05000336/LER-2011-002
| Millstone Power Station - Unit 2 | |
| Event date: | 06-20-2011 |
|---|---|
| Report date: | 08-18-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3362011002R00 - NRC Website | |
1. Event Description At 1152 on June 20, 2011, with Millstone Power Station Unit 2 operating at 59 percent power in Mode 1, the reactor tripped automatically on low steam generator (SG) water level 300 microseconds prior to a manual reactor trip initiated from the control room. The decreasing SG water level condition was due to a low suction pressure trip on the operating "B" steam generator main feedwater pump (SGFP) [SJ]. The "B" SGFP low suction pressure trip occurred while attempting to place the "A" steam generator main feedwater pump in service. Established operating procedures had recently been changed to permit starting the second SGFP at a higher power level than previously performed.
Auxiliary Feedwater (AFW) [BA] initiated as expected following loss of the operating SGFP. Standard post trip actions were carried out, and all other safety systems responded as expected.
The event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual or automatic actuation of any of the systems listed in 50.73(a)(2)(iv)(B), including the Reactor Protection System.
The actuation of the Auxiliary Feedwater Actuation System also is a reportable condition under the same paragraph.
2. Cause The cause of the event was gaps in the application of operator fundamentals and some procedure quality issues associated with operations procedure OP 2204, Load Changes.
3. Assessment of Safety Consequences The safety consequences of this event were low. The reactor automatically tripped on low steam generator level.
Steam generator levels remained in the visible range on narrow range SG level instrumentation. All other safety systems responded as expected.
There was no loss of decay heat removal capability, because main and auxiliary feedwater pumps were available to feed the steam generators. Neither departure from nucleate boiling nor fuel centerline melt design limits were challenged. As such, there were no challenges to the fuel, reactor coolant system or containment fission product barriers.
4. Corrective Action Procedure OP 2204 has been revised to ensure the second steam generator feedwater pump is placed in service at a lower power level. Additional corrective actions to address the underlying causes of the gaps in the application of operator fundamentals are being addressed in accordance with the station's corrective action program.
5. Previous Occurrences No previous similar events/conditions were identified.
Energy Industry Identification System (EDS) codes are identified in the text as [XX].