05000331/LER-2007-007, Regarding Reactor Scram Due to 1A2 Non-essential Bus Lockout
| ML071580204 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 06/01/2007 |
| From: | Vanmiddlesworth G Duane Arnold |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NG-07-0468 LER 07-007-00 | |
| Download: ML071580204 (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 |
| 3312007007R00 - NRC Website | |
text
0FPL Energy.
Duane Arnold Energy Center FPL Energy Duane Arnold, LLC 3277 DAEC Road Palo, Iowa 52324 June 1,2007 NG-07-0468 10 CFR 50.73 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Licensee Event Report #2007-007-00 Please find attached the subject Licensee Event Report (LER) submitted in accordance with 10 CFR 50.73. This letter makes no new commitments or changes to any existing commitments.
Gary Van Middlesworth Site Vice President, Duane Arnold Energy Center FPL Energy Duane Arnold, LLC cc:
Administrator, Region III, USNRC Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC (See reverse for required number of may not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
- 3. PAGE Duane Arnold Energy Center 05000 331 1
OF 4
- 4. TITLE Reactor Scram Due to 1A2 Non-essential Bus Lockout
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 04 02 2007 2007 7
0 06 01 2007 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 1 0
20.2201(b)
[] 20.2203(a)(3)(i)
[
50.73(a)(2)(i)(C)
[] 50.73(a)(2)(vii)
E] 20.2201(d)
E] 20.2203(a)(3)(ii)
[
50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
[]
20.2203(a)(4)
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50.73(a)(2)(ii)(B)
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50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
E] 50.36(c)(1)(i)(A)
[
50.73(a)(2)(iii)
[j 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[
20.2203(a)(2)(ii)
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50.36(c)(1)(ii)(A)
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50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x) 9L 20.2203(a)(2)(iii)
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50.36(c)(2)
LI 50.73(a)(2)(v)(A)
E] 73.71(a)(4) 98 E] 20.2203(a)(2)(iv)
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50.46(a)(3)(ii)
E] 50.73(a)(2)(v)(B)
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73.71 (a)(5)
[] 20.2203(a)(2)(v)
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50.73(a)(2)(i)(A)
LI 50.73(a)(2)(v)(C)
LI OTHER E] 20.2203(a)(2)(vi)
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50.73(a)(2)(i)(B)
LI 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Indude Aea Code)
Bob Murrell, Engineering Analyst (319) 851-7900MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPOA
CAUSE
SYSTEM COMPONENT MANU-REPOA FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION Q] YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On April 2, 2007, while operating at 98% power, during performance of planned preventative maintenance on Non-essential Electrical 4160V AC Bus 1A2, a bus lockout occured at 11:25. The loss of the 1A2 switchgear resulted in the loss of 'B' Reactor Feed Pump and 'B' Condensate Pump. At 11:25, a manual reactor scram was inserted due to Reactor Pressure Vessel level approaching 170 inches (automatic scram level). The scram went to completion. After the scram, vessel level rose to 211 inches, causing the 'A' Reactor Feed Pump to trip.
Before the 'A' Reactor Feed Pump could be restarted and adequate feed flow established, another scram signal was received when reactor level dropped below 170 inches. Control rods were already inserted and isolations had been completed prior to this second scram signal being received. The Reactor Operator successfully restored level using the Feedwater Regulating Valve as directed by the Operations Shift Manager.
The cause of this event is the failure to recognize the risk associated with performing the maintenance on-line without putting physical and process protections in place.
There were no actual safety consequences and no effect on public health and safety as a result of this event.
NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE.EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR I SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2007 007 00 2of4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
I. Description of Event
On April 2, 2007, Work Orders (WO) 1135564, "Clean and Inspect three 1A2 Bus Auxiliary 151-201 Over Current (OC) Relays" and 1135565 "Clean and inspect 1A2 Bus 151 N-201 Ground Relay" were signed on by the Work Control Center.
One Electrician and one Electrician Apprentice began work shortly after 0800 by removing the relays from panel in 1A2. The relays were taken to the corridor outside the Essential Switchgear room and calibrated. The as-found relay calibration results were found to be within specification.
Both Electricians and a Nuclear Oversight (NOS) observer entered the 1A2 (4160 volt non-essential AC Bus) Non-essential Switchgear room and began replacing the relays, starting with the 151 N-201 Ground Relay.
The relay panel case and internal panel contacts were inspected for foreign material prior to re-installation. The Ground Relay was replaced followed by the replacement of its relay contact paddle and finally its cover. The Phase 1 over current relay (151-201) replacement was completed in an identical manner. All actions were completed as required with no abnormalities noticed.
The same actions were started on the Phase 2 over current relay; however, approximately 20-30 seconds after the paddle was inserted and prior to the apprentice electrician installing the glass cover, the 186-2 Lockout Relay tripped, causing isolation of bus 1A2. The electricians stopped work and contacted the Shift Manager.
In the Control Room, at 1125, annunciator 1 C08 B (A-9) "Bus 1A2 Lockout Trip or Loss of Voltage" alarm was activated. Loss of 1A2 switchgear resulted in the loss of 'B' Reactor Feed Pump and 'B' Condensate Pump.
At 1125, a manual reactor scram was inserted due to Reactor Pressure Vessel level approaching 170 inches (automatic scram level). The scram went to completion. After the scram, vessel level rose to 211 inches, causing the 'A' Reactor Feed Pump to trip. Before the 'A' Reactor Feed Pump could be restarted and adequate feed flow established, another scram signal was received when reactor level dropped below 170 inches. Control rods were already inserted and isolations had been completed prior to this second scram signal being received. The Reactor Operator successfully restored level using the Feedwater Regulating Valve as directed by the Operations Shift Manager. The reactor was brought to MODE 3.
Notifications were made under 10 CFR 50.72(b)(3)(iv)(A) and 50.72(b)(2)(iv)(B) on April 2, 2007 and are listed as event number 43271.
I1. Assessment of Safety Consequences:
This event was a lockout of the 1A2 bus and subsequent manual scram from 98% power (1880 MWth). The manual scram was inserted in anticipation of an automatic scram on low reactor water level. The low reactor water level was due to the trip of a feedwater pump and a condensate pump.
The feedwater and condensate pump trips were due to the lockout of the 1A2 4160V AC electrical bus.U.S. NUCLEAR REGULATORY COMMISSION (1-200 1)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
I SEQUENTIAL REVISION Duan NUMBER NUMBER 2007 007 00 3of4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
A trip of one feedwater pump is an anticipated operational transient described in Updated Final Safety Analysis Report (UFSAR) 15.1.7.4. As described in UFSAR 15.1.7.4, the expected plant response to a trip of one feedwater pump is a reactor scram and both HPCI and RCIC initiate on low-low reactor water level (119.5"). The actual plant response for this scram was much milder and clearly bounded by the trip of one feedwater pump event described in UFSAR 15.1.7.4.
Therefore, the plant shutdown did not result in any radiological or nuclear concern which would impact the health and safety of the public.
This event did not result in a Safety System Functional Failure.
I1l. Cause of Event:
An investigation into this event was completed under Root Cause Evaluation (RCE) 1065.
Overall RCE Conclusions The human performance investigation conducted did not identify any discrepancies or inappropriate actions on the part of those involved in the field work or the control room crew on the day of the event.
Although not specifically targeted, equipment failures (relays, wiring, and panel configurations) were investigated. For the logic to be met (1A2 bus lockout) one of eight relays, or the lockout relay would have had to change state. Relays were sent off to an independent lab for testing. Detailed troubleshooting activities did not identify any source or cause that was consistent with or leading to the event. Given the fact that no equipment failures were identified, a human performance event (bumping or error) would be the likely cause of the actuation.
The RCE identified the main contributor of the event to be an organizational failure. Specifically, the organization failed to recognize the risk associated with performing the maintenance on-line without putting physical and process protections in place. Contributing to this cause was the successful performance of this maintenance on several previous occasions.
Root Causes An organizational failure, which allowed work on a system with the risk potential for plant impact is considered to be the cause of this event. The failure was the lack of the organization to recognize the risk and place physical and process protection associated with performing the maintenance on-line.
Contributing to this cause was the successful performance of this maintenance on several occasions previously. A human performance event (bumping or error) would be the likely cause of the actuation.U.S. NUCLEAR REGULATORY COMMISSION (1-200 1)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR I SEQUENTIAL REVISION Duane Arnold Energy Center 5000331 NUMBER NUMBER
- 1 02007 --
007 00 4 of 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
IV. Corrective Actions
Corrective Actions to Prevent Recurrence (CATPRs)
Change work orders / model tasks for bus relays associated with 1A1, 1A2, 1A3, and 1A4 to be done in a refueling outage mode. This corrective action should only be undone if a design / process change is made to facilitate work on-line without risk to tripping the lockout relays.
V. Additional Information
Previous Similar Occurrences:
From LER review over the previous 5 years, the following four similar occurrences were identified in:
LER 2007-006 - Reactor Shutdown as a Result of a Chemistry Excursion LER 2006-005 - Reactor Scram During Main Turbine Testing LER 2003-005 - Unplanned Manual Reactor Scram due to High Reactor Coolant Conductivity LER 2003-006 - Unplanned Manual Reactor Scram due to Degrading Condenser Vacuum EIIS System and Component Codes:
EB - Medium-Voltage Power System-Class 1 E
Reporting Requirements
This report is being submitted pursuant to 1 OCFR50.73(a)(2)(iv)(A).