Event Notification Report for April 15, 2005
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
04/14/2005 - 04/15/2005
EVENT NUMBERS
416044160541606416014160241770
Power Reactor
Event Number: 41604
Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: DAN KELLEY
HQ OPS Officer: MIKE RIPLEY
Region: 3 State: IA
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: DAN KELLEY
HQ OPS Officer: MIKE RIPLEY
Notification Date: 04/15/2005
Notification Time: 14:09 [ET]
Event Date: 04/15/2005
Event Time: 08:30 [CST]
Last Update Date: 04/15/2005
Notification Time: 14:09 [ET]
Event Date: 04/15/2005
Event Time: 08:30 [CST]
Last Update Date: 04/15/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
10 CFR Section:
26.73 - FITNESS FOR DUTY
Person (Organization):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
FITNESS FOR DUTY
A contract supervisor tested positive for alcohol during a for-cause fitness for duty test. The individual's access to the plant was terminated. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.
A contract supervisor tested positive for alcohol during a for-cause fitness for duty test. The individual's access to the plant was terminated. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.
Power Reactor
Event Number: 41605
Facility: KEWAUNEE
Region: 3 State: WI
Unit: [1] [] []
RX Type: [1] W-2-LP
NRC Notified By: DAVID KARST
HQ OPS Officer: MIKE RIPLEY
Region: 3 State: WI
Unit: [1] [] []
RX Type: [1] W-2-LP
NRC Notified By: DAVID KARST
HQ OPS Officer: MIKE RIPLEY
Notification Date: 04/15/2005
Notification Time: 16:32 [ET]
Event Date: 04/15/2005
Event Time: 07:35 [CST]
Last Update Date: 04/15/2005
Notification Time: 16:32 [ET]
Event Date: 04/15/2005
Event Time: 07:35 [CST]
Last Update Date: 04/15/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
10 CFR Section:
26.73 - FITNESS FOR DUTY
Person (Organization):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling Shutdown | 0 | Refueling Shutdown |
FITNESS FOR DUTY
A non-licensed employee supervisor tested positive for alcohol during a for-cause fitness for duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.
A non-licensed employee supervisor tested positive for alcohol during a for-cause fitness for duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.
Power Reactor
Event Number: 41606
Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFF HUMAN
HQ OPS Officer: MIKE RIPLEY
Region: 3 State: MN
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFF HUMAN
HQ OPS Officer: MIKE RIPLEY
Notification Date: 04/15/2005
Notification Time: 22:33 [ET]
Event Date: 04/15/2005
Event Time: 21:30 [CDT]
Last Update Date: 04/15/2005
Notification Time: 22:33 [ET]
Event Date: 04/15/2005
Event Time: 21:30 [CDT]
Last Update Date: 04/15/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 96 | Power Operation |
PLANT SHUTDOWN REQUIRED BY TECHNICAL SPECIFICATIONS - INOPERABLE EMERGENCY DIESEL GENERATOR
"Unit 2 Train A emergency diesel generator D5 was removed from service at 0838 CDT on 4/11/05 for surveillance testing and Technical Specification (TS) 3.8.1, 'AC Source - Operating,' Condition B, 'One DG inoperable,' was entered. TS Required Action 3.8.1.B.4 requires D5 be restored to operable status with a Completion Time of 7 days. At 1026 CDT, the test was halted due to high-indicated crankcase pressure on Engine 2 (D5 is a tandem engine generator). The test procedure specifies shutting down the DG if crankcase pressure exceeds 30 mm for more than a few minutes (the setpoint for the crankcase pressure trip is 52 mm).
"Investigation of the cause of the high-indicated crankcase pressure on Engine 2 (and whether Engine 1 was effected) started immediately. Unit 2 Train B emergency diesel generator was demonstrated operable by completing a surveillance test at 0423 CDT on 4/12/05. Evaluation of the scope of work to return D5 to operable status and the schedule for completing the work indicated that repairs could not be completed within the 7 days allowed outage time. Based on this assessment an orderly shutdown of Unit 2 is being performed.
"Shutdown of Unit 2 commenced at 21:30 CDT on 4/15/05. Unit 2 shutdown will continue until D5 is restored to operable status."
The licensee intends to issue a press release. The licensee notified the NRC Resident Inspector.
"Unit 2 Train A emergency diesel generator D5 was removed from service at 0838 CDT on 4/11/05 for surveillance testing and Technical Specification (TS) 3.8.1, 'AC Source - Operating,' Condition B, 'One DG inoperable,' was entered. TS Required Action 3.8.1.B.4 requires D5 be restored to operable status with a Completion Time of 7 days. At 1026 CDT, the test was halted due to high-indicated crankcase pressure on Engine 2 (D5 is a tandem engine generator). The test procedure specifies shutting down the DG if crankcase pressure exceeds 30 mm for more than a few minutes (the setpoint for the crankcase pressure trip is 52 mm).
"Investigation of the cause of the high-indicated crankcase pressure on Engine 2 (and whether Engine 1 was effected) started immediately. Unit 2 Train B emergency diesel generator was demonstrated operable by completing a surveillance test at 0423 CDT on 4/12/05. Evaluation of the scope of work to return D5 to operable status and the schedule for completing the work indicated that repairs could not be completed within the 7 days allowed outage time. Based on this assessment an orderly shutdown of Unit 2 is being performed.
"Shutdown of Unit 2 commenced at 21:30 CDT on 4/15/05. Unit 2 shutdown will continue until D5 is restored to operable status."
The licensee intends to issue a press release. The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 41601
Facility: COOPER
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVEN JOBE
HQ OPS Officer: PETE SNYDER
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVEN JOBE
HQ OPS Officer: PETE SNYDER
Notification Date: 04/15/2005
Notification Time: 09:26 [ET]
Event Date: 04/15/2005
Event Time: 04:36 [CDT]
Last Update Date: 04/15/2005
Notification Time: 09:26 [ET]
Event Date: 04/15/2005
Event Time: 04:36 [CDT]
Last Update Date: 04/15/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
REBECCA NEASE (R4)
REBECCA NEASE (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown |
REACTOR SCRAM AND SPECIFIED SYSTEM ACTUATIONS
"This is both a four hour and eight hour report for the following event:
"4 hour Non-Emergency for ECCS Discharge to the RCS and RPS Actuation;
"8 hour Non-Emergency for Groups 2 & 6. Also, HPCI [High Pressure Coolant Injection] was placed in pull-to-lock, thereby making HPCI unavailable.
"At 0436 [hrs.] CDT 04/15/2005, Cooper Nuclear Station experienced a reactor feed water level transient followed by reactor vessel water level low level alarm and an automatic scram. HPCI and RCIC [Reactor Core Isolation Cooling] automatically initiated and injected into the reactor vessel. Groups 2, 3, and 6 actuated on low reactor water level. Both reactor recirculation pumps tripped and RWCU [Reactor Water Clean-Up system] isolated. Reactor vessel water level reached approximately [minus] 70 inches (instrument zero, approximately 90 inches of water remained above the fuel) and recovered after HPCI and RCIC had begun to inject. All post-scram automatic actions were as expected. HPCI was placed in pull-to-lock to prevent over feeding. The plant is being cooled down to establish natural circulation flow and permit recovery of recirculation pumps.
"Investigation into the cause of the reactor feed water level transient is ongoing."
All rods inserted during the scram. No safety relief valves lifted during the transient. All group isolation signals have been reset. The electrical grid is stable. Decay heat is being removed via steam dumps to the main condenser.
The licensee has notified the NRC Resident Inspector.
"This is both a four hour and eight hour report for the following event:
"4 hour Non-Emergency for ECCS Discharge to the RCS and RPS Actuation;
"8 hour Non-Emergency for Groups 2 & 6. Also, HPCI [High Pressure Coolant Injection] was placed in pull-to-lock, thereby making HPCI unavailable.
"At 0436 [hrs.] CDT 04/15/2005, Cooper Nuclear Station experienced a reactor feed water level transient followed by reactor vessel water level low level alarm and an automatic scram. HPCI and RCIC [Reactor Core Isolation Cooling] automatically initiated and injected into the reactor vessel. Groups 2, 3, and 6 actuated on low reactor water level. Both reactor recirculation pumps tripped and RWCU [Reactor Water Clean-Up system] isolated. Reactor vessel water level reached approximately [minus] 70 inches (instrument zero, approximately 90 inches of water remained above the fuel) and recovered after HPCI and RCIC had begun to inject. All post-scram automatic actions were as expected. HPCI was placed in pull-to-lock to prevent over feeding. The plant is being cooled down to establish natural circulation flow and permit recovery of recirculation pumps.
"Investigation into the cause of the reactor feed water level transient is ongoing."
All rods inserted during the scram. No safety relief valves lifted during the transient. All group isolation signals have been reset. The electrical grid is stable. Decay heat is being removed via steam dumps to the main condenser.
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 41602
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: KEN PETERSEN
HQ OPS Officer: PETE SNYDER
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: KEN PETERSEN
HQ OPS Officer: PETE SNYDER
Notification Date: 04/15/2005
Notification Time: 10:02 [ET]
Event Date: 04/15/2005
Event Time: 05:17 [CDT]
Last Update Date: 04/20/2005
Notification Time: 10:02 [ET]
Event Date: 04/15/2005
Event Time: 05:17 [CDT]
Last Update Date: 04/20/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
REBECCA NEASE (R4)
REBECCA NEASE (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
STEAM GENERATOR BOWL DRAIN LINE WELD LEAK DISCOVERED DURING VISUAL EXAM
"During the performance of Alloy 600 bare metal visual examinations, the presence of boric acid deposits indicated a leak in a weld in the steam generator D (EBB01D) bowl drain line.
"This condition is being reported as a degraded condition on the primary coolant system under 10 CFR 50.72 (B)(3)(ii)(A).
"During investigation the defect was characterized as a 1/16" rounded indication in the drain line to boss weld.
"This condition cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws' or ASME Section XI, Table IWB-3410-1, 'Acceptable Standards.'
"The [NRC] Resident Inspector has been notified."
* * * UPDATE FROM S. GIFFORD TO J. KNOKE AT 15:10 EDT ON 04/20/05 * *
"On 04/20/05 at 12:10 CDT, during the performance of Alloy 600 Bare Metal Visual Examinations, the presence of boric acid deposits indicated a leak in a weld in the Steam Generator C (EBB01C) bowl drain line."
The inspection of Steam Generators A & B showed no presence of boric acid deposits.
The licensee notified NRC Resident Inspector. Notified the R4DO (Shannon).
"During the performance of Alloy 600 bare metal visual examinations, the presence of boric acid deposits indicated a leak in a weld in the steam generator D (EBB01D) bowl drain line.
"This condition is being reported as a degraded condition on the primary coolant system under 10 CFR 50.72 (B)(3)(ii)(A).
"During investigation the defect was characterized as a 1/16" rounded indication in the drain line to boss weld.
"This condition cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws' or ASME Section XI, Table IWB-3410-1, 'Acceptable Standards.'
"The [NRC] Resident Inspector has been notified."
* * * UPDATE FROM S. GIFFORD TO J. KNOKE AT 15:10 EDT ON 04/20/05 * *
"On 04/20/05 at 12:10 CDT, during the performance of Alloy 600 Bare Metal Visual Examinations, the presence of boric acid deposits indicated a leak in a weld in the Steam Generator C (EBB01C) bowl drain line."
The inspection of Steam Generators A & B showed no presence of boric acid deposits.
The licensee notified NRC Resident Inspector. Notified the R4DO (Shannon).
Power Reactor
Event Number: 41770
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: ERIC MARKS
HQ OPS Officer: CHAUNCEY GOULD
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: ERIC MARKS
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 06/14/2005
Notification Time: 13:45 [ET]
Event Date: 04/15/2005
Event Time: 15:15 [CDT]
Last Update Date: 06/14/2005
Notification Time: 13:45 [ET]
Event Date: 04/15/2005
Event Time: 15:15 [CDT]
Last Update Date: 06/14/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
PATRICK LOUDEN (R3)
PATRICK LOUDEN (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
INVALID ACTUATION DURING EXCESS FLOW CHECK VALVE TESTING
"At 1515 hours on April 15, 2005, while conducting excess flow check valve testing in conjunction with a reactor vessel pressure test during a scheduled refueling outage, Unit 1 received an invalid Anticipated Transient Without Scram (ATWS) Channel A trip signal and invalid reactor Lo-Lo water level trips on the A and C channels for ECCS. The operating Reactor Recirculation pump (1 A) tripped and both Emergency Diesel Generators (Unit 1 and Unit 0) auto-started and ran unloaded. RCIC received an initiation signal, which resulted in proper valve repositioning for the system but no injection due to the steam supply being isolated to support the reactor vessel pressure test. In addition, all appropriate running equipment tripped due to load shedding. Reactor pressure was 787 psig at the time of the event and no pressure transient or vessel injection occurred. All equipment responded to the actuation as expected for plant conditions. The excess flow check valve testing was stopped and the reactor vessel pressure test was terminated as a result of this invalid actuation.
"It was subsequently determined that a leaking drain valve at the instrument rack where the excess flow check valve testing was being performed caused a slow depressurization of the high-pressure side sensing line for the level transmitters associated with ATWS Channel A and the A and C channels for ECCS, resulting in an erroneous level signal and the invalid actuation. The instrument rack drain valve is scheduled for replacement."
The NRC Resident Inspector was notified
"At 1515 hours on April 15, 2005, while conducting excess flow check valve testing in conjunction with a reactor vessel pressure test during a scheduled refueling outage, Unit 1 received an invalid Anticipated Transient Without Scram (ATWS) Channel A trip signal and invalid reactor Lo-Lo water level trips on the A and C channels for ECCS. The operating Reactor Recirculation pump (1 A) tripped and both Emergency Diesel Generators (Unit 1 and Unit 0) auto-started and ran unloaded. RCIC received an initiation signal, which resulted in proper valve repositioning for the system but no injection due to the steam supply being isolated to support the reactor vessel pressure test. In addition, all appropriate running equipment tripped due to load shedding. Reactor pressure was 787 psig at the time of the event and no pressure transient or vessel injection occurred. All equipment responded to the actuation as expected for plant conditions. The excess flow check valve testing was stopped and the reactor vessel pressure test was terminated as a result of this invalid actuation.
"It was subsequently determined that a leaking drain valve at the instrument rack where the excess flow check valve testing was being performed caused a slow depressurization of the high-pressure side sensing line for the level transmitters associated with ATWS Channel A and the A and C channels for ECCS, resulting in an erroneous level signal and the invalid actuation. The instrument rack drain valve is scheduled for replacement."
The NRC Resident Inspector was notified