05000254/LER-2000-010-01, Re Automatic Reactor Scram from Low Reactor Vessel Level
| ML021340530 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/29/2002 |
| From: | Tulon T Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVP-02-032 LER-00-010-01 | |
| Download: ML021340530 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation |
| 2542000010R01 - NRC Website | |
text
ExelIo n.S Exelon Generation Company, LLC www.exeloncorp.com Nucea-r Quad Cities Nuclear Power Station 22710 206w' Avenue North Cordova, IL 61242-9740 April 29, 2002 SVP-02-032 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Generating Station, Unit 1 Facility Operating License No. DPR-29 NRC Docket No. 50-254
Subject:
Revision 01 of Licensee Event Report 254/00-010, "Automatic Reactor Scram from Low Reactor Vessel Level" Enclosed is Licensee Event Report (LER) 254/00-010, "Automatic Reactor Scram from Low Reactor Vessel Level," Revision 01, for Quad Cities Nuclear Power Station.
The root cause investigation for the event was completed subsequent to the submittal of the original LER. The LER has been revised to incorporate the pertinent root cause and corrective action information.
The following action is being committed to in this submittal:
The existing Feedwater Level Control System will be replaced with a more reliable Digital Feedwater Control System on Unit 1.
Any other actions described in the submittal represent intended or planned actions by Exelon Generation Company (EGC), LLC. They are described for the NRC's information and are not regulatory commitments.
Should you have any questions concerning this report, please contact Mr. W. J. Beck at (309) 227-2800.
Respectfully, Timey J.Tulon Si ice President Quad Cities Nuclear Power Station cc:
Regional Administrator - NRC Region Ill NRC Senior Resident Inspector - Quad Cities Generating Station
Abstract
At 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br /> on December 6, 2000, with reactor power at 100%, Unit One reactor water level started to decrease and the low-level alarm was received.
The operators took the Feedwater Level Control System (FWLC) out of three-element control and placed it in single-element control.
They then took manual control of the 'B' Feedwater Regulating Valve (FWRV) and opened the valve, causing reactor level to increase.
At about 0941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br /> reactor level started to decrease rapidly and the 'A' FWRV locked up.
A reactor scram signal was received at 0942 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.58431e-4 months <br />, with indicated reactor water level at 20 inches.
All automatic actions occurred consistent with a low-level reactor scram.
The root cause of this event was an age-related failure of a solder joint on the master controller for the 'A' FWRV.
Additionally, Operations procedures and training did not support reducing power to mitigate or prevent the reactor scram as an immediate operator action.
The safety significance of this event was minimal.
Although the event resulted in a reactor trip, all safety systems responded as designed.
This report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(iv).U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL IREVISION Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2000 010 01 2 of 3 (If more space is required, use additional copies of NRC Form 366A)(17)
PLANT AND SYSTEM IDENTIFICATION
Boiling Water Reactor, 2511 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION Automatic Reactor Scram from Low Reactor Vessel Level A.
CONDITION PRIOR TO EVENT
Unit: 1 Event Date: December 6, 2000 Event Time: 0942 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.58431e-4 months <br /> Reactor Mode: 1 Mode Name: Power Operation Power Level: 100%
Power Operation (1) - Mode switch in the RUN position with average reactor coolant temperature at any temperature.
B.
DESCRIPTION OF EVENT
At 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br /> on December 6, 2000, with reactor power at 100%, Unit One reactor water level started to decrease and the low-level alarm was received.
The operators took the Feedwater Level Control System (FWLC) [JB] out of three-element control and placed it in single-element control.
They then took manual control of the 'B' Feedwater Regulating Valve (FWRV) [LCV] and opened the valve, causing reactor level to increase.
At about 0941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br /> reactor level started to decrease rapidly and at the same time the 'A' FWRV locked up.
A reactor scram signal [JCI was received at 0942 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.58431e-4 months <br />, with indicated reactor water level at 20 inches.
All automatic actions occurred consistent with a low-level reactor scram.
C.
CAUSE OF EVENT
The root cause of this event has been determined to be an age-related failure of a solder joint on the master controller for the 'A' FWRV.
A contributing cause for this event was the fact that Operations procedures and training did not support reducing power to mitigate or prevent the reactor scram as an immediate operator action.
D.
SAFETY ANALYSIS
The safety significance of this event was minimal.
The failure of the 'A' FWRV caused reactor level to decrease below the scram setpoint.
After the scram reactor level returned to normal level.
Reactor water level did not go below 158.8" above the top of active fuel.
All safety systems responded as designed.
There were no safety system functional failures associated with this event.
This event was determined to be reportable under 10 CFR 50.73(a)(2)(iv) as an event that resulted in an automatic actuation of the Reactor Protection System.U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME 1 DOCKET NUMBER 2 LER NUMBER (6 PAGE 3 YEAR SEQUENTIAL REVISION Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2000 010 01 3 of 3 (If more space is required, use additional copies of NRC Form 366A)(17)
E.
CORRECTIVE ACTIONS
Immediate Actions An immediate inspection of the Feedwater Level Control system was conducted.
When the bad solder joint was located, it was repaired.
The other solder joints on the board were also inspected, and two additional joints were repaired as a result of that inspection.
Additionally, the spare controller in stock was visually inspected, with no discrepancies noted.
The Unit 2 controller was modified and inspected in February 2000 with no discrepancies noted.
At Quad Cities Nuclear Power Station this type of controller is used only in this system.
The annunciator procedure (QCAN 901(2)-5 F-8) and Abnormal Operating Procedure (QCOA 0201-09) associated with this event were revised to allow operators, as an immediate action, to reduce recirculation pump speed to maintain level.
Corrective Actions Completed:
Abnormal Operating Procedures were reviewed and changes made to move some subsequent operator actions to the immediate operator actions section of the procedures.
Training has been provided to all Operations crews emphasizing power reduction as a means of controlling level during a reactor level transient.
The Unit 2 Feedwater Level Control System has been replaced with a more reliable Digital Feedwater Control System.
Corrective Actions to be Completed:
The existing Feedwater Level Control System will be replaced with a more reliable Digital Feedwater Control System on Unit 1.
F.
PREVIOUS OCCURRENCES
There were no previous occurrences identified at Quad Cities Nuclear Power Station involving a scram due to failure of a solder joint.
G.
COMPONENT FAILURE DATA
The feedwater level master controller is a General Electric Controller, Part #50-548-OllFAK11.