05000348/LER-2004-001, Reactor Trip Due to Steam Generator Feedwater Pump Speed Control Failure
| ML041240203 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/29/2004 |
| From: | Stinson L Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-04-0702 LER 04-001-00 | |
| Download: ML041240203 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 3482004001R00 - NRC Website | |
text
L M. Stinson (Mike)
Vice President Southern Nuclear Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992:5181 Fax 205.992.0341 SOUTHERN A COMPANY Energy to Serve Your World' NL-04-0702 April 29, 2004 Docket No.: 50-348 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit I Licensee Event Report 2004-001 -00 Reactor Trip Due to Steam Generator Feedwater Pump Speed Control Failure Ladies and Gentlemen:
Joseph M. Farley Nuclear Plant - Licensee Event Report (LER) No. 2004-001 -00 is being submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A).
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, L. M. Stinson LMS/WAS/sdl Enclosure: Licensee Event Report 2004-001-00 cc:
Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. D. E. Grissette, General Manager - Plant Farley RTYPE: CFAO4.054; LC# 14021 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. S. E. Peters, NRR Project Manager - Farley Mr. C. A. Patterson, Senior Resident Inspector - Farley e','T I -z"-Z
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NR OM36US ULA EUAOYAPPROVEDOBY OMB NO. 3150-0104 EXPIRES 7-31.2004 NRC FORM 366 U.S. NUCLEAR REGULATORY
, the NRC may not conduct or sponsor. and a cers for ech bo person Is not required to respond to. the Information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Joseph M. Farley Nuclear Plant - Unit 1 05000 348 j
1 OF 4
TITLE (4) Reactor Trip Due to Steam Generator Feedwater Pump Speed Control Failure EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
FACILITY NAME DOCKET NUMBER SEQUENTIAL REV MO DAY YEAR YEAR NUMBER NO MO DAY YEAR 05000 3
O 2
00 4
_200 -
001
- - 001 04 29 2004 FACILITY NAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THEREQUIREMENTS OF 10 CFR §: (Check all that apply) (11)
MODE (9) 1 l 20.2201 (b) 20.2203(aX3)(ii) 50.73(aX2)(iiXB) 50.73(a)(2)(ix)(A)
POWER 100 20.2201(d)
=
20.2203(aX4) 0_
5.73(aX2)(iii) 50.73(a)(2)(x)
LEVEL (10) 0 20.2203(aX1) 50.36(cX1)(i)(A)
X 50.73(aX2)(iv)(A) 73.71 (a)(4) 20.2203(aX2)(i) 50.36(c)(1 XiiXA) 50.73(aX2)(v)(A) 73.71 (aX5) 20.2203(a)(2Xii) 50.36(c)(2) 50.73(a)(2)(vXB)
OTHER 20.2203(a)(2Xiii) 50.46(aX3)(ii) 50.73(a)(2Xv)(C)
Specify in Abstract below or in 20.2203(a)(2Xiv) 50.73(aX2)(i)(A) 150.73(a)(2Xv)(D)
N
=
20.2203(a)(2)(v) 50.73(a)(2Xi)(B) 50.73(a)(2)(vii)
=
20.2203(a)(2)(vi)
_ 50j73(a)(2XiXC) 50.73(a)(2)(viiiXA) 20.2203(aX3)(i) 50.73(a)(2Xii)(A) 50.73(aX2)(viiiXB)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
D. E. Grissette, General Manager Nuclear Plant__
334-899-5156 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
MANU-REPORTABLE l
MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX X
JB SIC W120 Yes l I
I I
I SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR SUBMISSION l
l l
YES (If yes, complete EXPECTED SUBMISSION DATE).
IX INO DATE (15) l_l l
ABSTRACT (Limit to 1400 spaces, Ie.. approximately 15 single-spaced typewritten lines) (16)
On March 1, 2004 at 0522, with the reactor at 100% power, Unit I automatically tripped due to Steam Generator IC level reaching its high level setpoint. At approximately 0521, the lead-lag card in the Steam Generator Feedwater Pump (SGFP) master speed control circuit failed, causing a ramp increase in the speed of both SGFPs. The speed increase caused an increase in feedwater flow to all three steam generators. As a result of the increase in SGFP speed, a low SGFP suction pressure condition occurred. Operators responded to the low suction pressure condition and the increasing feedwater flow by starting a stand-by condensate pump and manually closing the Main Feedwater Regulating Valves (MFRVs). The MFRVs responded to the transient, but the integrated reactor operator and system response was not rapid enough to prevent steam generator level reaching the high level setpoint. Per design, the main turbine, both SGFPs, and the reactor tripped. All safety systems functioned as designed.
This event was caused by failure of the Lead-Lag (NLL) card in the master SGFP speed control circuit, resulting in a ramp increase in SGFP speed. The unexpected SGFP low suction pressure alarm delayed the operators' diagnosis of the failure, thus delaying operator action which might have prevented the trip.
The failed lead-lag (NLL) card has been replaced. Annunciator response procedures have been revised to provide additional guidance for malfunctions of the feedwater control system.
NRC FORM 366 (72001J)
(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)
Additional Information
A program to address generic aging issues in the plant process control system (7300 system) is currently being implemented.
The following LERs have been submitted in recent years on Steam Generator Feedwater Pump associated events:
LER 2000-004-00 Unit 2 Reactor Trip Due to Degraded Main Feedwater Regulating Valve Transient Response.
LER 2002-004-00 Unit I Manual Reactor Trip on Loss of Both Steam Generator Feed Pumps.
LER 2003-003-00 Unit I Unplanned Auxiliary Feedwater Actuation upon Trip of Steam Generator Feed Pump.