05000247/LER-2008-001, Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of Feedwater Pump Speed Control Malfunction
| ML081490318 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/22/2008 |
| From: | Joseph E Pollock Entergy Nuclear Indian Point 2 |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-08-075 LER-08-001-00 | |
| Download: ML081490318 (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)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2472008001R00 - NRC Website | |
text
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 Administration May 22, 2008 Indian Point Unit No. 2 Docket No. 50-247 NL-08-075 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
Subject:
Licensee Event Report # 2008-001-00, "Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of a Feedwater Pump Speed Control Malfunction"
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a) (1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2008-001-00. The enclosed LER identifies an event where the reactor was manually tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A).
As a result of the reactor trip, the Auxiliary Feedwater System was actuated which is a system listed in 10 CFR 50.73(a)(2)(iv)(B) that 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-IP2-2008-01333.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, IPEC 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 2 Mr. Paul Eddy, New York State Public Service Commission INPO Record Center
Abstract
On March 23, 2008, at 2216, with reactor power at 94.5%,
as part of a planned coast down for a scheduled refueling outage, a manual reactor trip (RT) was initiated as a result of 22 Main Boiler Feed Pump (MBFP) rapid speed reduction causing lowering steam generator levels.
All control rods 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 off-site power remained available.
The Auxiliary Feedwater System automatically started as expected due to Steam Generator low level from shrink.
The direct cause was radio frequency interference (RFI) from camera use near a MBFP speed control signal processer.
The root cause of the event was lack of knowledge that a digital camera is an RFI source that can produce adverse effects on digital control components.
Contributing causes were a poor choice of the controlling procedure for camera activity, poor change management in implementing the review requirements without providing adequate review tools, failure to follow procedure.
Corrective actions
include preparation of specific procedural guidance on electronic interference sources, creating a change management plan to track implementation of the new procedure and process, perform a needs analysis for training, and a site communication of this event and lessons learned from the event.
The event had no effect on the 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)
" A Shift Order was issued to operations providing the event, lessons learned, and the direction to prohibit the use of all cameras or any known RF emitter in the CR and referenced the guidance in procedure SMM-MA-102.
- Develop guidance/fleet procedure on electronic interference sources and their control and use in sensitive areas. Scheduled completion is July 31, 2008.
- Evaluate this CR topic and related information for both initial orientation and General Employee continuing training. Issue any tracking/implementation actions resulting from the needs analysis process.
Scheduled completion is July 31, 2008.
- Develop a change management plan related to electronic interference based on digital equipment.
Scheduled completion is July 31, 2008.
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 22:16 hours, on March 23, 2008, and the AFWS actuated as a result of the RT.'
The malfunction of the MBFP speed control 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'was performed of the past three years of Licensee Event Reports (LERs) for events that involved a RT from a malfunction of the MBFP speed control.
There were no similar LERs identified.
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 TS related components out of service or off normal status of any safety systems 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 reduced 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.
Main FW remained available.
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.