05000286/LER-2003-005
| Indian Point Unit 3 | |
| Event date: | 08-14-2003 |
|---|---|
| Report date: | 10-14-2003 |
| 2862003005R00 - NRC Website | |
DESCRIPTION OF EVENT
Note: The Energy Industry Identification System Codes are identified within the brackets {1.
On August 14. 2003, at approximately 1611 hours0.0186 days <br />0.448 hours <br />0.00266 weeks <br />6.129855e-4 months <br />, during 100% steady state power, Indian Point Unit 3 experienced an automatic reactor trip (RT) {JE} initiated by a loss of off-site power due to a grid disturbance. The loss of off-site power (LOOP) was associated with the blackout that affected parts of northeastern United States and Ontario, Canada. The degraded grid caused an under-frequency breaker trip on the 34 Reactor Coolant Pump (RCP). The trip of 34 RCP breaker initiated a RT on low approximately 35% power, permissive net point). The plant stabilized in natural circulation and the Emergency Diesel Generators (EDGs)(EK) 31, 32, and 33 started automatically and energized the 480V buses. Main feedwater system isolated and the Auxiliary Peed Water System (AFW)(EA) pumps automatically started. The AFW flow control valves associated with AFW pumps 31 and 33 subsequently lost pneumatic control and manual control was assumed. Other equipment that failed to operate properly included the 34 Main Steam Line (MSL) Safety Valve (lifted prematurely), 32 Source Range Monitor Nuclear Instrument failed, Spent Fuel Cooling was lost for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 32 CRDM fan tripped and the Technical Support Center (TSC) Diesel tripped on over speed. The condition for the TSC Diesel was recorded as CR-IP3-2003-04706.
The TSC Diesel failure did not prevent activation of the emergency plan when required. None of the equipment issues precluded the return of the Unit to power.
No actuation of the Safety Injection System occurred nor was required as a result of this trip and no Power Operated Relief Valves actuated during this event. The Pressurizer Code Safety Valves remained closed throughout this transient. This event was entered into the Entergy Corrective Action Process under CR-11,3-2003- 04698.
CAUSE OF EVENT
The cause of the reactor trip was a loss of off-site power due to grid disturbance. The root cause of the grid disturbance which resulted in a blackout for parts of northeastern United States and Ontario, Canada is under investigation by a joint United States and Canadian government special task force. The grid disturbance caused the vain generator to have lower frequency.
The 34 RCP breaker tripped on under-frequency, which resulted in a RPS logic trip of the reactor on loss of RCS loop flow.
CORRECTIVE ACTIONS
The reactor experienced an automatic trip and the plant shutdown as designed. All emergency systems initiated as required. Corrective actions for the event included a post trip review, a root cause evaluation, and plant walkdown. No specific corrective actions to preclude loss of off-site power due to a similar event were identified.
EVENT REPORTING
This event is reportable under 10 CFR 50.73 (a) (2) (iv) (A). The licensee shall report any event or condition that resulted in manual or automatic actuation of any of the systems listed in 10 CFR 50.73 (a) (2) (iv) (B). Systems to which the requirements of 10 CFR 50.73 (a) (2) (iv) (A) apply includes the Reactor Protection System including reactor scram or reactor trip, AFW system and the MOM.
PAST SIMILAR EVENTS
A review of previous occurrences when IP3 had experienced unit trip due to a loss of off-site power was performed. Within the past three years, no such occurrences were identified. However, unit trips as a result of loss of off-site power were reported to the NRC for Indian Point 2 via LER 2003-004-00, 2003-003-00 and 2001- 007-00. These trips were for a single unit unlike the present event where both Indian Point 2 and 3 tripped simultaneously.
EVENT SAFETY SIGNIFICANCE
There were no significant safety consequences for this event because the plant systems responded as expected except as noted. No pressurizer safety valves lifted and no actuation of the safety injection system was required. There were no nuclear safety concerns exhibited during the event and all fission product barriers remained intact. There was no significant impact on the health and safety of the general public.
The loss of a reactor coolant pump is described in the UFSAR Section 14.1.6, "Loss of Reactor Coolant Flow.
- This event was initiated when the Unit was operating at 100 1 power and is bounded by the UFSAR analysis.
The loss of power to station auxiliaries is described in UFSAR Section 14.1.12, "Loss of Station Auxiliaries.' The design event as described in the UFSAR results in a loss of offsite power to both 6.9kV and 480V busses. In this event, the loss of power was per this design event and was bounded by the UFSAR analysis.
MC FORM SEM (14001)