05000316/LER-2013-001, Regarding Unit 2 Manual Reactor Trip Due to Lowering Steam Generator Level
| ML13269A367 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/24/2013 |
| From: | Gebbie J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2013-77 LER 13-001-00 | |
| Download: ML13269A367 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(iv)(A), System Actuation 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 |
| 3162013001R00 - NRC Website | |
text
INDIANA MICHIGAN POWER A unit of American Electric Power Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 IndianaMichiganPower.com September 24, 2013 AEP-NRC-2013-77 10 CFR 50.73 Docket No.: 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike, Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2013-001-00 Unit 2 Manual Reactor Trip In accordance with 10 CFR 50.73, Indiana Michigan Power Company, the licensee for Donald C.
Cook Nuclear Plant Unit 2, is submitting the following report as an enclosure to this letter:
LER 316/2013-001-00: "Unit 2 Manual Reactor Trip" There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely, Joel P. Gebbie Site Vice President SJM/amp Enclosure c:
J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosure MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III T. J. Wengert, NRC Washington, DC rg
NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Donald C. Cook Nuclear Plant Unit 2 05000-316 1 of 4
- 4. TITLE Unit 2 Manual Reactor Trip due to Lowering Steam Generator Level
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV FACILITY NAME DOCKET NUMBER NUMBER NO.
1 05000 FACILITY NAME DOCKET NUMBER 07 28 2013 2013 001 00 09 24 2013 1 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[E 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
[1E 20.2201(d)
El 20.2203(a)(3)(ii)
[1 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B)
[E 50.73(a)(2)(viii)(B)
[1 20.2203(a)(2)(i)
[E 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [I
20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100 El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
H 73.71(a)(5)
El 20.2203(a)(2)(v)
E]
50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)S i
r OTHER Specify in Abstract below El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D) or in
Cause of Event
The initiating event was a failure of air operated valve 2-MRV-411 failing to the closed position due to fretting of the control air signal air line by a tubing support bracket.
The Root Cause of the event was determined to be a lack of procedural guidance and adequate controls for the Condensate Heater Condensate Bypass Control Valve, 2-CRV-224 [SD][V], automatic control functions.
Analysis of Event
The event was analyzed regarding the contributing factors leading up to the manual reactor trip on lowering steam generator levels caused by the low suction pressure automatic trip of the West Main Feedwater pump.
Investigation following the event concluded that 2-CRV-224 [SD][V] did not open to help prevent a low feedwater suction pressure from tripping the main feedwater pump. The associated controller [SD][PMC]
was found at a setpoint lower than the intended design, which prevented 2-CRV-224 from opening.
Controller upgrades occurred in 1995 and 2003 that changed the methodology for setpoint inputs. The root cause was stated as a lack of procedural guidance and controls for the valve controller functions following controller upgrades.
A failure of 2-LPD-320N, North Heater Drain Pump PP-22N Discharge Check Valve [SN][CKV], was discovered following the manual reactor trip. The check valve was found in a partially open position which allowed a diversion of a portion of condensate flow away from the main feedwater pump suction.
The TDAFP discharge valve that did not automatically position as expected following the manual reactor trip was found with limit switches adjusted incorrectly. The incorrect limit switch settings did not affect the main feedwater pump suction during the event or impact feedwater flow to #3 steam generator. The limit switches were adjusted and tested satisfactory.
A risk impact review for this event indicates the trip was uncomplicated. All control rods [JC][JD] inserted, the main turbine-generator [TB][TG] tripped, offsite power transferred and remained energized by the station Reserve Auxiliary Transformers [EB][XFMR] as designed. No safety injection or engineered safety feature actuations occurred or were warranted beyond those expected for a nominal reactor trip. Main feedwater function was recoverable if it had been needed. The control room operators did not enter any additional emergency operating procedures after the trip, except those optimally expected. The operators noted that 2-FMO-221 (TDAFP Discharge to #3 SG flow control valve) ran fully closed rather than stopping at its intermediate flow retention setting. Discussion with the control room operators afterward and review of the Reactor Trip Report indicates there was sufficient flow from the motor driven AFW pumps such that the closed AFW valve had no significant impact on plant trip response. The operators were not required to manually reopen valve 2-FMO-221 immediately following the trip. The operators subsequently opened 2-FMO-221 to re-align the TDAFP to an available status and the valve did respond as expected to manual operation. The operators could have opened the valve during the event for additional AFW flow to #3 SG if needed.
Overall, the trip was an uncomplicated event with a malfunction of one AFW valve. The crew had observed and was aware of the AFW valve condition, and could have manually used the valve if needed. For these reasons, this trip did not pose any significant risk.
Corrective Actions
Completed Corrective Actions Plant procedures were revised to provide guidance to establish and maintain condensate bypass valve 2-CRV-224 controller setpoint to modulate beginning at 240 psig.
North Heater Drain Pump discharge check valve 2-LPD-320N was replaced.
Failed control air tubing supply line to 2-MRV-411 was replaced.
The motor operated valve actuator limit switch settings for 2-FMO-221 were adjusted and tested.
Planned Corrective Actions
No additional corrective actions are planned.
Previous Similar Events
LERs for CNP Unit 1 and Unit 2 were reviewed for the previous five years and found no similar events.