05000316/LER-2003-005, Regarding Reactor Trip Due to Steam Flow/Feed Flow Mismatch
| ML040550411 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/17/2004 |
| From: | Jensen J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2573-18 LER 03-005-00 | |
| Download: ML040550411 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
| 3162003005R00 - NRC Website | |
text
'Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER February 17, 2004 AEP:NRC:2573-18 10 CFR 50.73 Docket No.
50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 2 UNIT 2 REACTOR TRIP DUE TO STEAM FLOW/FEED FLOW MISMATCH In accordance with 10 CFR 50.73, "Licensee Event Report System," the following report is submitted:
Licensee Event Report (LER) 316/2003-005-00: "Unit 2 Reactor Trip Due to Steam Flow/Feed Flow Mismatch." identifies the commitments included in this submittal.
Should you have any questions regarding this correspondence, please contact Mr. Toby K. Woods, Compliance Supervisor, at (269) 466-2798.
Sincerely, Joseph N. Jensen Site Vice President RAM/jen Attachments
C U. S. Nuclear Regulatory Commission AEP:NRC:2573-1B Page 2 C:
J. L. Caidwell - NRC Region III K. D. Curry - AEP Ft. Wayne J.T.King-MPSC MDEQ - WHMDi'HWRPS NRC Resident Inspector Records Center - INFO J. F. Stang, Jr. - NRC Washington DC
AT[ACHMENT I TO AEP:NRC:2573-18 REGULATORY COMMITMENTS The following table identifies those actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.
Commitment
Date I&M will incorporate the enhanced standards for disconnecting and connecting leads into procedure 12-IIIP-5021-IMP-OO1 March 19, 2004 (CRA 03365009-03).
I&M will incorporate the enhanced standards for disconnecting and connecting leads into procedure PMP-2291-PLN-OO1 March 19, 2004 (CRA 03365009-04).
'NIC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150.0104 EXPIRES 7-31-2004 (7-2001)
Estimated briden p response to cnpty citi lis rnandatoy kVaination cotleclon request: 50 hcoxs. Reputed lessons leaned are Incorporated Into tie ticenang process and fed back to Industiy. Send comnients regarrtng LICENSEE EVENT REPORT (LER) tuden esdmale to lie Records Management Branch (T.6 EG). 11.5. Nucteor Reatory Cormusmon.
20555-0001. or by Internet emat to bjs1nrc.gov. aid to tie Desk Ofticer, Ofdce of Information aid ReWlatory Altars. NEOB-l 0202(3150-0104). Oftice of Management and Budget. Wastington. DC 20503. tf (See reverse for required num5r of a means used to impose Information cofeciwi 6ses not isday a crxrerrfy valid 0MB conbot razrber, tie NRC dlgrtslcharacters for each block) may not conduct or sponsor. and a person is not req.ired to respond to, the Information coltectisa
- 1. FACIUTY NAME
- 2. DOCKET NUMBER
- 3. PAGE Donald C. Cook Nuclear Plant Unit 2 05000-316 I of 4
- 4. TITLE Unit 2 Reactor Trip Due To Steam Flow/Feed Flow Mismatch
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACIUTIES INVOLVED SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 12 30 2003 2003 005 00 02 17 2004
- 9. OPERATING MODE 1
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR Check all that a I
__ 20.2201(b) 20.2203(a)(3)(ii) 50.73(a)(2Wti)(B)
__ 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 100
- - 20.2201(d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x) 20.2203(afll) 50.36(c)(1)(iflA)
X 50.73(a)(2)Ctv)(A)
__ 73.71(a)(4) 20 2203(a)(2)(l)
__ 50.36(c(1 )(iiflA) 50.73(a)(2)(vflA)
__ 73.71 (afl5) 20 2203(a)(2)Oi) 50.36(c)t2) 50.73(a)(2)(v)(B)
OTHER Specify In Abstract below 20 2203(a)(2)(iii) 50.46(a)(3)Oi) 50.73(aX2)(vXC) or In
Abstract
On December 30, 2003, at approximately 1328 Eastern Standard Time, the Unit 2 reactor automatically tripped on low steam generator water level coincident with feed flow less than Steam flow for Steam Generator #22. Just prior to the trip, the Control Room Instrument Distribution (CRID) IV inverter transferred to the alternate power supply. A Feedwater Isolation Relay (K666X2), powered from CRID IV, opened. The opening of the relay initiated a seal-in closure signal to SteamGenerator #22 and #23 Feedwater Isolation Valves (2-FMO-202 and 2-FMO-203), resulting in a loss of feedwater flow to Steam Generators #22 and #23 and produced a low steam generator water level trip coincident with feed flow less than Steam flow. Following the reactor trip, the Unit Output Breakers failed to automatically open because a Main Turbine Stop Valve position indicator failed to provide a closed indication. The Unit Output Breakers were manually opened in accordance with the reactor trip response procedure. All safety components started and performed as expected. The initiator of the reactor trip was determined to be a momentary ground on the CRID IV bus, which occurred during the conduct of instrument calibration. The ground caused the automatic transfer of CRID IV to the alternate power source. The Voltage drop on the CRID IV bus Caused the Feedwater Isolation Relay to open. The root cause of the event has been determined to be inattention on the part of the Instrument and Control Technician who was performing the calibration. Corrective actions include the establishment of enhanced controls for the disconnecting and connecting of electrical leads.
NRC FORM 366 (7-2001)U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
- LICENSEE EVENT REPORT (LER);
TEXT CONTINUATION
- 1. FACIUTY NA1IIE
- 2. DOCKET NUMBER
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Donald C. Cook Nuclear Plant Unit 2 05000-316 NUMBER NUMBER 2 of4 2003 005 00
- 17. TEXT (If more space Is required, use additionalcopies of NRC Form (366A)
Conditions Prior to Event
Unit I = 100% Power Unit 2-100% Power
Description of Event
On December 30, 2003, at approximately 1328 EST, the Unit 2 reactor automatically tripped on low steam generator water level coincident with feed flow less than steam flow for Steam Generator #22.
Just prior to the trip, Instrument and Control Technicians were performing a calibration of 2-IFC-325 [BP - FCVJ, West Residual Heat Removal Pump (PP-35W) Discharge Flow Switch. While connecting a lead, a momentary ground occurred at the terminal box. The resulting ground fault current and subsequent inrush current caused the Control Room Instrument Distribution (CR10) IV [ED - INVT] inverter to transfer to the alternate power supply. The voltage transient on the CRID IV bus was enough to open the Feedwater Isolation Relay (K666X2). The opening of the relay initiated a seal-in closure signal to Steam Generator #22 and #23 Feedwater Isolation Valves (2-FMO-202 and 2-FMO-203) [ISV]. Control Room personnel observed the CR10 IV abnormal annunciator and Steam Generator water level alarms and immediately implemented actions to respond to the alarms. An automatic reactor trip occurred due to a low steam generator level coincident with feed flow less than steam flow.
Following the reactor trip, the Unit 2 Output Breakers failed to automatically open due a Main Turbine Stop Valve [SB] position indicator failing to provide a closed indication. The Unit Output Breakers were manually opened in accordance with the reactor trip response procedure. All safety components started and performed as expected.
At 1501 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.711305e-4 months <br /> on December 30, 2003, Indiana Michigan Power Company (l&M) made a four-hour notification to the NRC Operations Center, in accordance with 10 CFR 50.72(b)(2)(iv)(B) to report the actuation of the reactor protection system and resultant reactor scram.
Cause of Event
The cause of the Unit 2 automatic reactor trip was the loss of feedwater flow to Steam Generator #22. The cause of the loss of feedwater flow was the closure of the Feedwater Isolation Valve when the Feedwater Isolation Relay opened following a momentary electrical disturbance on CR10 IV. The electrical disturbance was caused by a momentary ground fault created while connecting a lead during the calibration of 2-IFC-325, which is powered by CR10 IV.
The root cause of the Unit 2 trip was inattention on the part of the Instrument & Control Technician performing the connection of live leads in the terminal box for 2-IFC-325. The technician brought the lead in close proximity to the terminal box edge and caused a ground fault. The current drawn as a result of the fault caused a voltage transient resulting in an unanticipated voltage drop below the design opening value of relay 2-K666-X2-B. The opening of this relay sent a close signal to the Feedwater Isolation Valves, which subsequently resulted in the reactor trip.U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Donald C. Cook Nuclear Plant Unit 2 05000-316 NUMBER NUMBER 3 of4 2003
-
005 00
- 17. TEXT (If more space Is required, use additional copies of NRC Form (366A)
Analysis of Event
An assessment of this event was performed and it was determined that this event was bounded by the existing accident analysis associated with unplanned reactor trips with the main condenser available. The change in risk with respect to core damage and large early release frequency due to the inadvertent CRID inverter automatic transfer, together with the associated inadvertent closure of the Feedwater Isolation Valves and subsequent plant trip, have been qualitatively assessed and judged no different than any other Unplanned reactor trips with the main condenser available. This assessment is based on the following considerations:
The automatic plant trip, due to low main feedwater flow, functioned properly. The automatic trip features also functioned dependably, with the exception of a failed Main Turbine Stop Valve limit switch, which resulted in failure of the Unit Output Breaker to automatically open. However, the operators responded in an appropriate and timely manner, resulting in a safe and stable plant configuration.
The inadvertent CRID inverter automatic transfer did not contribute to the increased likelihood of any initiators, other than transients that result in or from a reactor trip.
Neither the CRID inverter automatic transfer nor the associated inadvertent closure of the Feedwater Isolation Valves degraded any system used to prevent core damage, assure containment integrity, or maintain defense-in-depth and safety margins.
Accordingly, l&M concludes that there was no impact on the health and safety of the public as a result of this event.
Corrective Actions
Plant Eauioment:
Verification of proper operation of the CRID.
Replacement of Main Turbine Stop Valve position indicator and verification of its proper operation.
Immediate Actions:
An event specific causal evaluation was performed.
- A human performance root cause was performed for this human error event as well as other recent similar human performance events within the Maintenance Department.
As a result of the unacceptable human performance trend associated with maintenance work activities, a Stop Work Order was issued effective immediately to the Maintenance Department for all maintenance activities.
This Stop Work Order is documented in Condition Report 04009044. The Stop Work Order will be rescinded when permanent preventive actions have been put into place to address barrier failures that occurred leading up to this event. The approval of the Plant Manager and the Site Vice President or Chief Nuclear Officer is required for rescinding this Stop Work Order.
Enhanced standards were established for disconnecting and connecting leads.U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LiCENSEE EVENT REPORT (LER)
TEXT CONTINUATION
- 1. FACIUTY NAME
- 2. DOCKET NUMBER
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REViSION Donald C. Cook Nuclear Plant Unit 2 05000-316 NUMBER NUMBER 4 of4 2003
-
005 00
- 17. TEXT (If more space Is required, use additionalcapies of NRC Form (366A)
Corrective Action to Prevent Recurrence:
Incorporation of the enhanced standards for disconnecting and connecting leads into Procedure 12-IHP-5021-IMP-001, "Lead Lifting/Landing And Electrical Jumper/Fuse Installation And Removal."
(Reference Corrective Action Item 3 of CR 03365009)
Incorporation of the enhanced standards for disconnecting and connecting leads into Procedure PMP-2291-PLN-001, "Work Control Activity Planning Process." (Reference Corrective Action Item 4 of CR 03365009)
Previous Similar Events
None. A review of the LERs issued by I&M from 1998 to present, for both units, identified no similar occurrences (i.e., reactor trips due to maintenance personnel errors).