05000247/LER-2010-001, Regarding Automatic Reactor Trip as a Result of a Turbine-Generator Trip Due to a Loss of Generator Field Excitation Caused by a Failed Exciter Rectifier
| ML100750250 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/09/2010 |
| From: | Joseph E Pollock Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-10-0195 LER 10-001-00 | |
| Download: ML100750250 (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 |
| 2472010001R00 - NRC Website | |
text
-i-fntrai 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-10-019 March 9, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-Pl-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2010-001-00, "Automatic Reactor Trip as a Result of a Turbine-Generator Trip Due to a Loss of Generator Field Excitation Caused by a Failed Exciter Rectifier" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2010-001-00. 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-IP2-2010-00157.
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 I*s' 4, ?o A.OCK.
cc:
Mr. Samuel J Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Mr. Paul Eddy, New York State Public Service Commission LEREvents@inpo.org 4/
Abstract
On January 11, 2010, an automatic reactor trip (RT) was initiated as a result of turbine trip due to a loss of main generator excitation.
The plant was stabilized in hot standby with decay heat being removed by the main condenser {SG}.
The Auxiliary Feedwater System automatically started as expected due to SG low level from shrink effect.
Prior to the RT one of four rectifiers (#24) of the main generator excitation system was believed to be electrically isolated and its cooling was isolated to stop an existing cooling leak.
Investigations determined the rectifier disconnect switch that was believed to be open remained in a closed condition.
This condition allowed current to continue to flow through the 24 rectifier diodes and because there was no cooling water the diodes failed resulting in a loss of field voltage to the exciter actuating a trip signal.
The direct cause of the RT was loss of generator field excitation due to loss of two diodes within the # 24 Generrex rectifier cabinet.
The root cause was failure of management to implement critical decision making.
A significant contributing cause was an improper lubricant used on disconnect switch contact surfaces.
Corrective actions included a brief of station personnel on the event and lessons learned.
A Generrex upgrade modification will be implemented including new disconnect switches in the spring 2010 refueling outage, case study training will be completed, computer based training (CBT) from the case study will be prepared and included in training curriculum, Alarm Response Procedure (2-ARP-SJF) and System Operating Procedure (2-SOP-24.4) will be revised for operation of the new disconnect switches, and maintenance personnel will be instructed on appropriate application of greases.
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)
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 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.
There were no significant potential safety consequences of this event under reasonable and credible alternative conditions.
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 (P-8).
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 over temperature delta temperature (OTDT) or over power delta temperature (OPDT) trip would prevent safety limits from being exceeded.
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 TT.
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 the reactor trip.
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.