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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54257 | 6 September 2019 02:15:00 | 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 | En Revision Imported Date 11/8/2019 | EN Revision Text: CONTAINMENT PENETRATION DISCOVERED NOT ISOLATED At 2115 CDT on 9/5/2019, an inside containment test connection and inoperable outside containment isolation valve were discovered to be open for a containment air sample penetration. This resulted in the containment penetration not being isolated. The inside containment test connection was closed at 2322 CDT on 9/5/2019. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
This event was originally reported on September 6, 2019 under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B). Upon completion of the investigation of the event, it was determined that the event had insignificant safety consequences because the containment breach was disconnected from the Reactor Coolant System by a series of closed valves for the duration of the event. Additionally, the lines to the inside containment connection and the outside inoperable containment isolation valve that was found to be open as well as the main line connecting and passing through the penetration were one-inch diameter lines. Analysis determined that containment breaches that are less than a three-inch diameter do not lead to a large radiation release. The event did not place the plant in an unanalyzed condition that significantly degrades plant safety. Therefore, 10 CFR 50.72(b)(3)(ii)(B) did not apply to this event and this notification is to retract reporting under that criterion. The licensee notified the NRC Resident Inspector. Notified R4DO (Drake). | Reactor Coolant System | |
| ENS 49490 | 31 October 2013 22:12:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Fire Event Could Result in a Hot Short That Could Adversely Impact Safe Shutdown Equipment | While performing a review of industry OE (operating experience) concerning unfused ammeter circuits on station batteries, it was discovered that the ammeter circuits for all of the non-1E batteries are of a similar design to that described in the OE. Also, while reviewing additional DC circuits, it was discovered that the control circuit for the Turbine Generator Emergency Lube Oil pump is unfused, protected only by the motor circuit breaker with a trip setting of 350 amps. The concern is that under the fire safe shutdown rules it is postulated that a fire in one fire area can damage these circuits and cause short circuits without protection that would overheat the cables and possibly result in secondary fires in other fire areas where the cables are routed. The secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. | 05000289/LER-2014-001 05000498/LER-2013-003 | |
| ENS 48277 | 22 August 2012 17:36:00 | 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 | Missing Flood Seal | During flooding walkdowns being performed on August 22, 2012, with the unit at 100 percent power, South Texas Project Unit 2 discovered the potential for water intrusion into the 10' Elevation Electrical Auxiliary Building (EAB) via a 2-inch underground conduit that was found to be missing its flood seal. It has been determined that the missing flood seal compromised the external flood design controls for the EAB. If flooding of the 10 (foot) EAB were to occur as a result of the missing flood seal, the operability of the Train A Engineered Safety Features (ESF) switchgear and the ESF Sequencers for all three Standby Diesel Generators could have been affected. Additionally, the Qualified Display Parameter System process cabinets (which control Auxiliary Feedwater flow and Steam Generator PORVs) and the Auxiliary Shutdown Panel are also located on the 10' Elevation. Repairs have been made and the 2-inch conduit is sealed. The event is being reported under 10 CFR 50.72(b)(3)(ii)(B) for Unit 2 being in an unanalyzed condition that significantly degraded plant safety, and under 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
The purpose of this call is to retract the notification made on 09/05/2012, Event Number 48277. Further analysis indicates that water intrusion resulting from the missing 2-inch conduit seal would not have been sufficient to affect the operability of the equipment located on the 10-foot elevation of the Unit 2 Electrical Auxiliary Building. It has been determined that the maximum water depth would not have exceeded 2 inches in depth and all safety related equipment on the 10-foot elevation is greater than 2 inches above the floor, therefore there would be no impact to any safety-related equipment. Accordingly, this event notification is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Geoffrey Miller). | Steam Generator Auxiliary Feedwater | |
| ENS 45631 | 8 July 2008 03:47:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Loss of Fire Suppression Capability | This is an 8-hour notification being made in accordance with 10CFR 50.72(b)(3)(ii)(B) for an event or condition that results in the plant being in an unanalyzed condition that significantly degrades plant safety. This notification is being made as the result of the re-review of a July 7, 2008 occurrence which resulted in an inadvertent isolation of a large portion of the (fire suppression) ring header affecting all Unit 2 fire suppression and a portion of Unit 1 fire suppression. The Unit 1 ring header isolation did not have an affect on the fire safe shutdown capability of Unit 1. However, three areas in Unit 2 which credit the availability of fire suppression to assure that the safe shutdown capability could have been achieved did not have fire suppression for approximately 3 hours. A Licensee Event Report will be submitted within 60 days. The licensee notified the NRC Resident Inspector. | 05000499/LER-2010-002 | |
| ENS 40612 | 27 March 2004 04:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Regarding Hhsi Flush Line Isolation Valve Leakage - | While performing a High Head Safety Injection (HHSI) Train-B inservice test, a Plant Operator noted leakage coming through HHSI Flush Line Isolation Valve #SI-0120B. This valve is located on a flush line coming from the discharge of the HHSI pump and directs water to the Safety Injection Pump room sump located within the Fuel Handling Building. The operator investigated and determined that the leakage was due to the valve not being fully seated. The Licensee has come to the conclusion that the estimated leakage through the valve exceeded the allowable value to remain in compliance with 10 CFR 50, Appendix A, GDC 19 limits for Control Room Envelope during the post LOCA Containment recirc phase of a large break LOCA. This condition resulted in an unanalyzed condition that significantly degraded plant safety requiring a notification to the NRC within eight hours per 10 CFR 50.72 (b) (3) ii. Subsequently, flush line valve #SI-0120B has been fully seated, and the leakage stopped. The licensee considers this valve to be operable. The Licensee notified the NRC Resident Inspector. | Control Room Envelope | |
| ENS 40380 | 9 December 2003 21:38:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unable to Achieve Cold Shutdown Using Natural Circulation at a Cooldown Rate of 50 Degrees Per Hour | At 1538 on 12/09/03, it was determined that the plant could not achieve cold shutdown using Natural Circulation with the available volume of the Auxiliary Feed Water Storage Tank (AFWST) and at an allowed plant cooldown rate of 50 degrees F/hour. The plant has a functional objective to achieve Plant Shutdown from Operating Condition to Cold Shutdown Condition with the Natural Circulation Cooldown process with the volume of water available in the AFWST. Natural Circulation cooldown is required during Loss of Offsite Power (LOOP), long term cooling, Safety Grade Cold Shutdown and Appendix R type fire with LOOP. This condition also applies to Unit 2. STP (South Texas Project) changed from a Reactor Vessel Head (RVH) T-Hot Plant to a RVH T-Cold Plant after replacement of the plant's steam generators. As a result, the plant cooldown rate under natural circulation conditions was changed from 25 degrees F/hour to 50 degrees F/hour based on a Westinghouse evaluation. While performing confirmatory calculations to address a fire safe shutdown issue, it was determined that the plant can be cooled down with two steam generators using natural circulation. However, at a cooldown rate of 50 degrees F/hour, the two steam generators not being used for cooldown would stagnate with hot water in the primary side tubes. During the depressurization process, the hot water flashes to steam such that depressurization below approximately 1000 psig is significantly challenged. This results in filling the pressurizer and a loss of RCS pressure control. The fluid in the RCS (Reactor Coolant System) loops not receiving AFW will remain at saturated conditions until cooling is achieved by other means. Cooldown and depressurization of the RCS in this configuration will take substantially longer than assumed in the current analyses. Although there is adequate core cooling, the plant will be in a condition that is outside the existing design basis and not specifically addressed by the Emergency Operating Procedures. Therefore, this condition resulted in the plant being in an unanalyzed condition that significantly degraded plant safety. Thermal Hydraulic Analyses at a cooldown rate of 25 degrees F/hr demonstrated that natural circulation will occur in the two steam generators that are not being used for cooldown such that the stagnated hot water conditions would not occur in these steam generators. This would allow for depressurization of the reactor coolant system and successfully meet the functional objective. A compensatory action is in place to limit natural circulation cooldown rate to no greater than 25 degrees F/hr. The licensee has notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Auxiliary Feedwater |