05000483/LER-2008-006, Regarding Reactor Trip Due to Turbine Trip During Power Reduction Following Loss of C Condensate Pump
| ML090540058 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/06/2009 |
| From: | Neterer D AmerenUE, Union Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| ULNRC-05593 LER 08-006-00 | |
| Download: ML090540058 (9) | |
| 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4832008006R00 - NRC Website | |
text
AmerenUE Callaway Plant PO Box 620 Fulton, MO 65251 February 6, 2009 ULNRC-05593 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 WAmeren ff 1 OCFR50.73(a)(2)(iv)(A)
Ladies and Gentlemen:
DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.
FACILITY OPERATING LICENSE NPF-30 LICENSEE EVENT REPORT 2008-006-00 REACTOR TRIP DUE TO TURBINE TRIP DURING POWER REDUCTION FOLLOWING LOSS OF 'C' CONDENSATE PUMP The enclosed licensee event report is submitted in accordance with 10CFR50.73(a)(2)(iv)(A) to report an event of a reactor trip due to a turbine trip during power reduction following a loss of the 'C' condensate pump.
This letter does not contain new commitments.
Sincerely, David W. Neterer Plant Director KRA/nls Enclosure a subsidiary of Ameren Corporation
ULNRC-05593 February 6, 2009 Page 2 cc:
Mr. Elmo E. Collins, Jr.
Regional Administrator U.S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mohan C. Thadani (2 copies)
Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-8G14 Washington, DC 20555-2738
ULNRC-05593 February 6, 2009 Page 3 Index and send hardcopy to QA File A160.0761 Hardcopy:
Certrec Corporation 4200 South Hulen, Suite 422 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)
LEREvents@inpo.org (must send the WORD version of the LER to this address)
Electronic distribution for the following can be made via LER ULNRC Distribution:
A. C. Heflin F. M. Diya D. W. Neterer T. E. Herrmann L. S. Sandbothe S. A. Maglio S. L. Gallagher L. M. Belsky (NSRB)
T. B. Elwood D. E. Dumbacher (NRC)
B. A. Brook (WCNOC)
Ms. Diane M. Hooper (WCNOC)
Mr. Dennis Buschbaum (TXU)
Mr. Scott Bauer (Palo Verde)
Mr. Stan Ketelsen (PG&E)
Mr. Wayne Harrison (STPNOC)
Mr. John ONeill (Pillsbury Winthrop Shaw Pittman LLP)
Missouri Public Service Commission Records Center (INPO)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Callaway Plant Unit 1 05000483 1 OF 6
- 4. TITLE Reactor Trip Due To Turbine Trip During Power Reduction Following Loss Of 'C' Condensate Pump
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED RSEQENTIALuM I REV MO NT DA YR IFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEUMENILRE MO NT.A YA IFACILITY NAME DOCKET NUMBER 12 11 2008 2008 006 00 02 06 2009[
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
MODE 1 El 20.2201(d)
[3 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[1 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[E 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 17 20.2203(a)(2)(ii)
[] 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
[1 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
[E 50.73(a)(2)(v)(B)
[1 73.71(a)(5) 100%
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
[E OTHER El 20.2203(a)(2)(vi)
[E 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES At 2254 on December 11, 2008 the 'C' condensate pump [EIIS system: SD, component: P] tripped due to a motor [EIIS system: SD, component: MO] ground fault failure (turn-to-turn failure on the stator). At 2255 the operating crew noticed computer point MAJO001, Main Generator Power, had turned yellow and was indicating approximately 1314 MWe. The operators immediately noticed that the 'C' condensate pump had tripped, commenced a reduction in power, and entered Callaway off-normal procedure OTO-AE-00001, Feedwater System Malfunction. However, the manual load reduction rate was approximately 31% in one minute, exceeding the procedural guidance of 5% per minute. At 2256 the control rods [ElIS system: AA, component: ROD] automatically stepped in and the condenser steam dumps [EIIS system: SB, component: RV] opened as designed. Level transients subsequently occurred in the steam generators (S/Gs) [EIIS system: AB, component: SG], and level deviation annunciators actuated in the main control room for all four S/Gs. As a result of high level in the 'B' S/G, a turbine [EIIS system: TA, component: TRB] trip/reactor trip occurred at 2300.
The turbine tripped on a Hi-Hi S/G level (P-14) feedwater isolation signal (FWIS) [EIIS system: SJ, component: ISV] from the 'B' S/G, and the reactor trip immediately followed. Additionally, as a result of the P-14 FWIS, both main feedwater pumps (MFPs) [EIIS system: SJ, component: P] tripped and hence a motor-driven auxiliary feedwater actuation [EIIS system: BA, component: P] occurred as expected. The Auxiliary Building sample station Chemical and Volume Control System (CVCS)
Letdown Gamma Detector [EIIS system: IL, component: DET] went into Alert as a result of the transient (due to a pre-existing fuel defect) and a Loose Parts Monitor [EIIS system: II, component: MON]
alarmed, as expected, due to the control rods dropping. The 'B' and 'C' atmospheric steam dumps (ASDs) [EIIS system: SB, component: RV] opened and re-closed at the time of the reactor trip.
Two root cause investigations were performed to determine causes and corrective actions for the pump motor failure and for the automatic actuations of the Main Feedwater Isolation, RPS & Auxiliary Feedwater System (AFWS). The pump motor failure which initiated the sequence of events was the result of one root cause associated with two causal factors. The investigation for the automatic (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 (If more space is required, use additional copies of NRC Form 366A) (17)
One event in the Callaway corrective action program documents an Institute of Nuclear Power Operations (INPO) Area for Improvement, from 2002, in which "Operations does not set and reinforce high standards in some activities. Specifically, Operations lags the industry in implementing tools to improve human performance." The CATPRs were to develop a 6-year plan for the performance of Self-Assessments and Benchmarking trips as well as conducting evaluations periodically with each crew to focus on maintaining and reinforcing high standards.
The 6-year plan has only gone through a single cycle and may not have had time to be effective.
V.
ADDITIONAL INFORMATION
The system and component codes listed below are from the IEEE Standard 805-1984 and IEEE Standard 803A-1983, respectively.
System:
AA Control Rod Drive System AB Reactor Coolant System (PWR)
BA Auxiliary/Emergency Feedwater System (PWR)
II Loose Parts Monitoring System IL Radiation Monitoring System SB Main/Reheat Steam System (PWR)
SD Condensate System SJ Feedwater System TA Main Turbine System Component:
DET Detector ISV Valve, Isolation MO Motor MON Monitor P
Pump ROD Rod RV Valve, Relief SG Generator, Steam TRB Turbine