Event Notification Report for March 11, 2006
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/10/2006 - 03/11/2006
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 42405
Facility: LASALLE
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: TERRY NANCE
HQ OPS Officer: BILL GOTT
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: TERRY NANCE
HQ OPS Officer: BILL GOTT
Notification Date: 03/11/2006
Notification Time: 06:36 [ET]
Event Date: 03/11/2006
Event Time: 00:30 [CST]
Last Update Date: 03/15/2006
Notification Time: 06:36 [ET]
Event Date: 03/11/2006
Event Time: 00:30 [CST]
Last Update Date: 03/15/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
JAMNES CAMERON (R3)
JAMNES CAMERON (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
UNANALYZED CONDITION
"This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B), Degraded or Unanalyzed Condition. During Division 1 Core Standby Cooling System (CSCS) valve replacement work, it was discovered that temporary seismic supports required to maintain this piping within the Safe Shutdown Earthquake (SSE) design envelope were not installed. This condition was discovered during the restoration phase of the maintenance and the requirements for the temporary supports no longer exist. At the time of discovery, the required permanent seismic qualified support system had been re-established. The Division 1 CSCS System is common to both units and circulates lake water from the Ultimate Heat Sink (UHS) to the Division 1 Residual Heat Removal (RHR) heat exchangers, Division 1 Diesel Generator cooler, Division 1 Emergency Core Cooling System (ECCS) area room coolers, Division 1 RHR Pump seal coolers, and the Low Pressure Core Spray (LPCS) motor coolers. During the period with the required seismic supports not installed, both Unit 1 and Unit 2 were in an unanalyzed condition due to the inability of the Division 1 CSCS piping to meet its seismic qualification. A postulated structural failure of the Division 1 CSCS piping during an SSE could affect areas containing redundant equipment due to flooding."
The licensee is investigating the failure to install the supports. The supports were only needed during the maintenance.
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 0025 EST ON 3/15/06 FROM T. NANCE TO W. GOTT * * *
"On March 11, 2006, it was identified that temporary seismic supports required to maintain the Core Standby Cooling System piping within the Safe Shutdown Earthquake (SSE) envelope during valve replacement activities were not installed. This was reported under 10 CFR 50.72 (b)(3)(ii)(B) as a degraded or unanalyzed condition that significantly degrades plant safety.
"Subsequent evaluations have been performed to demonstrate that the piping remained within the SSE envelope with the missing supports. These evaluations have concluded that the remaining supports were sufficient to ensure that piping would remain intact during an SSE. As such, this event notification is being retracted."
The licensee notified the NRC Resident Inspector.
Notified R3DO (D. Passehl)
"This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B), Degraded or Unanalyzed Condition. During Division 1 Core Standby Cooling System (CSCS) valve replacement work, it was discovered that temporary seismic supports required to maintain this piping within the Safe Shutdown Earthquake (SSE) design envelope were not installed. This condition was discovered during the restoration phase of the maintenance and the requirements for the temporary supports no longer exist. At the time of discovery, the required permanent seismic qualified support system had been re-established. The Division 1 CSCS System is common to both units and circulates lake water from the Ultimate Heat Sink (UHS) to the Division 1 Residual Heat Removal (RHR) heat exchangers, Division 1 Diesel Generator cooler, Division 1 Emergency Core Cooling System (ECCS) area room coolers, Division 1 RHR Pump seal coolers, and the Low Pressure Core Spray (LPCS) motor coolers. During the period with the required seismic supports not installed, both Unit 1 and Unit 2 were in an unanalyzed condition due to the inability of the Division 1 CSCS piping to meet its seismic qualification. A postulated structural failure of the Division 1 CSCS piping during an SSE could affect areas containing redundant equipment due to flooding."
The licensee is investigating the failure to install the supports. The supports were only needed during the maintenance.
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 0025 EST ON 3/15/06 FROM T. NANCE TO W. GOTT * * *
"On March 11, 2006, it was identified that temporary seismic supports required to maintain the Core Standby Cooling System piping within the Safe Shutdown Earthquake (SSE) envelope during valve replacement activities were not installed. This was reported under 10 CFR 50.72 (b)(3)(ii)(B) as a degraded or unanalyzed condition that significantly degrades plant safety.
"Subsequent evaluations have been performed to demonstrate that the piping remained within the SSE envelope with the missing supports. These evaluations have concluded that the remaining supports were sufficient to ensure that piping would remain intact during an SSE. As such, this event notification is being retracted."
The licensee notified the NRC Resident Inspector.
Notified R3DO (D. Passehl)
Power Reactor
Event Number: 42406
Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: GEORGE HERNEISEY
HQ OPS Officer: STEVE SANDIN
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: GEORGE HERNEISEY
HQ OPS Officer: STEVE SANDIN
Notification Date: 03/11/2006
Notification Time: 08:35 [ET]
Event Date: 03/11/2006
Event Time: 01:00 [EST]
Last Update Date: 03/11/2006
Notification Time: 08:35 [ET]
Event Date: 03/11/2006
Event Time: 01:00 [EST]
Last Update Date: 03/11/2006
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 WHITE (R1)
JOHN WHITE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
UNIT ENTERED A T.S. REQUIRED SHUTDOWN AFTER DECLARING BOTH CONDENSER OFFGAS RADIATION MONITORS INOPERABLE
"Condenser offgas radiation monitor flow was blocked. Both Offgas Radiation Monitors were affected and out of service. We entered T.S. 3.0.1 - When a Limiting Condition for Operation is not met, within 1 hour action shall be initiated to place the unit in a condition in which the specification does not apply:
"1. Hot Standby within 6 hours.
"2. Hot Shutdown with following 6 hours.
"3. Cold Shutdown with subsequent 24 hours.
"We started to shutdown the Unit at 0100 [hours] on 11 March 2006. Removed blockage from condenser offgas radiation monitors common exhaust and evaluated operability of the condenser offgas radiation monitors. At 0315 [hours] on 11 March 2006, we returned RM-A-5 and RM-A-15 [Condenser Offgas Radiation Monitors] to service and exited T.S. 3.0.1.
"We stopped the power reduction at 0251 [hours] on 11 March 2006 while performing the evaluation on the condenser offgas radiation monitors. We stopped the power reduction at 73% reactor power. We started power escalation to 100% at 0338 [hours] on 11 March 2006."
The blockage was discovered at 2125 hours on 3/10/2006 during rounds by Secondary Plant Operator who noted that the normal 1 SCFM exhaust flow was reading zero flow. During troubleshooting to discover the cause of the blockage, a bluish foreign material was found in the common exhaust piping. An investigation is underway to identify the source of the foreign material.
The licensee informed the NRC Resident Inspector.
"Condenser offgas radiation monitor flow was blocked. Both Offgas Radiation Monitors were affected and out of service. We entered T.S. 3.0.1 - When a Limiting Condition for Operation is not met, within 1 hour action shall be initiated to place the unit in a condition in which the specification does not apply:
"1. Hot Standby within 6 hours.
"2. Hot Shutdown with following 6 hours.
"3. Cold Shutdown with subsequent 24 hours.
"We started to shutdown the Unit at 0100 [hours] on 11 March 2006. Removed blockage from condenser offgas radiation monitors common exhaust and evaluated operability of the condenser offgas radiation monitors. At 0315 [hours] on 11 March 2006, we returned RM-A-5 and RM-A-15 [Condenser Offgas Radiation Monitors] to service and exited T.S. 3.0.1.
"We stopped the power reduction at 0251 [hours] on 11 March 2006 while performing the evaluation on the condenser offgas radiation monitors. We stopped the power reduction at 73% reactor power. We started power escalation to 100% at 0338 [hours] on 11 March 2006."
The blockage was discovered at 2125 hours on 3/10/2006 during rounds by Secondary Plant Operator who noted that the normal 1 SCFM exhaust flow was reading zero flow. During troubleshooting to discover the cause of the blockage, a bluish foreign material was found in the common exhaust piping. An investigation is underway to identify the source of the foreign material.
The licensee informed the NRC Resident Inspector.
Power Reactor
Event Number: 42489
Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GRANT FERNSIER
HQ OPS Officer: BILL GOTT
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GRANT FERNSIER
HQ OPS Officer: BILL GOTT
Notification Date: 04/11/2006
Notification Time: 09:35 [ET]
Event Date: 03/11/2006
Event Time: 04:32 [EST]
Last Update Date: 04/11/2006
Notification Time: 09:35 [ET]
Event Date: 03/11/2006
Event Time: 04:32 [EST]
Last Update Date: 04/11/2006
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):
JAMES NOGGLE (R1)
JAMES NOGGLE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
INVALID ACTUATION OF CONTAINMENT ISOLATION VALVES
"On 03/11/06 at approximately 0432 hours, with Susquehanna Unit 1 in the Refueling Mode (0% power), the Unit experienced a partial isolation of Primary Containment isolation valves. The actuation occurred when a blown fuse disrupted power to a containment isolation logic relay. The inboard isolation valve to the 'B' Drywell Floor Drain Sump and the Containment Instrument Gas purge supply valve to the TIP indexer both closed. The 'A' Standby Gas Treatment system fan and the 'A' Reactor Building Recirculation fan successfully auto started during the event.
"This event constitutes an invalid system actuation and is reportable under 10CFR 50.73(a)(2)(iv)(A). The condition meets the criteria of 10CFR 50.73(a)(2)(iv)(B)(2) because a general containment isolation signal affected containment isolation valves in more than one system. This notification is being provided via a 60-day optional phone call as permitted under 10CFR 50.73(a)(1) in lieu of a written LER.
"As stated above, both valves affected by the invalid signal fulfilled their isolation function by successfully closing. At the time of the event, no other Primary Containment isolation valve was being maintained in a manner that required re-positioning. There were no challenges to the Reactor as a result of this event. The plant responded as expected.
"Following replacement of the subject fuse, the isolation logic was successfully reset and affected equipment was restored to the desired status."
The licensee notified the NRC Resident Inspector.
"On 03/11/06 at approximately 0432 hours, with Susquehanna Unit 1 in the Refueling Mode (0% power), the Unit experienced a partial isolation of Primary Containment isolation valves. The actuation occurred when a blown fuse disrupted power to a containment isolation logic relay. The inboard isolation valve to the 'B' Drywell Floor Drain Sump and the Containment Instrument Gas purge supply valve to the TIP indexer both closed. The 'A' Standby Gas Treatment system fan and the 'A' Reactor Building Recirculation fan successfully auto started during the event.
"This event constitutes an invalid system actuation and is reportable under 10CFR 50.73(a)(2)(iv)(A). The condition meets the criteria of 10CFR 50.73(a)(2)(iv)(B)(2) because a general containment isolation signal affected containment isolation valves in more than one system. This notification is being provided via a 60-day optional phone call as permitted under 10CFR 50.73(a)(1) in lieu of a written LER.
"As stated above, both valves affected by the invalid signal fulfilled their isolation function by successfully closing. At the time of the event, no other Primary Containment isolation valve was being maintained in a manner that required re-positioning. There were no challenges to the Reactor as a result of this event. The plant responded as expected.
"Following replacement of the subject fuse, the isolation logic was successfully reset and affected equipment was restored to the desired status."
The licensee notified the NRC Resident Inspector.