Event Notification Report for November 04, 2003
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
11/03/2003 - 11/04/2003
EVENT NUMBERS
4029540296402974029840324
Power Reactor
Event Number: 40295
Facility: WATERFORD
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: RON WILLIAMS
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: RON WILLIAMS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 11/04/2003
Notification Time: 12:08 [ET]
Event Date: 11/04/2003
Event Time: 00:00 [CST]
Last Update Date: 11/04/2003
Notification Time: 12:08 [ET]
Event Date: 11/04/2003
Event Time: 00:00 [CST]
Last Update Date: 11/04/2003
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):
MARK SHAFFER (R4)
MARK SHAFFER (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | N | 0 | Refueling | 0 | Refueling |
NUMBER OF STEAM GENERATOR TUBES REQUIRING REPAIR EXCEEDS 1 PERCENT
"At 0400 CDT on November 4, 2003 with Waterford 3 refueled, steam generator inspection was being performed in accordance with Technical Specification (TS) 4.4.4.2, Steam Generator Tube Sample Selection and Inspection. The results of the on-going inspection performed on Steam Generator 31 indicated that greater than 1% of the inspected tubes require repair. The number of inspected tubes for SG 31 is 8906 and to date 98 tubes require repair. Based on TS 4.4.4.2.C, this determination results in the Category C-3. The current inspection results do not meet the criteria specified for steam generator tube degradation in NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73. However, for Category C-3, TS Table 4.4-2, Steam Generator Tube Inspection, requires that the results of the inspection be reported under 10 CFR 50.72(b)(2), which was superceded by an 8 hr notification in accordance with 50.72(b)(3)(ii).
"Per TS 4.4.4.5, the final results of the SG tube inspection, which fall into Category C-3, shall be reported in a special report within 30 days and prior to resumption of plant operation."
The licensee has notified the NRC Resident Inspector.
"At 0400 CDT on November 4, 2003 with Waterford 3 refueled, steam generator inspection was being performed in accordance with Technical Specification (TS) 4.4.4.2, Steam Generator Tube Sample Selection and Inspection. The results of the on-going inspection performed on Steam Generator 31 indicated that greater than 1% of the inspected tubes require repair. The number of inspected tubes for SG 31 is 8906 and to date 98 tubes require repair. Based on TS 4.4.4.2.C, this determination results in the Category C-3. The current inspection results do not meet the criteria specified for steam generator tube degradation in NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73. However, for Category C-3, TS Table 4.4-2, Steam Generator Tube Inspection, requires that the results of the inspection be reported under 10 CFR 50.72(b)(2), which was superceded by an 8 hr notification in accordance with 50.72(b)(3)(ii).
"Per TS 4.4.4.5, the final results of the SG tube inspection, which fall into Category C-3, shall be reported in a special report within 30 days and prior to resumption of plant operation."
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 40296
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: ADAM MILLER
HQ OPS Officer: GERRY WAIG
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: ADAM MILLER
HQ OPS Officer: GERRY WAIG
Notification Date: 11/04/2003
Notification Time: 15:50 [ET]
Event Date: 11/04/2003
Event Time: 14:30 [EST]
Last Update Date: 11/24/2003
Notification Time: 15:50 [ET]
Event Date: 11/04/2003
Event Time: 14:30 [EST]
Last Update Date: 11/24/2003
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):
RICHARD BARKLEY (R1)
RICHARD BARKLEY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
PRESSURIZER HEATER BUNDLE DIAPHRAGM PLATE DEGRADED CONDITION
"On November 4, 2003, during the TMI Unit 1 15th (T1R15) refueling outage, an inspection of the Pressurizer Heater Bundle (PHB) Diaphragm Plate was completed. This Inspection identified a leak path emanating from the lower Pressurizer Heater Bundle.
"The initial indication of a potential Reactor Coolant System (RCS) leak was boric acid residue located between the PHB Diaphragm Plate and the PHB Cover Plate. Initially the leak was believed to be from a seal weld, which is considered comparable to a gasket leak. Following disassembly of the PHB Cover Plate and performance of NDE, it was determined that the pathway was through the edge of the PHB Diaphragm Plate. This degraded condition of the PHB Diaphragm Plate is indicative of a RCS pressure boundary leak. This notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A)."
The licensee provided the following pre-refuel outage information:
RCS Leak Rate = less than 0.1 gallons per minute
Activity = 0.45 microcuries/milliliter
TS [Technical Specifications] Limits: No leakage (3.1.6.4)
Secondary System Activity = less than 1E-10 microcuries/milliliter
The licensee has notified the NRC Resident Inspector.
* * * UPDATE ON 11/24/03 AT 0114 EST FROM JOHN SCHORK TO HOWIE CROUCH * * *
"Subsequent to the initial report made on 11/4/03, the Pressurizer Heater Bundle (PHB) Diaphragm Plate was repaired. On November 23, 2003 during performance of post-maintenance testing inspections with the plant in Hot Shutdown, steam was observed emanating from either the seal weld or the PHB diaphragm plate.
"This update is being made to EN 40296 because the observed leak is being conservatively classified as a leak from the Primary System Pressure Boundary and is being addressed in a manner consistent with TMI Technical Specifications 3.1.6.4 and 3.1.6.6. An evaluation of the safety implications of the leak has been initiated. A condition report has been generated to capture all of the actions that have been and will be taken in response to the leak.
"The plant is being taken to cold shutdown in order to perform an inspection and repair of the leak. The plant continues to be subcritical with all control rods fully inserted and the Reactor Coolant System boron concentration is at the refueling boron concentration. The plant continues to remain in the T1R15 refueling outage.
"The NRC TMI-1 Sr. Resident Inspector has been notified of the leak. No other notifications were made to the State, Local or other governmental agencies. No press release has been issued regarding the event. The cause and corrective action to repair the leak will be addressed in the licensee event report being submitted in response to EN 40296.
"The location of the leak observed on November 23, 2003 is in the immediate vicinity of the lower Pressurizer Heater Bundle. There has been no determination of the volumetric leak rate. The leak consists [of] steam wisping from the location. The start date of the leak is November 23, 2003 and the leak was initially observed during hot shutdown checks at 1930 hours on November 23, 2003. There has been no radiological release to the environment as a result of this leak."
The licensee has notified the NRC Senior Resident Inspector.
Notified R1DO (John Rogge) and NRR (William Ruland).
"On November 4, 2003, during the TMI Unit 1 15th (T1R15) refueling outage, an inspection of the Pressurizer Heater Bundle (PHB) Diaphragm Plate was completed. This Inspection identified a leak path emanating from the lower Pressurizer Heater Bundle.
"The initial indication of a potential Reactor Coolant System (RCS) leak was boric acid residue located between the PHB Diaphragm Plate and the PHB Cover Plate. Initially the leak was believed to be from a seal weld, which is considered comparable to a gasket leak. Following disassembly of the PHB Cover Plate and performance of NDE, it was determined that the pathway was through the edge of the PHB Diaphragm Plate. This degraded condition of the PHB Diaphragm Plate is indicative of a RCS pressure boundary leak. This notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A)."
The licensee provided the following pre-refuel outage information:
RCS Leak Rate = less than 0.1 gallons per minute
Activity = 0.45 microcuries/milliliter
TS [Technical Specifications] Limits: No leakage (3.1.6.4)
Secondary System Activity = less than 1E-10 microcuries/milliliter
The licensee has notified the NRC Resident Inspector.
* * * UPDATE ON 11/24/03 AT 0114 EST FROM JOHN SCHORK TO HOWIE CROUCH * * *
"Subsequent to the initial report made on 11/4/03, the Pressurizer Heater Bundle (PHB) Diaphragm Plate was repaired. On November 23, 2003 during performance of post-maintenance testing inspections with the plant in Hot Shutdown, steam was observed emanating from either the seal weld or the PHB diaphragm plate.
"This update is being made to EN 40296 because the observed leak is being conservatively classified as a leak from the Primary System Pressure Boundary and is being addressed in a manner consistent with TMI Technical Specifications 3.1.6.4 and 3.1.6.6. An evaluation of the safety implications of the leak has been initiated. A condition report has been generated to capture all of the actions that have been and will be taken in response to the leak.
"The plant is being taken to cold shutdown in order to perform an inspection and repair of the leak. The plant continues to be subcritical with all control rods fully inserted and the Reactor Coolant System boron concentration is at the refueling boron concentration. The plant continues to remain in the T1R15 refueling outage.
"The NRC TMI-1 Sr. Resident Inspector has been notified of the leak. No other notifications were made to the State, Local or other governmental agencies. No press release has been issued regarding the event. The cause and corrective action to repair the leak will be addressed in the licensee event report being submitted in response to EN 40296.
"The location of the leak observed on November 23, 2003 is in the immediate vicinity of the lower Pressurizer Heater Bundle. There has been no determination of the volumetric leak rate. The leak consists [of] steam wisping from the location. The start date of the leak is November 23, 2003 and the leak was initially observed during hot shutdown checks at 1930 hours on November 23, 2003. There has been no radiological release to the environment as a result of this leak."
The licensee has notified the NRC Senior Resident Inspector.
Notified R1DO (John Rogge) and NRR (William Ruland).
Power Reactor
Event Number: 40297
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: DANIEL HARDIN
HQ OPS Officer: GERRY WAIG
Region: 2 State: NC
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: DANIEL HARDIN
HQ OPS Officer: GERRY WAIG
Notification Date: 11/04/2003
Notification Time: 21:20 [ET]
Event Date: 11/04/2003
Event Time: 17:32 [EST]
Last Update Date: 11/04/2003
Notification Time: 21:20 [ET]
Event Date: 11/04/2003
Event Time: 17:32 [EST]
Last Update Date: 11/04/2003
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
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
Person (Organization):
STEPHEN CAHILL (R2)
STEPHEN CAHILL (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | N | 0 | 0 |
BRUNSWICK UNIT 2 SCRAM DUE TO GENERATOR/TURBINE TRIP
"On November 4, 2003, at approximately 1732 hours, Unit 2 received a generator/turbine trip due to loss of excitation, which resulted in a Reactor Protection System (RPS) trip. Plant response to the reactor shutdown resulted in High Pressure Coolant Injection and Reactor Core Isolation Cooling [RCIC] system actuations on low reactor coolant level. Additionally, primary containment isolation system actuation signals for valve groups 1, 2, 3 , 6, 8, and 10 were received and the valves closed as required. All four emergency diesel generators automatically started, but did not load because power was never lost to the emergency buses. The loss of power from the generator trip resulted in reactor building ventilation isolation and automatic start of both trains of the Standby Gas Treatment (SBGT) system. SBGT Train A immediately tripped, but was successfully placed in service. All control rods fully inserted into the core. At approximately 1857 hours, another RPS trip was received due to low reactor coolant level while cycling Safety Relief Valves; however, all control rods were already inserted.
"The safety significance of this event is considered to be minimal. The plant responded as designed to the transient with the exception of the SBGT Train A initial starting issue.
"An event investigation team has been assembled to determine the cause of the event. Plant response to the event
is being evaluated and identified issues will be addressed prior to plant restart."
The licensee reported that the station electrical grid is normal and the emergency diesel generators have been shutdown and returned to standby status. The current plant conditions are 550 psi, 496 degrees F with RCIC operating to maintain reactor water level. The main condenser is available and being used to dump steam from the reactor. Two safety relief valves opened during the transient and reclosed as expected. The reactor water level decreased to minimum of approximately 90 inches during the transient. The cause of main steam isolation valve (MSIV) closure is under investigation.
The licensee also reported that this event caused a Group 6 Isolation, reactor building ventilation isolation, and standby gas treatment actuation signal on Brunswick Unit 1. These Unit 1 systems have been returned to a normal configuration.
The licensee has notified the NRC Resident Inspector.
"On November 4, 2003, at approximately 1732 hours, Unit 2 received a generator/turbine trip due to loss of excitation, which resulted in a Reactor Protection System (RPS) trip. Plant response to the reactor shutdown resulted in High Pressure Coolant Injection and Reactor Core Isolation Cooling [RCIC] system actuations on low reactor coolant level. Additionally, primary containment isolation system actuation signals for valve groups 1, 2, 3 , 6, 8, and 10 were received and the valves closed as required. All four emergency diesel generators automatically started, but did not load because power was never lost to the emergency buses. The loss of power from the generator trip resulted in reactor building ventilation isolation and automatic start of both trains of the Standby Gas Treatment (SBGT) system. SBGT Train A immediately tripped, but was successfully placed in service. All control rods fully inserted into the core. At approximately 1857 hours, another RPS trip was received due to low reactor coolant level while cycling Safety Relief Valves; however, all control rods were already inserted.
"The safety significance of this event is considered to be minimal. The plant responded as designed to the transient with the exception of the SBGT Train A initial starting issue.
"An event investigation team has been assembled to determine the cause of the event. Plant response to the event
is being evaluated and identified issues will be addressed prior to plant restart."
The licensee reported that the station electrical grid is normal and the emergency diesel generators have been shutdown and returned to standby status. The current plant conditions are 550 psi, 496 degrees F with RCIC operating to maintain reactor water level. The main condenser is available and being used to dump steam from the reactor. Two safety relief valves opened during the transient and reclosed as expected. The reactor water level decreased to minimum of approximately 90 inches during the transient. The cause of main steam isolation valve (MSIV) closure is under investigation.
The licensee also reported that this event caused a Group 6 Isolation, reactor building ventilation isolation, and standby gas treatment actuation signal on Brunswick Unit 1. These Unit 1 systems have been returned to a normal configuration.
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 40298
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES SMIT
HQ OPS Officer: RICH LAURA
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES SMIT
HQ OPS Officer: RICH LAURA
Notification Date: 11/05/2003
Notification Time: 03:45 [ET]
Event Date: 11/04/2003
Event Time: 21:32 [CST]
Last Update Date: 11/05/2003
Notification Time: 03:45 [ET]
Event Date: 11/04/2003
Event Time: 21:32 [CST]
Last Update Date: 11/05/2003
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
DAVID HILLS (R3)
DAVID HILLS (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
BRAIDWOOD 2 AFW SUPPORT SYSTEM ACTUATION DURING OUTAGE
"Unit 2 auxiliary feedwater support systems actuated during scheduled ATWS testing when an unrelated clearance order placement de-energized two 6.9 kv busses (256 and 258). The 2 of 4 6.9 kv bus undervoltage coincidence initiated a valid auto-start signal causing lube oil pumps 2AF01PA-A, 2AF01PB-A and 2AF01PB-C to start. Auxiliary Feedwater Pump AOV discharge valves 2AF004A and 2AF004B auto opened. 2AF01PA 4kv breaker, which was in the equipment test position. Neither auxiliary feedwater pump started and no water transferred to the steam generators."
The licensee notified the NRC resident inspector.
"Unit 2 auxiliary feedwater support systems actuated during scheduled ATWS testing when an unrelated clearance order placement de-energized two 6.9 kv busses (256 and 258). The 2 of 4 6.9 kv bus undervoltage coincidence initiated a valid auto-start signal causing lube oil pumps 2AF01PA-A, 2AF01PB-A and 2AF01PB-C to start. Auxiliary Feedwater Pump AOV discharge valves 2AF004A and 2AF004B auto opened. 2AF01PA 4kv breaker, which was in the equipment test position. Neither auxiliary feedwater pump started and no water transferred to the steam generators."
The licensee notified the NRC resident inspector.
Hospital
Event Number: 40324
Rep Org: QUEENS MEDICAL CENTER
Licensee: QUEENS MEDICAL CENTER
Region: 4
City: HONOLULU State: HI
County:
License #: 531653302
Agreement: N
Docket:
NRC Notified By: SCOTT DUPE
HQ OPS Officer: CHAUNCEY GOULD
Licensee: QUEENS MEDICAL CENTER
Region: 4
City: HONOLULU State: HI
County:
License #: 531653302
Agreement: N
Docket:
NRC Notified By: SCOTT DUPE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 11/14/2003
Notification Time: 14:41 [ET]
Event Date: 11/04/2003
Event Time: 00:00 [HST]
Last Update Date: 11/14/2003
Notification Time: 14:41 [ET]
Event Date: 11/04/2003
Event Time: 00:00 [HST]
Last Update Date: 11/14/2003
Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(ii) - LOST/STOLEN LNM>10X
10 CFR Section:
20.2201(a)(1)(ii) - LOST/STOLEN LNM>10X
Person (Organization):
WILLIAM JONES (R4)
ROBERTO TORRES (NMSS)
WILLIAM JONES (R4)
ROBERTO TORRES (NMSS)
REPORTED LOSS OF NINE IMPLANTED I-125 SEEDS
The licensee reported that 9 out of 156 I-125 seeds are unaccounted for following an autopsy performed on the recipient of an implant. The implant recipient was a dolphin which received the seeds on 08/16/03 and it died 11/02/03 and an autopsy was performed 11/04/03. The seeds probably came out of the dolphin where it was kept in a tank and washed into the harbor.
The licensee reported that 9 out of 156 I-125 seeds are unaccounted for following an autopsy performed on the recipient of an implant. The implant recipient was a dolphin which received the seeds on 08/16/03 and it died 11/02/03 and an autopsy was performed 11/04/03. The seeds probably came out of the dolphin where it was kept in a tank and washed into the harbor.