05000424/LER-2001-001-01, Re Reactor Trip Due to Loss of Generator Excitation
| ML020930168 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/28/2002 |
| From: | Beasley J Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LCV-1566-A LER-01-001-01 | |
| Download: ML020930168 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4242001001R01 - NRC Website | |
text
J. Barnie Beasley, Jr., P.E.
Vice President Southern Nuclear Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7110 Fax 205.992.0403 4
SOUTHERNAZ COMPANY E~nergy to Serve Your Wo rid SM March 28, 2002 LCV-1566-A Docket No.: 50-424 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT LICENSEE EVENT REPORT 1-01-001, REV. 1 REACTOR TRIP DUE TO LOSS OF GENERATOR EXCITATION In accordance with the requirements of 10 CFR 50.73, Southern Nuclear Operating Company hereby submits a revision to Vogtle Electric Generating Plant licensee event report for a condition that occurred on Unit 1 on August 24, 2001.
Sincerely, B.
sley, i f
JBB/BHW Enclosure: LER 1-2001-001, REV. 1 cc:
Southern Nuclear Operating Company Mr. J. T. Gasser Mr. M. Sheibani SNC Document Management U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. Frank Rinaldi, Vogtle Project Manager, NRR Mr. J. Zeiler, Senior Resident Inspector, VEGP
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NRC FORM 366 U.S.NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 EXPIRES: 06f30=2001 (6-1998)
.Estimated burden per responae to comply with this mandatory information request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 LICENSEE EVENT REPORT (LER)
F33), U.S. Nucear Regulatory Commission. Washington. DC 20555-(See reverse for required number of 0001, and to the Paperwork Reduction Project (3150-0104), Office of digits/characters for each block)
Management and Budget, Washington, DC 20503.
If an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to. the information collection.
CKET NUMBER (2)
PAGE (3)
Vogtle Electric Generating Plant - Unit 1 0 5 0 0 0 4 2 4 1 OF 4
TITLE (4)
REACTOR TRIP DUE TO LOSS OF GENERATOR EXCITATION EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MONTH DAYl YEAR YEAR I SEQUENTIAL I REVISION MONTH YEAR FACILITY NAME DOCKET NUMBER NUMBER55 NŽUMBER MOT DAFAIIYN EDCKTUBR 0 5 000 FACILITY NAME 0 81241200112001 0 0 110 0
0 0 l
5 0 0 ° OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR*: (Check one or more)
(11)
MODE (9) 1 J
20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)
POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)
LEVEL (10) 1 0
0 ll 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.7t 20.2203(a)(2)(ii) 20.2203(a)(4)
X 50.73(a)(2)(iv)
OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v)
Specify in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2)
T50.73(a)(2)(vii) or in NRC Form 366A LICENSEE CONTACT FOR THIS LER (12) l NAME ELEPHONE NUMBER (indude area code)
Mehdi Sheibani, Nuclear Safety and Compliance 7 0 6 -
8 2 6 3 2 0 91 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) l
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE
CAUSE
I SYSTEM COMPONENT MANUFACTURER REPORTABLE TOTSCR GP8X TO EPIX B _T L S C R G 0 8 0 A..<
f SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR SUBMISSION i
YES (If yes, complete EXPECTED SUBMISSION DATE)
NO DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-space typewritten lines) (16)
On August 24, 2001, with the unit at 100% power, personnel were returning to service the Main Generator Rectifier Bridge 1 after performing corrective maintenance. Upon closure of the disconnect switch, a trip of the main generator occurred on loss of field excitation, causing an automatic turbine/reactor trip at 2307 EDT. All control rods were observed to fully insert, and control room operators acted properly to control steam generator water levels and stabilize the unit in mode 3 (hot standby).
An investigation found that, upon returning Rectifier Bridge 1 to service, silicon controlled rectifiers (SCRs) in Rectifier Bridge 4 failed due to a short circuit. This led directly to the loss of the generator excitation field and the generator/turbine/reactor trip. The malfunctioning rectifier bridges were repaired and the unit was returned to service. The failed SCRs that were removed from the rectifier bridges were sent to a laboratory for failure analysis. The analysis was unable to determine the root cause of the failure.
NR eor SE on of~
NRC F-ormr 366 (6-1Vt9t)U.S.NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SECUENTIAL REVIS IN NUMBER NUMBER VoptleElectricGeneratingPlant -Unit 1 0 5 0 0 0 4 2 4 2 0 0 1 0 0 1 0
1 2 OF 4
TEXT (If more space is required, use additional copies of NRC Form 366A)(17)
A. REQUIREMENT FOR REPORT This report is required per 10 CFR 50.73 (a)(2)(iv) because an unplanned actuation of the reactor protection system occurred.
B. UNIT STATUS AT TIME OF EVENT At the time of this event, Unit 1 was operating in Mode 1 (power operation) at 100% of rated thermal power. The generator was operating with Rectifier Bridge 1 out of service and Rectifier Bridges 2, 3 and 4 in service.
C. DESCRIPTION OF EVENT
Following a lightning strike on the unit's 230 kV line on August 18, 2001, anomalies were found on phase A silicon controlled rectifiers (SCRs) of the main generator Rectifier Bridge 1 in the form of failed LEDs and resistors. Upon removing Rectifier Bridge 1 from service for repairs, an anomaly (flickering LEDs) began to occur on phase C of Rectifier Bridge 2 due to a voltage imbalance.
Following consultation with the generator vendor, it was decided to repair Bridge 1, return it to service, then remove Bridge 2 from service for repair.
On August 24, 2001, after repairing Rectifier Bridge 1, personnel were placing the bridge back in service in accordance with procedure 13830- 1, "Main Generator Operation." Upon closure of the disconnect switch for Rectifier Bridge 1, a trip of the main generator occurred on loss of field excitation, causing an automatic turbine/reactor trip at 2307 EDT. The first-out annuciator seen by operators in the control room was "Turbine Trip/P-9 Reactor Trip." All control rods were observed to fully insert, and a main feedwater system isolation (FWI) and an auxiliary feedwater system (AFW) actuation occurred as expected. Control room operators acted properly to control steam generator water levels and stabilize the unit in mode 3 (hot standby).
D. CAUSE OF EVENT
An investigation found that, upon returning Rectifier Bridge 1 to service, both phase A SCRs failed in Bridge 4 due to a short circuit. This led directly to the loss of generator excitation field and the generator/turbine/reactor trip. Several possible reasons for the short circuit were proposed, and the failed SCRs were sent to a laboratory for failure analysis. The analysis determined that overheatingU.S.NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER Vogtle Electric Generatina Plant - Unit 1 05000424 2 0 0 1 0 0 1 0
1 3 OF 4
TEXT (If more space is required, use additional copies of NRC Form 366A)(17) of the SCRs was the cause of the failures. Although possible causes for the overheating have been identified, the root cause of the overheating, and of this event, cannot be determined with certainty.
E. ANALYSIS OF EVENT
The reactor protection system, the main feedwater isolation function, and the auxiliary feedwater actuation function performed as designed. Control room operators acted properly to control steam generator water levels and stabilize the unit in mode 3 (hot standby). Based on these considerations, there was no adverse effect on plant safety or on the health and safety of the public as a result of this event.
This event does not represent a safety system functional failure.
F. CORRECTIVE ACTIONS
- 1) Rectifier bridges were repaired and the unit was returned to service.
- 2) Based on new information received from the generator vendor, operating and maintenance procedures have been revised and generator training for appropriate engineering and maintenance personnel is scheduled for later this year.
- 3) The failed SCRs were sent to a testing laboratory for failure analysis. The analysis determined that overheating of the SCRs was the cause of the the failures. Although possible causes for the overheating have been identified, the root cause of the overheating, and of this event, cannot be determined with certainty.
G. ADDITIONAL INFORMATION
- 1) Failed Components:
Silicon Controlled Rectifiers manufactured by General Electric Corporation Part # 44C338754G01
- 2) Previous Similar Events:U.S.NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER Voetle Electric Generating Plant - Unit 1 0°50°00 4 242 0 01 _0 01 _0 1
40OF 4
TEXT (If more space is required, use additional copies of NRC Form 366A)(17)
LER 5000425/1991-007-00, dated June 4, 1991. This LER described a generator/turbine/reactor trip caused by problems with the generator control circuits.
- 3) Energy Industry Identification System Code:
Main Generator System - TB Main Generator Excitation System - TL Reactor Control System - JD Main Feedwater System - SJ Auxiliary Feedwater System - BA