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| Entered date | Event description | |
|---|---|---|
| ENS 50686 | 16 December 2014 17:31:00 | On 12/13/14 at 1700 (EST) hours, during a planned Unit 1 maintenance outage the licensee identified leakage from a weld on the (.75 inch) lower seal vent piping connected to the 1B reactor recirculation pump lower seal area. The location is within the reactor recirculation loop isolation valves, therefore isolable from the reactor vessel. The piping is ASME Class 2 and is a reactor coolant pressure boundary. The reactor was in mode 3 at the time of discovery. Control Room notified at 1345 (EST) on 12/16/14, that requirements for 10CFR50.72(b)(3)(ii)(A) were not met. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). The unit was taken to Mode 4 and the seal vent piping repair was completed. The NRC Resident Inspector has been notified. |
| ENS 50061 | 27 April 2014 01:45:00 | On 4/26/14 at 2322 EDT it was determined that the combined leakage for Main Steam Isolation Valves (including MSIV's, Main Steam Line Drains, HPCI Steam Supply and RCIC Steam Supply) per SR (Surveillance Requirement) 3.6.1.3.12 exceeded the minimum pathway limit of 300 scfh (standard cubic feet per hour). The MSIV Combined leakrate of 309 scfh exceeded the limit of 300 scfh with the Local Leak Rate Test failure of the HPCI Steam Supply Outboard Isolation Valve. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii), as it was discovered that the required leakage limits were exceeded. The licensee informed the NRC Resident Inspector.
The licensee is retracting this event based on the following: Subsequent engineering review identified an administrative error with procedures used to calculate MSIV leakage. Recalculation using revised procedures resulted in a MSIV Combined leakrate of 129 scfh, which is below the associated minimum pathway limit of 300 scfh specified in SR 3.6.1.3.12. Therefore, this condition is not reportable and EN 50061 is being retracted. The licensee informed the NRC Resident Inspector. Notified R1DO (Jackson). |
| ENS 49588 | 27 November 2013 14:58:00 | On November 27, 2013 at 1001 (EST), Susquehanna Steam Electric Station operators observed secondary containment differential pressure was at 0.04 inches water gauge for Zone I (Unit 1 Reactor Building). Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. Zone II (Unit 2 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable. Zone I differential pressure was impacted due to equipment malfunction. The inservice reactor building exhaust fan discharge damper developed an air leak at the solenoid operator. Zone 1 Building D/P was restored to within the required band at 1111 (EST) by placing the standby train exhaust fan, which was out of service for maintenance, in operation and verified to be stable. LCO 3.6.4.1 was entered for both units at 1001 (EST) and exited at 1131 (EST). This event is being reported under 10 CFR 50.72(b)(3)(v) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System. The NRC Resident Inspector has been informed. |
| ENS 49351 | 16 September 2013 21:11:00 | At 1455 EDT, (on 91/16/13), (Susquehanna) Engineering determined the leakage from the 2B RHR Pump Suction Relief valve caused the Engineered Safety Feature (ESF) Leakage to exceed the 2.5 gpm which was provided to the NRC during the implementation of the Alternate Source Term (AST) submittal. The calculated leakage rate was 7.5 gpm. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii). Unit 2 is currently in Mode 4 (cold shutdown) for a maintenance outage. This leaking RHR pump suction relief valve, previously identified in EN #49344, is being evaluated and repaired. The licensee has notified the NRC Resident Inspector and the state. |
| ENS 49342 | 14 September 2013 05:15:00 | At approximately 0330 hours on September 14, 2013, Susquehanna Steam Electric Station Unit 2 reactor was manually scrammed while transitioning the 'A' reactor feed pump from flow control mode to discharge pressure mode. Reactor water level rose to +54 inches causing a trip of reactor feedpumps. Subsequently the mode switch was taken to shutdown to manually scram the Unit 2 reactor. All control rods inserted. Reactor water level lowered to approximately +18 inches. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. No containment isolations occurred. All safety systems operated as expected. RCIC system was manually initiated for level control until a reactor feedpump was recovered, then RCIC was manually shutdown. The cause of the feedwater flow transient and trip of the reactor feedwater pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is being removed via the turbine bypass valve to the condenser. Offsite power remains stable, and there was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. The Pennsylvania Emergency Management Agency will be notified, and the licensee will be making a press release. |
| ENS 49263 | 10 August 2013 10:58:00 |