control, but does not account for the potential loss of 125 VDC or 120 VAC to the Unit 1 TDAFW pump controls. Without procedural guidance to transfer the 120 VAC and 125 VDC supplies, the operation of the Unit 1 TDAFW pump may not be completed within the required time. Failure to take all the required actions to control the Unit 1 TDAFW Pump places WBN Unit 1 in an unanalyzed condition. WBN has instituted a fire impairment for the affected rooms and established an hourly roving fire watch. The NRC Resident Inspector has been notified of this condition. The affected components are the trip throttle valve and the local control panel. Procedures should be in place by 10/16/2014 to rectify this situation. | Feedwater | |
| ENS 50245 | 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 Watts Bar Nuclear Plant (WBN) a deviation from the current hydrologic analysis. The flooding analysis in Section 2.4.3 of the WBN 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 WBN-UFSAR-12-01 (Application to Revise Watts Bar Nuclear Plant Unit 1 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 WBN Plant Site. The current licensing basis PMF level for WBN is 734.9 feet as stated in Section 2.4.3.5 of the WBN UFSAR. In addition, it should be noted that by letter dated July 19, 2012, TVA proposed a revised PMF level of 739.2 feet. Introducing non failure of the Watts Bar West Saddle Dike indicated a potential increase of approximately 1.7 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 WBN 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, TVA reported (Event 50245) 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 Watts Bar Nuclear Plant (WBN) by 1.7 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 WBN. 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 49818 | 11 February 2014 23:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Nonconservative Operator Manual Actions Identified in Appendix R Analysis | Preliminary Westinghouse transient analysis calculations of Watts Bar Nuclear (WBN) Unit 1 fire protection features revealed that there is less time than previously credited to perform certain operator manual actions (OMA) to prevent Pressurizer (PZR) overfill during an Appendix R fire. Specifically, an assumed Appendix R fire in rooms 713.0-A28, 737.0-A1A, 757.0-A2, 757.0-A5, 757.0-A9, 772.0-A1, 772.0-A2, or 772.0-A5 could result in spurious operation of multiple components in the normal and emergency charging flow paths. Westinghouse's analysis indicates the required time to isolate the normal charging path is approximately two minutes, securing the second charging pump is approximately four minutes and isolating the emergency charging path is approximately 12.5 minutes. Based on this preliminary analysis, WBN procedures are non-conservative since they require these actions to be completed in 18 minutes. TVA has verified that potentially impacted Appendix R equipment remains functional; however, a compensatory fire watch has been established for the above listed areas. This immediate action will ensure that the approved Appendix R Fire Safe Shutdown Plan can be achieved. The licensee has notified the NRC Resident Inspector. | 05000390/LER-2014-002 | |
| ENS 49541 | 14 November 2013 21:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Postulated Fire Induced Failure of Centrifugal Charging Pumps | During analysis of Watts Bar Nuclear (WBN) Unit 2 fire protection features, it was revealed that a potential fire induced failure of centrifugal charging pumps could occur in Unit 1. Specifically, a potential fire induced failure of both Unit 1 Chemical and Volume Control System centrifugal charging pumps (CCPs) (1-PMP-62-108-A and 1-PMP-62-104-B) could occur due a fire in either auxiliary building room 737.0-A1 (general area for elevation 737.0) or 757.0-A2 (6.9 kV and Shutdown Board Room A). It is postulated that a fire in these rooms could cause a spurious closure of the CCP suction valve (1-LCV-62-133-B) from the volume control tank (VCT) (1-TANK-62-129) and could disable the control circuit which opens the flow from the refueling water storage tank (RWST) suction valve (1-LCV-62-135-A). The fire safe shutdown analysis (Fire Protection Report, Part VI) currently addresses this occurrence via the performance of a prompt main control room operator action to open the RWST suction path. However, this procedurally directed action may require several minutes to complete and due to the potentially short duration (possibly as short as a few seconds) for CCP survivability without suction flow, the action has now been determined to be unacceptable. As a result, the loss of charging flow could result in a loss of injection to the reactor coolant pump (RCP) seals which could subsequently lead to a RCP seal failure and a small break loss of coolant event. WBN engineering is continuing to validate whether the CCP minimum flow recirculation would protect the pumps with both suction paths (VCT and RWST) isolated and with the reactor at normal operating pressure. WBN has established compensatory measures to ensure that a fire in affected rooms will not cause a spurious closure of the CCP suctions valves. The licensee has notified the NRC Resident Inspector. | ||
| ENS 49389 | 27 September 2013 16:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Calculation Error Resulted in Unanalyzed Condition for Appendix R Event | On July 18, 2012, TVA identified that a calculation error resulted in the inability to establish Essential Raw Cooling Water (ERCW) supply to the Component Cooling System (CCS) heat exchanger within the required time to ensure cooling is available to Reactor Coolant System (RCS) seal injection water during an Appendix R event. TVA promptly established fire watches as a compensatory measure to prevent initiation of fires in the areas of concern. Subsequently, abnormal operating instructions were revised to take action to ensure that adequate time is available to establish ERCW cooling to CCS, and ensure cooling to (Reactor Coolant Pump) RCP seal injection water. The calculation error was identified on July 18, 2012, however, it was not recognized at the time that an unanalyzed condition existed that was reportable under 10 CFR 50.72(b)(3)(ii)(B). This report documents that the condition that existed until fire watches were established is reportable under 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified of this condition.
Watts Bar Nuclear Plant Unit 1 is retracting this 8 hour non-emergency notification made on September 27, 2013, at 1335 EDT (EN #49389). The notification on September 27, 2013, reported a calculation error which resulted in the inability to establish Essential Raw Cooling Water (ERCW) to supply the Component Cooing System (CCS) heat exchanger within the required time to ensure cooling is available to Reactor Coolant System (RCS) seal injection water during an Appendix R event. Subsequent analysis of actual plant data concluded that TVA could have established cooling to RCS seal injection during the subject Appendix R event, and achieved and maintained fire safe shutdown. Therefore, this event was not reportable under 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. R2DO (Bartley) notified. | Reactor Coolant System | |
| ENS 48056 | 29 June 2012 14:28:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Increase in Probable Maximum Flood Level | On June 29, 2012, TVA issued an updated calculation titled 'PMF Determination for Tennessee River Watershed' The calculation resulted in an increase in the Watts Bar Nuclear (WBN) probable maximum flood (PMF) level from Elevation 734.9 to Elevation 739.2. All flood sensitive safety related systems, structures, and components have been reviewed and been determined to remain unaffected by the revised PMF surge elevation, with the exception of the Thermal Barrier Booster Pump Motors and Essential Raw Cooling Water (ERCW) equipment required for flood mode operation located on Elevation 722 of the Intake Pumping Station (IPS). The updated PMF of Elevation 739.2 could impact the ability of the thermal barrier booster pumps and the Elevation 722 IPS ERCW equipment to perform their design accident protection function. Because of the unanalyzed condition. the potential existed for WBN to exceed its PMF design basis and adversely affect plant safety. This notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures have been prepared to install a temporary flood protection barrier around the thermal barrier booster pumps and provide additional protection of Elevation 722 of the IPS in the event of a flood alert. The potential for the increased PMF level and the associated compensatory measures were previously discussed in a public meeting between TVA and the NRC on May 31, 2012 and in correspondence between TVA and the NRC dated June 13, 2012 and June 25, 2012. All safety related equipment is currently operable. There are no indications of conditions that might result in a flood in the near term. The licensee notified the NRC Resident Inspector of this condition.
Based upon continuing engineering reviews, the chilled water circulating pump motors for the Train A and B Main Control Room and 6.9kV Shutdown Board Room, including various sub-components, would be partially submerged during a Probable Maximum Flood (PMF) event. These components were not previously considered as affected by the PMF. The affected components are located on floor elevation 737.0 of the auxiliary building. This notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures have been prepared to install temporary flood protection barrier around the chilled water circulating pump motors and provide additional protection of Elevation 722 of the IPS in the event of a flood. All safety related equipment is currently operable. There are no indications of conditions that might result in a flood in the near term. The licensee notified the NRC Resident Inspector of this condition. Licensee Event Report 50-390/2012-002-00 will be supplemented to include a description of the potential PMF affects on the chilled water circulating pump motors. Notified the R2DO (Ernstes). | ||
| ENS 48723 | 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 Watts Bar Unit 1. TVA initiated immediate actions to address the condition by conducting additional analyses and the development of 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, the essential raw cooling water system, the thermal barrier booster pumps and the control room chillers. Additional details regarding the modular flood barriers and the results of TVA's subsequent hydrologic analyses for Watts Bar 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 July 19, 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 Browns Ferry (EN #48724) and Sequoyah (EN #48725). | Emergency Diesel Generator |