05000286/LER-2009-006, Re Automatic Trip Due to a Turbine-Generator Trip Caused by Actuation of the Generator Protection System Lockout Relay During a Severe Storm with Heavy Lightning
| ML092930088 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/06/2009 |
| From: | Joseph E Pollock Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-09-120 LER 09-006-00 | |
| Download: ML092930088 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2862009006R00 - NRC Website | |
text
.Entergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 734-6700 J. E. Pollock Site Vice President NL-09-120 October 6, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2009-006-00, "Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by Actuation of the Generator Protection System Lockout Relay During a Severe Storm with Heavy Lightning" Indian Point Unit No. 3 Docket No. 50-286 DPR-64
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2009-006-00. The attached LER identifies an event where the reactor automatically tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A). As a result of the reactor trip, the Auxiliary Feedwater system was actuated and one of three Emergency Diesel Generators started and energized it's assigned bus which is also reportable under 10 CFR 50.73(a)(2)(iv)(A). During the subsequent investigation it was determined that the unit operated outside of Technical Specifications for one offsite circuit which was inoperable. This is reportable under 10 CFR 50.73(a)(2)(i)(B). These conditions were recorded in the Entergy Corrective Action Program as Condition Reports CR-IP3-2009-03375 and CR-IP3-2009-03380.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 734-6710.
Sincerely JEP/cbr cc:
Mr. Samuel J Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Mr. Paul Eddy, New York State Public Service Commission LEREvents@inpo.org
Abstract
On August.10,: 2009-;7:during-a. severe thunderstorm,, the Control-Room received indication of.a. Turbine Trip-and-Reactor-Trip initiated.-by the Generator Primary Lockout Relay (86P).
All control rods fully inserted and all required safety systems functioned properly.
6.9 kV Bus 2 failed to auto transfer to 6.9 kV bus 5 resulting in the trip of the 34 Reactor Coolant Pump and de-energization of 480 volt safeguards Bus 2A.
Emergency Diesel Generator (EDG) 31 automatically started and re-energized 480 volt bus-.2A.
The failure of the fast bus transfer meant one offsite circuit was inoperable and applicable Technical Specifications were not met.
The plant was stabilized in hot standby with decay heat being removed by the main condenser.
There was no radiation release.
EDG-32 and 33 remained in standby as offsite power remained available.
The Auxiliary Feedwater (AFW)
System automatically started as expected due to Steam Generator low level from shrink effect.
6.9kV Bus 2 was energized from 6.9 kV bus 5 via closure of 6.9 kV bus 2-5 tie Breaker.
The cause of the trip was indeterminate.
The trip was a lightning induced event with two possible causes: Actuation of 1) one of the generator protection relays 87G,
- 59N, 87T21,
- 87T22, or 85LI/345, or 2) issues with the IPEC ground mat.
Each possible cause will be evaluated after additional data is collected and analyzed.
Corrective actions completed included inspection of 345 kV output feeder W96, and testing of pilot wire cables.
Calibration and testing of likely relays will be performed and data collected and analyzed.
The pilot wire will be evaluated for susceptibility to lightning induced fault and the ground grid condition will be evaluated. The event had no effect on public health and safety.
(If more space is required, use additional copies of (if more space is required, use additional copies of (if more space is required, use additional copies of (if more space is required, use additional copies of NRC Form 366A) (17) between July 6, 2009 and July 11, 2009 when the breaker was replaced.
Past Similar Events A review was performed of Unit 3 Licensee Event Reports (LERs) over the past six years for Unit 3 events that involved a RT from a lightning strike on the unit's 345 kV feeder to the Buchanan Substation.
No Unit 3 LERs were identified.'
A review of Unit.2 LERs identified LER-2003-004 that reported a TT-RT on August 3,' 2003 due to a lightning strike on Unit 2 tower No.
51 on 345kV feeder W93.
There was one other Unit 2 trip reported in LER-1980-006 for a direct tip from the Buchanan Substation on June 3, 1980 due to a lightning strike on one of the 345kV/138kV transmission towers between the Buchanan Substation North Ring Bus and Millwood Substations.
A similar review was performed for-events that involved a failure of a fast bus transfer breaker to operate properly.
No similar events were identified.
Safety Significance
This event had no effect on the health and safety of.the public.'
There were no actual safety consequences for the event because there were no other transients or accidents. at the time of the RT and no other event requiring fast bus transfer.
Required primary safety systems performed as designed when the RT was initiated.
The AFWS actuation was an expected reaction as a result of low SG water level due to SG void fraction (shrink), which occurs after a RT and main steam back pressure as a result of. the rapid reduction of steam flow due to turbine control-valve closure.
As a result of 6.9 kV breaker UT2-ST5 failing to close on Turbine Trip Fast Transfer the EDG-31 started and energized its assigned Bus 2A in accordance with design.
There were no significant potential safety consequences of this event under reasonable and credible alternative conditions.
A RT and the reduction in SG level is a condition for which the plant is analyzed.
This event was bounded by the analyzed event described in FSAR Section 14.1.8, Loss of External Electrical Load.
The response of the plant is evaluated for a complete loss of steam load from full power without a direct RT and includes the acceptability of a loss of steam load without direct RT on turbine trip below 35 percent power.
The analysis shows that the plant design is such that there would be no challenge to the integrity of the reactor coolant system or main steam system and no core safety limit would be violated.
A low SG water level initiates actuation of the AFWS whose design has adequate capability to provide the minimum required flow assuming a single failure.
For this event, rod controls were in Auto and all rods inserted upon initiation of the automatic RT.
The AFWS actuated and provided required FW flow to the SGs*
RCS pressure remained below the set point for pressurizer PORV or code safety valve operation and above the set point for automatic safety injection'actuation.
Pressurizer level remained on scale.
Following the RT, the plant was stabilized in hot standby.