05000315/LER-2007-001, Unit 1 Automatic Reactor Trip
| ML073050327 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/22/2007 |
| From: | Jensen J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2573-40 LER 07-001-00 | |
| Download: ML073050327 (6) | |
| 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 |
| 3152007001R00 - NRC Website | |
text
INDIANA MICHIGAN POWER A unit of American Electric Power Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 AEPcom October 22, 2007 Docket No. 50-315 U. S. Nuclear Regulatory Commission' Attn: Document Control Desk Mail Stop O-Pi-17 Washington, DC 20555-000.1 AEP:NRC:2573-40 10 CFR 50.73 10 CFR 50.4 Donald C. Cook Nuclear Plant Unit 1 LICENSEE EVENT REPORT 315/2007-001 -00 UNIT 1 AUTOMATIC REACTOR TRIP In accordance with the criteria established by 10 CFR 50.73, Licensee Event Report.System, the following report is being submitted:
LER 315/2007-001-00: "Unit. Automatic Reactor Trip" There are no commitments contained in this submittal.
Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.
Sincerely, v
?Jensen Site Vice President RAM/jen Attachment
U. S. Nuclear Regulatory Commission AEP:NRC:2573-40 Page 2 c:
J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachment INPO Records Center J. T. King, MPSC - w/o attachment MDEQ - WHMD/RPMWS - w/o attachment NRC Resident Inspector P. S. Tam - NRC Washington DC
NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES 6/3012007 (6-2004)
, 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 1 05000315 1 of 4
- 4. TITLE Unit 1 Automatic Reactor Trip
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 08 28 2007 2007 001 00 10 22 2007
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1E 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
E3 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[] 50.73(a)(2)(viii)(A) 0l 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[_ 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
E3 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
E] 50.36(c)(1)(ii)(A)
Z 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 100 [1 20.2203(a)(2)(iii)
El 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71 (a)(4)
El 20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[] 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[E 50.73(a)(2)(v)(C)
U 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 (If more space is required, use additional copies of NRC Form (366A)
Conditions Prior to Event
100 percent reactor power
Description of Event
On August 28,
- 2007, at 1354 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.15197e-4 months <br />, Donald C. Cook Nuclear Plant (CNP)
Unit 1 received a reactor trip and main turbine trip signal as a result of low Steam Generator #11 water level coincident with a steam flow -
feedwater flow mismatch.
All control rods [AA] fully inserted and the Auxiliary Feedwater System (AFW)
[BA]
started and performed as designed.
The reactor trip was uncomplicated and all major plant components functioned as designed.
The reactor trip was reported in accordance with 10 CFR 50.72(b) (2)(iv)(B) and the AFW actuation was reported in accordance with 10 CFR 50.72(b) (3) (iv)
(A).
The reactor trip and AFW actuation are reportable as a Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(iv)(A).
The initial' design of the Feed Pump Turbine Digital Control System (DCS)
[JJ] did not include an air conditioning system.
However, shortly after initial installation and operation of the DCS system, CNP recognized the need for and installed an air conditioning system to supply cool air to DCS cabinets.
The installed air conditioning system failed, causing temperatures to elevate in the DCS cabinet resulting in a DCS trouble alarm in the Control Room.
CNP personnel were troubleshooting the alarm and identified the DCS cabinet air conditioning had failed.
Actions were implemented to restore the cabinet's air conditioning at about the time the trip occurred and cabinet temperatures returned to normal shortly thereafter.
The DCS has two power supplies, which are designed to work in parallel and share the DCS power loads.
Each of the power supplies is capable of carrying the full load upon loss of the other power supply.
One of the two DCS power supplies had degraded and become overly temperature sensitive.
This degraded power supply, when exposed to elevated temperatures, produced an elevate~d output voltage.
The over voltage condition caused the alternate power supply to default to a zero output condition, per design, and the over voltage supply then carried the full load.
The degraded power supply output voltage increased to a point that the protective circuit on the East Main Feed Pump DCS controller tripped on high supply voltage level.
The trip of the controller caused the loss of the East Main Feed Pump.
This resulted in inadequate feedwater flow to the steam generators.
Subsequently, the reactor and main turbine tripped due to low Steam Generator #11 water level coincident with a steam flow -
feedwater flow mismatch.
(If more space is required, use additional copies of NRC Form (366A)
Cause of Event
The design of the DCS failed to ensure that DCS Power Supply Trip Setpoints were adequate to ensure that a degraded/failing power supply would not cause loss of downstream components, such as the feed pump control cards, prior to the power supply tripping and transferring the downstream loads to the alternate power supply.
In February 2007.,
CNP recognized that the DCS power supplies were temperature sensitive and that one power supply was degraded.
Cabinet air conditioning was installed to ensure the power supplies were maintained within an acceptable temperature band.
The failure of the DCS cabinet air conditioner in August 2007 resulted in elevated temperatures within the DCS cabinets, which in turn caused the degraded power supply to produce an over voltage condition.
This failure created an elevated output voltage from the power supply, which ultimately resulted in the tripping of the controller for the East Main Feed Pump.
This adverse impact on the Main Feed Pump controllers was not previously considered in the design or previous evaluation-of the degraded power supply.
Therefore, actions were not implemented to replace the power supply prior to the next scheduled refueling outage.
Analysis of Event
This event was consistent with the primary success path in the existing CNP risk analysis associated with unplanned reactor trips with the main condenser available.
There were no significant post-trip complications or failures.
This trip resulted from the loss of the Unit 1 East Main Feedwater Pump as a result of a short-term overheating condition in the DCS cabinets.
The three AFW pumps in the CNP Unit 1 AFW System, the cross-ties from the two Motor Driven AFW Pumps in the CNP Unit 2 AFW System, and the CNP Unit 1 West Main Feed Pump, along with the remainder of the Unit 1 Condensate System [SD],
were available to supply feedwater to the Unit 1 Steam Generators (SG) for heat removal.
Loss of the East Main Feedwater Pump causes no appreciable increase in CNP's estimated zero test and maintenance core damage frequency or large earlyrelease frequency.
This low risk significance is due to the equipment available to supply feedwater to a unit's SG in a post-trip condition.
Failure of the DCS cabinet air' conditioning and the subsequent reactor trip did not degrade systems that maintain core decay heat removal, assure containment integrity, or maintain defense-in-depth and safety margins.
Operators took procedurally-directed actions and responded to the transient in an appropriate and timely manner, resulting in a safe and stable plant configuration.
Automatic post-trip features functioned dependably.
For these reasons, this event is not considered to represent a significant risk to the plant or surrounding population.
(If more space is required, use additional copies of NRC Form (366A)
Corrective Actions
The DCS power supplies were replaced.
The air conditioning system was returned to operational status.
The evaluation of why CNP failed to recognize design limitations of the DCS system and failed to recognize system vulnerability associated with the degraded power supply are captured within the CNP corrective action program.
Previous Similar Events
The following LERs identify automatic reactor trips in the past three years.
The causes of these reactor trips were not similar in nature to the cause of this trip.
05000316/2004-001-00, Automatic Reactor Trip Due to RPS Actuation, While Manipulating Reactor Trip Bypass Breaker 05000316/2004-002-00, Unplanned Automatic Reactor Protection System Actuation Due to Feedwater Transient During a Power Reduction 05000316/2005-001-00, Reactor Trip from RCP Bus Undervoltage Signal Complicated by Diesel Generator Output Breaker Failure