05000286/LER-2007-001, Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels as a Result of the Loss of Feedwater Flow Caused by the Failure of 32 Main Feedwater Pump Train a Control Logic Power Supply
| ML071620123 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/04/2007 |
| From: | Dacimo F Entergy Nuclear Indian Point 3 |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-07-052 LER 07-001-00 | |
| Download: ML071620123 (6) | |
| 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)(v), Loss of Safety Function |
| 2862007001R00 - NRC Website | |
text
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 nBuchanan, N.Y. 10511-0249 Int..-
Tel (914) 734-6700 Fred Dacimo Site Vice President June 4, 2007 Indian Point 3 Docket No. 50-286 NL-07-052 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1 -17 Washington, D.C. 20555-0001
Subject:
Licensee Event Report # 2007-001 -00, "Manual Reactor Trip Due to Decreasing Steam Generator Levels as a Result of the Loss of Feedwater Flow Caused by the Failure of 32 Main Feedwater Pump Train A Control Logic Power Supply"
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-001-00. The attached LER identifies an event where the reactor was manually tripped while critical, 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-IP3-2007-01775.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. T. R. Jones, Manager, 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 3 Mr. Paul Eddy, New York State Public Service Commission INPO Record Center
Abstract
On April 3, 2007, with the 31 Main Boiler Feedwater Pump (MBFP) shutdown, the 31 MBFP control circuit was de-energized in preparation for maintenance, when unexpectedly the logic control circuit for the 32 MBFP was lost resulting in loss of feedwater (FW) flow.
Control room operators manually initiated a reactor trip (RT) at approximately 0417 hours0.00483 days <br />0.116 hours <br />6.894841e-4 weeks <br />1.586685e-4 months <br />, after observing rapidly decreasing 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 initially 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 decreasing SG levels due to loss of 32 MBFP FW control as a result of the failure of the 32 MBFP Train A Lovejoy control logic power supply when the 31 MBFP instrument bus power supply was isolated.
The 32 MBFP Train A control logic power supply failure was due to a bad pin connection on the power supply voltage regulator.
The root cause was a failure of an auctioneered power supply that was not self revealing and was undetectable until it was required to carry load.
Significant corrective actions include replacement of 32 MBFP Train A control logic power supply, establishment of a the 10 year Preventive Maintenance (PM) to replace all power supplies in the Lovejoy system, performance of an assessment for identification of other auctioneered power supplies and verification that PMs confirm that the power supplies can carry their load. 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)
This event meets the reporting criteria because a manual RT was initiated at 0417 hours0.00483 days <br />0.116 hours <br />6.894841e-4 weeks <br />1.586685e-4 months <br />, on April 3,
- 2007, and the AFWS actuated as a result of the RT.
Although the MSIVs were closed due to low reactor coolant system average temperature, the action was not to mitigate the consequences of the event.
The failure of the MBFP speed controller power supply 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 unit 3 events that involved a RT from loss of a power supply identified no applicable LERs.
However, unit 2 had three LERs in the past two years that reported RTs as a result of power supply failures.
LER-2006-003 reported a RT due to a mismatch of reactor power to turbine load during a power reduction from a loss of heater drain tank (HDT) pumps.
The HDT pumps were lost due to a failure of the power supply for the HDT level transmitter (LT).
The direct cause of the HDT LT failure was a faulty capacitor.
The root cause was a programmatic weakness with classification and management of critical PM tasks.
LER-2006-005 reported a RT due to a generator exciter trip caused by a Generrex power supply failure.
The power supply failed due to a bad common ground connection.
LER-2007-004 reported a RT due loss of FW flow as a result of a failure of the power supply for the MBFP suction pressure transmitter.
The power supply failed due to capacitor aging.
The unit 2 LER-2007-004 included a CA to develop and implement at IPEC an Instrument Power Supply PM in accordance with the Entergy Nuclear South (ENS)
PM Template.
The Instrument Power Supply PM CA also included auctioneered power supplies discovered in this event.
The CA for LER-2007-004 could have prevented this event but the CA had not yet been implemented due to the short time period from the previously reported event (LER-2007-004).
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 main steam isolation valves were manually closed by operations as a result of low RCS temperature that was due to excessive cooling as a result of a new core with low decay heat.
Reactor core cooling through the SG was controlled by use of the main steam power operated relied valves (PORVs) rather than by use of the steam dumps to the condenser.
A condition of excessive heat removal from the SGs is bounded by the analysis in FSAR Section 14.2.5, "Rupture of a Steam Pipe," which bounds excessive FW flow with reactor at zero power analyzed in Section 14.1.10, "Excessive Heat Removal Due to FW System Malfunctions."
The AFWS actuation was an expected reaction 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 loss of FW flow for this event was bounded by the analysis in FSAR Section 14.1.9, "Loss of Normal FW."
The AFWS actuated and provided required FW flow to the SGs.
The AFW capacity is sufficient to provide the minimum required FW flow to the SGs.
For this event the 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.