05000313/LER-2003-001, Unit 1 Reactor Trip Due to Automatic Actuation of the Reactor Protection System on High Reactor Coolant System Pressure and Actuation of the Emergency Feedwater System Resulting from a Lightning-Induced Closure
| ML033030398 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/28/2003 |
| From: | Ashley G Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN100301 LER 03-001-00 | |
| Download: ML033030398 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3132003001R00 - NRC Website | |
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Et-egy Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 501 858 5000 1CAN 100301 October 28, 2003 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station OPl-17 Washington, DC 20555
Subject:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Licensee Event Report 50-313/2003-001-00
Dear Sir or Madam:
In accordance with 10CFR50.73(a)(2)(iv)(A), enclosed is the subject report concerning an automatic actuation of the Reactor Protection and Emergency Feedwater Systems.
The enclosure contains no commitments.
Sincerely, Glenn R. Ashley Manager, Licensing GRA/dh enclosure 1-
1CAN100301 Page 2 cc:
Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, GA 30339-5957 LEREvents@inpo.org
NRC -ORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1-2001)
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Nuclear Regulatory Commission, Washington. DC 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0104), Office of Management and Budget.
Washington. DC 20503.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Arkansas Nuclear One - Unit 1 05000313 1 OF 4 TITLE(4) Reactor Trip due to Automatic Actuation of the Reactor Protection System on High Reactor Coolant System Pressure and Actuation of the Emergency Feedwater System Resulting From a Lightning-Induced Closure of the Main Turbine Governor Valves EVENT DATE (5)
LER NUMBER MO DAY YEAR
+ YEAR SEQUENTIAL NUMBER 08 29 2003 2003 001 OPERATING MODE (9)
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR: (Check one or more) 11) 1 20.2201 (b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)
POWER 1
20.2201 (d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A)
LEVEL (10) 100 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1)(fi)(A)
X 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(ii)
=
50.36(c)(2)
=
50.73(a)(2)(v)(A) 73.71 (a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)ii)
=
50.73(a)(2)(v)(B) 73.71 (a)(5) 20.2203(a)(2)(v) 50.73(a)(2)Ci)(A)
=
50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi)
=
50.73(a)(2)(i)(B)
=
50.73(a)(2)(v)(D)
Specify in Abstract or (17)
A.
Plant Status At the time this event occurred, Arkansas Nuclear One, Unit 1 (ANO-1) was operating in Mode 1 at approximately 100 percent power.
The ANO-1 area was experiencing inclement weather in the form of thunderstorm activity.
B.
Event Description
At 1428 CDT on August 29, 2003, ANO-1 experienced an automatic actuation of the Reactor Protection System (RPS)[JC] due to high Reactor Coolant System (RCS)[AB) pressure.
The RCS high pressure condition was caused by the closure of the main turbine governor valves.
The closure of these valves was likely due to a lightning-induced Electro Hydraulic Control (EHC)[TG] System malfunction.
All control rods fully inserted and the unit was in stable Hot Standby conditions (Mode 3) shortly after the reactor trip.
Following the reactor trip, the main generator output breakers failed to open automatically due to failure of both the primary and secondary reverse power relays to trip.
The output breakers were manually opened approximately 1.5 minutes after the reactor trip in accordance with post-trip procedures.
The main turbine tripped as a result of the reactor trip.
Main Feedwater Block Valve CV-2675 to the "B" Once-Through Steam Generator (OTSG)[AB] failed to fully close resulting in erratic control of feedwater flow to the steam generator.
The
B" Main Feedwater (MFW)[SJJ Pump was placed in manual in an attempt to control steam generator level.
When steam generator level reached the automatic actuation setpoint, the Emergency Feedwater Initiation and Control (EFIC)[JB] System actuated as expected and initiated Emergency Feedwater (EFW)[BA] flow to the affected steam generator.
EFW flow quickly restored the steam generator level to setpoint.
Both MFW pumps remained operable throughout the transient and MFW flow was never lost.
The EFW System recirculated for approximately two hours.
Valve CV-2675 was manually closed at 1713.
The Auxiliary Feedwater [SJ] Pump was placed in service and the EFW and MFW pumps were secured.
Fifteen main steam safety valves (MSSVs)[SB] lifted.
All valves performed satisfactorily.
MSSV PSV-2697 did not appear to lift.
The highest pressure indicated for the PSV-2697 steam header was less than the +3 percent setpoint margin for this valve; therefore, the performance of PSV-2697 was determined to be acceptable.
The Reactor Protection and Emergency Feedwater Systems performed as designed.
No technical specification limits were exceeded.
(17)
C.
Root Cause The reactor trip was initiated by an electrical perturbation that caused a malfunction of the Main Turbine EHC System.
The lightning-induced electrical perturbation simulated an "output breaker open" permissive causing the EHC System to change control states briefly.
The change in the control signal resulted in closure of the main turbine governor valves.
The main generator output breakers failed to open automatically due to failure of both the primary and secondary reverse power relays to trip.
The primary reverse power relay is set to actuate three seconds after sensing a reverse power of greater than 2MWe.
If the primary relay fails to operate, the backup reverse power relay should actuate in 30 seconds.
The primary relay failed due to binding.
The secondary relay was found to be nonfunctional due to a flaw in the control logic that deenergized the relay before it could actuate.
In addition, Main Feedwater Block Valve CV-2675 did not fully close due to the failure of its valve operator.
CV-2675 is an 18-inch gate valve with a two-speed motor operator powered by 480 volt AC.
The speed change in the operator, accomplished by the use of high-speed and slow-speed clutches, gives the motor operator the ability to close the valve at high speed, but protects the valve from seating damage by slowing during the last 5 percent of travel.
The high-speed and slow-speed clutches are DC powered. To obtain the required DC supply from the 480 volt source, the motor operator has an integral full-wave bridge rectifier, followed by fly-back capacitors.
Troubleshooting revealed that two of the four diodes of the full-wave rectifier were damaged.
This condition degraded the DC supply to the clutches and prevented clutch operation.
The damage to the rectifier diodes was likely caused by a bad connection on one of the fly-back capacitors.
This allowed the collapsing magnetic field from the high-speed clutch to be impressed across the diode bridge.
D.
Corrective Actions
Following replacement of the damaged components, testing was performed to verify operability of Main Feedwater Block Valve CV-2675.
The primary reverse power relay for the main generator output breaker was replaced with a new unit, eliminating the binding condition that was found on the failed relay.
The main generator output breaker was tested to verify operability of the system.
The trip logic of the backup reverse power relay will be corrected during the next refueling outage scheduled for the (17) spring of 2004.
Operations personnel have been provided with contingency actions during the interim period of plant operation.
A functional test of the EHC System was conducted and no problems were identified.
The system's power supplies were tested and found satisfactory.
Improvements to the EHC System are being developed to provide circuit isolation for the main generator output breaker contact to prevent false breaker indication due to lightning.
E.
Safety Significance
The RPS initiated a reactor trip upon reaching the high RCS pressure setpoint as designed.
The plant response to the transient was complicated by the failure of a main feedwater block valve to fully close.
The EFIC System actuated as a result of low steam generator level and initiated EFW flow to the affected steam generator.
Although this event was complicated by equipment malfunctions, the unit was returned to normal post-trip parameters shortly after the trip. Therefore, this event is considered to be of low
safety significance
F.
Basis for Reportability Automatic actuations of the Reactor Protection System and Emergency Feedwater System are being reported in accordance with 10CFR50.73(a)(2)(iv)(A).
An initial report of this event was made to the NRC Operations Center at 1559 CDT on August 29, 2003, in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A).
Supplemental notifications were made to the NRC Operations Center at 0044 CDT and 1815 CDT on August 30, 2003.
G.
Additional Information
Although there have been previous lightning-induced plant transients, the only previous similar event was reported in LER 50-313/86-004-00 (letter 1CAN058604 dated May 8, 1986).
The valve operator for Main Feedwater Block Valve CV-2675, model SMB-4, was manufactured by Limitorque Corporation (MFR Code L200).
The main generator output breaker primary reverse power relay, model 12GGP53C1A, was manufactured by General Electric Company (MFR Code G080).
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
NRC FORM 3e6A (1 2001)
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