05000286/LER-2012-004, Regarding Automatic Reactor Trip as a Result of a Turbine-Generator Trip Due to a Loss of 345 Kv Feeders W97 and W98 Caused by Storm Damage to Feeder Insulators
| ML13002A019 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/11/2012 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-12-175 LER 12-004-00 | |
| Download: ML13002A019 (5) | |
| 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 |
| 2862012004R00 - NRC Website | |
text
- Entergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 John A. Ventosa Site Vice President NL-12-175 December 11, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2012-004-00, "Automatic Reactor Trip as a Result of a Turbine-Generator Trip Due to a Loss of 345 kV Feeders W97 and W98 Caused by Storm Damage to Feeder Insulators" 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) 2012-004-00. The attached LER identifies an event where the reactor was 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, which is also reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2012-03425.
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) 254-6710.
Sincerely, cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Mrs. Bridget Frymire, New York State Public Service Commission LEREvents@INPO.org
Abstract
On October 29,
- 2012, an automatic reactor trip (RT) was initiated as a result of a main turbine-generator trip due to a trip of the Generator Primary and Backup Lockout relay 86P and 86BU on a direct trip from the Buchanan switchyard.
The Buchanan switchyard south ring bus was isolated from output feeder W96 as a result of faults on 345 kV feeders W97 and W98 causing the generator output breakers 1 and 3 to open initiating a direct trip signal from Buchanan.
All control rods fully inserted and all required safety systems functioned properly.
The plant was stabilized in hot standby with decay heat being removed by the main condenser.
The Auxiliary Feedwater System automatically started as expected due to SG low level from shrink effect.
The Emergency Diesel Generators did not start as offsite power remained available and stable.
The cause of the RT was the trip of generator output breakers 1 and 3 due to isolation of the south ring bus from output feeder W96.
Generator output breakers 1 and 3 tripped as a result of a fault on feeders W97 and W98 due to Con Edison 345 kV feeder line insulator damage from the effects of superstorm Sandy.
Corrective action was taken to repair 345 kV feeders W97 and W98 by Con Edison.
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 NRC Form 366A) (17)
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 the event was an uncomplicated reactor trip with no other transients or accidents.
Required primary safety systems performed as designed when the RT was initiated.
The AFWS actuation was an expected reaction as a result of low Steam Generator (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 stop and control valve closure.
There were no significant potential safety consequences of this event.
The RPS is designed to actuate a RT for any anticipated combination of plant conditions including a direct RT on TT unless the reactor is below approximately 20% power.
The analysis in UFSAR Section 14.1.8 concludes an immediate RT on TT is not required for reactor protection.
A RT on TT is provided to anticipate probable plant transients and to avoid the resulting thermal transient.
If the reactor is not tripped by a TT, the primary tripping functions of over temperature delta temperature (OTDT) or over power delta temperature (OPDT) trip would prevent safety limits from being exceeded.
Additional tripping functions are provided as a backup for specific accident conditions and mechanical failures.
The generator is protected by the generator protection system (GPS) which is designed to protect the generator from internal and external faults by tripping the output breakers.
During this event the GPS functioned as designed and initiated a turbine-generator trip.
This event was bounded by the analyzed event described in UFSAR 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.
The RT and the reduction in SG level is also a condition for which the plant is analyzed.
A low water level in the SGs initiates actuation of the AFWS.
The AFW System has adequate redundancy to provide the minimum required flow assuming a single failure.
The analysis of a loss of normal FW (UFSAR Section 14.1.9) shows that following a loss of normal FW, the AFWS is capable of removing the stored and residual heat plus reactor coolant pump waste heat thereby preventing either over pressurization of the RCS or loss of water from the reactor.
For this event, rod control was in automatic and all rods inserted upon initiation of a 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.
Following the RT, the plant was stabilized in hot standby.