05000247/LER-2007-004, Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of Feedwater Pump Suction Pressure Transmitter Power Supply Failure
| ML071280694 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 04/30/2007 |
| From: | Dacimo F Entergy Nuclear Operations |
| To: | Document Control Desk, Plant Licensing Branch III-2 |
| References | |
| NL-04-044 LER 07-004-00 | |
| Download: ML071280694 (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 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2472007004R00 - NRC Website | |
text
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 E lfjffgy Buchanan, N.Y. 10511-0249 Tel (914) 734-6700 Fred Dacimo Site Vice President Administration April 30, 2007 Indian Point 2 Docket No. 50-247 NL-07-044 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, D.C. 20555-0001
Subject:
Licensee Event Report # 2007-004-00, "Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of Feedwater Pump Suction Pressure Transmitter Power Supply Failure"
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-004-00. The enclosed LER identifies an event where the reactor was manually tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A).
This condition has been recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2007-01046.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. T. R. Jones, Manager, IPEC Licensing at (914) 734-6670.
Sincerely, Fred R. Dacimo Site Vice President Indian Point Energy Center 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 INPO Record Center
Abstract
On February 28, 2007, control room operators manually initiated a reactor trip (RT) at approximately 0633 hours0.00733 days <br />0.176 hours <br />0.00105 weeks <br />2.408565e-4 months <br />, after receiving speed control trouble alarms for the 21 and 22 Main Boiler Feedwater Pumps (MBFPs) and observing decreasing MBFP speed and steam generator (SG) levels.
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.
There was no radiation release.
The Emergency Diesel Generators did not start as offsite power remained available.
The Auxiliary Feedwater System automatically started as expected due to Steam Generator low level from shrink effect.
The cause of the RT was loss of main feedwater (FW) flow due to the failure of the power supply (PQ-408B) for the MBFP suction pressure transmitter (PT-408B).
The power supply failed due to a failure of its filter capacitors as a result of age degradation.
The root cause of the power supply failure was insufficient verification of the existing plant programs to address capacitor age degradation due to human error.
Significant corrective actions include replacement of power supply PQ-408B and pressure transmitter PT-408B.
An instrument power supply Preventive Maintenance (PM) will be implemented in accordance with the ENS PM template, and a capacitor program will be developed and implemented to address age degradation of capacitors.
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 NRC Form 366A) (17)
Event Analysis
The event is reportable under 10CFR50.73(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 under 10CFR50.73(a) (2) (iv) (B).
Systems to which the requirements of 10CFR50.73(a) (2) (iv) (A) apply for this event include the Reactor Protection System (RPS) including RT and AFWS actuation.
This event meets the reporting criteria because a manual RT was initiated at 0633 hours0.00733 days <br />0.176 hours <br />0.00105 weeks <br />2.408565e-4 months <br />, on February 28, 2007, and the AFWS actuated as a result of the RT.
The failure of the MBFP suction pressure transmitter power supply PQ-408B did not result in the loss of any safety function.
Therefore, there was no safety system functional failure reportable under 10CFR50.73(a) (2) (v).
PAST SIMILAR EVENTS A review of the past two years of Licensee Event Reports (LERs) for events that involved a RT from loss of a power supply was performed.
There were two LERs identified that reported a RT due to a power supply failure.
LER-2006-003 reported a manual RT due to a mismatch between reactor power and turbine load caused by cycling of steam dump valves after a power reduction for loss of Heater Drain Tank (HDT) pumps.
The cause of the loss of the HDT pumps was a failed HDT level controller power supply.
LER-2006-003 was initiated by a level controller whose aging power supply failed.
LER-2006-005 reported an automatic RT due to a turbine trip as a result of a Main Generator Exciter protective trip caused by a Generrex Power Supply loss.
The power supply failed due to a loose terminal screw where the grounds are mounted.
The cause of the event for LER-2006-005 was a loose power supply connection associated with a degraded power supply due to high resistance connections as a result of oxidation residue (aging).
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 RT with no other transients or accidents.
Required primary safety systems performed as designed when the RT was initiated.
There were no risk related components out of service at the time of the RT.
The AFWS actuation was expected as a result of low SG water level due to SG void fraction (shrink), which occurs after automatic RT from full load.
There were no significant potential safety consequences of this event under reasonable and credible alternative conditions.
The AFWS actuated and provided required FW flow to the SGs.
There are two motor driven AFW pumps and one steam driven pump, any one of which could provide the minimum required FW flow to the SGs.
Main FW remained available but due to the cutback signal from PT-408B (failed low) was at minimum flow.
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.