05000286/LER-2008-006, Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing Due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt Safeguards Bus 6A
| ML083450062 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/03/2008 |
| From: | Joseph E Pollock Entergy Nuclear Indian Point 3, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-08-168 LER 08-006-00 | |
| Download: ML083450062 (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)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2862008006R00 - NRC Website | |
text
Indian Point Energy Center 450 Broadway, GSB En P.O. Box249 tBuchanan, N.Y. 10511-0249 EntaWTel (914) 734-6700 J. E. Pollock Site Vice Piesident December 3, 2008 Indian Point Unit No. 3 Docket No. 50-286 NL-08-168 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
Subject:
Licensee Event Report # 2008-006-00, "Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt Safeguards Bus 6A"
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2008-006-00. The attached LER identifies an event where there was an automatic actuation of an emergency diesel generator and two auxiliary feedwater pumps, systems listed in 10 CFR 50.73(a)(2)(iv)(B), which is reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2008-02519.
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, J. E. Pollock 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 October 9, 2008, Instrumentation & Control (I&C) Technicians were performing surveillance testing, when the normal power supply breaker for 480 Volt AC safeguards bus 6A inadvertently opened, causing the bus to de-energize.
In accordance with design, loads on bus 6A were stripped and then sequentially re-loaded back onto the bus.
The bus undervoltage control circuitry caused an automatic actuation of emergency diesel generator (EDG) 32 which started and re-energized the bus, actuation of motor driven auxiliary feedwater pump (AFWP) 33 and steam driven AFWP 32.
All systems performed as designed.
AFW was injected into the steam generators (SG) that resulted in a 1-2% change in SG level, an approximate 0.1% increase in reactor power and no control rod movement.
The most probable cause of the inadvertent actuation was an intermittent fault in the digital volt meter used for the test.
Corrective actions
included revision of the test procedures for 480 Volt buses so that the voltage checks of the degraded grid voltage Agastat are not taken across contacts.
An independent vendor will perform an analysis of the digital volt meter used during the test. 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)
This event meets the reporting criteria because the 32 EDG actuated to start and the 32 and 33 AFWP actuated to start when the UV control circuit on 480 Volt AC Bus 6A actuated.
On October 9, 2008, the normal power supply to safeguards bus 6A was inadvertently de-energized and in accordance with design the 32 EDG, AFWP-32 and 33 automatically started at approximately 12:54 hours.
At 12:58 hours, the 33 AFWP was secured and at 13:19 hours the 32 AFWP was secured.
At 14:35 hours, 480 Volt AC Bus 6A was returned to its normal power supply and the condition for Technical Specification 3.8.1 was exited.
At 14:41 hours, the 32 EDG was secured and returned to its normal standby condition.
All required safety systems performed as designed.
As a result of the event, there were no safety systems that were not capable of performing their safety function.
In accordance with reporting guidance in NUREG-1022, an additional random single failure need not be assumed in that system during the condition.
Therefore, there was no safety system functional failure reportable under 10 CFR 50.73 (a) (2) (v).
Past Similar Events A review was performed of Licensee Event Reports (LERs) for the past three years for any events reporting inadvertent Engineered Safety Feature actuation during testing.
The review identified three LERs: LER-2005-003, LER-2008-003, and LER-2008-004.
LER-2005-003 reported an inadvertent actuation of AFWPs 31 and 33 on May 16, 2005, during Reactor Protection Logic Functional testing.
The cause of the event was human error due to inadequate work practices where the technician performing the test failed to adhere to the procedure.
LER-2008-003 reported the actuation of an EDG due to the inadvertent action of the UV sensing circuit on bus 5A.
The cause of the inadvertent actuation was procedure use and adherence.
LER-2008-004 reported inadvertent actuation of AFWPs 31 and 33 during Reactor Protection Logic Channel Functional testing caused by incorrect jumper connection due to personnel error.
The events reported in LER-2005-003, LER-2008-003, and LER-2008-004 do not have a similar cause to this event because those events were caused by human performance errors and not potentially faulty test equipment.
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 accidents or transients requiring the EDGs.
Required power from both offsite sources and onsite emergency power were available and the actuation circuitry and EDG performed in accordance with design and minimum safeguards power was available to power safety loads.
There was no significant core reactivity change as there was no automatic movement of the control rods, reactor power increased approximate 0.1%, and there was a 1-2% change in SG level.
The changes were well within the actuation limits of the reactor protection system.