Event Notification Report for May 19, 2002
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/18/2002 - 05/19/2002
EVENT NUMBERS
38927389283892939034
Power Reactor
Event Number: 38927
Facility: LIMERICK
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PETE GARDNER
HQ OPS Officer: LEIGH TROCINE
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PETE GARDNER
HQ OPS Officer: LEIGH TROCINE
Notification Date: 05/19/2002
Notification Time: 03:33 [ET]
Event Date: 05/19/2002
Event Time: 00:54 [EDT]
Last Update Date: 05/19/2002
Notification Time: 03:33 [ET]
Event Date: 05/19/2002
Event Time: 00:54 [EDT]
Last Update Date: 05/19/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
DANIEL HOLODY (R1)
DANIEL HOLODY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 94 | Power Operation | 0 | Hot Shutdown |
AUTOMATIC REACTOR SCRAM DUE TO A TURBINE TRIP FOR UNKNOWN REASONS
The following text is a portion of a facsimile received from the licensee:
"[A] Unit-1 reactor scram occurred at 0054 hours on May 19, 2002. The reactor scram occurred as a result of a main turbine trip. The cause of the turbine trip is under investigation. All systems actuated as designed. The unit is being maintained in Hot Shutdown at normal [reactor pressure vessel] level. This report is being made pursuant to [10 CFR] 50.72(b)(2)(iv)(B) due to actuation of the reactor protection system (RPS). No [safety relief valves] lifted, and currently [one] reactor recirculation pump has been restarted[. The] scram has been reset."
The licensee stated that reactor vessel level control is currently being maintained via normal condensate and feedwater. The main steam isolation valves are open, the condenser is available as a heat sink, and reactor vessel pressure control is being maintained via normal electrohydraulic control pressure control. Containment parameters are normal. The licensee also stated that the all control rods fully inserted, that there were no engineered safety feature actuations or emergency core cooling system injections, and that none were required. There was nothing unusual or not understood with the exception of the cause of the turbine trip.
Prior to this event, the unit had been operating at reduced power to facilitate the performance of turbine stop valve testing. The licensee stated that a combined intermediate valve (CIV) was being stroked at the time of the turbine trip/reactor scram but that stroking of the CIV is currently believed to be unrelated to the turbine trip. At this time, the licensee believes that the cause of the turbine trip may have been related to the thrust bearing wear detector. However, the final root cause of the turbine trip remains under investigation.
The licensee notified the NRC resident inspector, applicable State and local agencies (including Montgomery, Chester, and Berks Counties), and the Pennsylvania Emergency Management Agency.
The following text is a portion of a facsimile received from the licensee:
"[A] Unit-1 reactor scram occurred at 0054 hours on May 19, 2002. The reactor scram occurred as a result of a main turbine trip. The cause of the turbine trip is under investigation. All systems actuated as designed. The unit is being maintained in Hot Shutdown at normal [reactor pressure vessel] level. This report is being made pursuant to [10 CFR] 50.72(b)(2)(iv)(B) due to actuation of the reactor protection system (RPS). No [safety relief valves] lifted, and currently [one] reactor recirculation pump has been restarted[. The] scram has been reset."
The licensee stated that reactor vessel level control is currently being maintained via normal condensate and feedwater. The main steam isolation valves are open, the condenser is available as a heat sink, and reactor vessel pressure control is being maintained via normal electrohydraulic control pressure control. Containment parameters are normal. The licensee also stated that the all control rods fully inserted, that there were no engineered safety feature actuations or emergency core cooling system injections, and that none were required. There was nothing unusual or not understood with the exception of the cause of the turbine trip.
Prior to this event, the unit had been operating at reduced power to facilitate the performance of turbine stop valve testing. The licensee stated that a combined intermediate valve (CIV) was being stroked at the time of the turbine trip/reactor scram but that stroking of the CIV is currently believed to be unrelated to the turbine trip. At this time, the licensee believes that the cause of the turbine trip may have been related to the thrust bearing wear detector. However, the final root cause of the turbine trip remains under investigation.
The licensee notified the NRC resident inspector, applicable State and local agencies (including Montgomery, Chester, and Berks Counties), and the Pennsylvania Emergency Management Agency.
Power Reactor
Event Number: 38928
Facility: SEQUOYAH
Region: 2 State: TN
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MICHAEL HOWARD
HQ OPS Officer: LEIGH TROCINE
Region: 2 State: TN
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MICHAEL HOWARD
HQ OPS Officer: LEIGH TROCINE
Notification Date: 05/19/2002
Notification Time: 06:03 [ET]
Event Date: 05/19/2002
Event Time: 04:47 [EDT]
Last Update Date: 05/19/2002
Notification Time: 06:03 [ET]
Event Date: 05/19/2002
Event Time: 04:47 [EDT]
Last Update Date: 05/19/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
PAUL FREDRICKSON (R2)
PAUL FREDRICKSON (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 0 | Startup | 0 | Hot Standby |
MANUAL REACTOR TRIP DUE TO A CONTROL ROD SYSTEM URGENT FAILURE ALARM DURING LOW POWER PHYSICS TESTING
While in Mode 2 (Startup) with reactor power at 10E-6 in the intermediate range as part of a refueling outage, Unit 2 was manually tripped at 0447 EDT on 05/19/02. During the performance of low power physics testing, the unit experienced a control rod system urgent failure alarm. Operators were inserting Shutdown Bank 'B,' and the Shutdown Bank 'B' Group 2 step counter failed to step. Rod movement was stopped. After performance of an evaluation, operators noticed that reactivity was increasing slightly. Operators then attempted to insert Control Bank 'D,' and the Control Bank 'D' Group 2 step counter also failed to step. After evaluating rod control problems, the reactor was manually tripped.
All control rods fully inserted as a result of the manual reactor trip, and all systems functioned as required. There were no engineered safety feature actuations or emergency core cooling system injections, and none were required. The licensee currently suspects that there may be a blown fuse in the rod control cabinet. The investigation is currently underway.
The unit is currently stable in Mode 3 (Hot Standby). Normal charging and makeup, pressurizer heaters and sprays, and the reactor coolant pumps are being utilized for primary system inventory, pressure, and transport control. Main feedwater is supplying water to the steam generators, and secondary steam is being dumped to the condenser. Containment parameters are also normal.
The licensee notified the NRC resident inspector.
While in Mode 2 (Startup) with reactor power at 10E-6 in the intermediate range as part of a refueling outage, Unit 2 was manually tripped at 0447 EDT on 05/19/02. During the performance of low power physics testing, the unit experienced a control rod system urgent failure alarm. Operators were inserting Shutdown Bank 'B,' and the Shutdown Bank 'B' Group 2 step counter failed to step. Rod movement was stopped. After performance of an evaluation, operators noticed that reactivity was increasing slightly. Operators then attempted to insert Control Bank 'D,' and the Control Bank 'D' Group 2 step counter also failed to step. After evaluating rod control problems, the reactor was manually tripped.
All control rods fully inserted as a result of the manual reactor trip, and all systems functioned as required. There were no engineered safety feature actuations or emergency core cooling system injections, and none were required. The licensee currently suspects that there may be a blown fuse in the rod control cabinet. The investigation is currently underway.
The unit is currently stable in Mode 3 (Hot Standby). Normal charging and makeup, pressurizer heaters and sprays, and the reactor coolant pumps are being utilized for primary system inventory, pressure, and transport control. Main feedwater is supplying water to the steam generators, and secondary steam is being dumped to the condenser. Containment parameters are also normal.
The licensee notified the NRC resident inspector.
Power Reactor
Event Number: 38929
Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: STEVE DAVID
HQ OPS Officer: JOHN MacKINNON
Region: 4 State: CA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: STEVE DAVID
HQ OPS Officer: JOHN MacKINNON
Notification Date: 05/19/2002
Notification Time: 20:59 [ET]
Event Date: 05/19/2002
Event Time: 16:00 [PDT]
Last Update Date: 05/19/2002
Notification Time: 20:59 [ET]
Event Date: 05/19/2002
Event Time: 16:00 [PDT]
Last Update Date: 05/19/2002
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):
GREG PICK (R4)
GREG PICK (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
GREATER THAN 1% OF STEAM GENERATOR 1-2 TUBES ARE DEFECTIVE
On May 19, 2002, during the Unit 1 eleventh refueling outage, analysis of eddy current data on steam generator (SG) 1-2 indicated that greater than one percent of the total tubes inspected on SG 1-2 were defective. Thirty-six (36) defective tubes in SG 1-2 were detected and are being plugged. Most of the pluggable indications are due to outside diameter stress corrosion cracking at the tube support plate intersections. Review of the eddy current inspections are complete. Results of the SG tube inspection fall into Category C-3, which requires a 4-hour non-emergency report in accordance with Technical Specification Table 5.5.9-2 and 10 CFR 50.72(b)(3)(ii) an 8-hour requirement.
The other three SGs' tubes have been checked.
An NRC teleconference in accordance with Technical Specification 5.6.10.d to report the results of the voltage-based repair criteria implemented for the tube support plate intersections is currently scheduled for Monday May 20, 2002, at 1230 hours EDT.
The NRC Resident Inspector will be notified of this event by the licensee.
On May 19, 2002, during the Unit 1 eleventh refueling outage, analysis of eddy current data on steam generator (SG) 1-2 indicated that greater than one percent of the total tubes inspected on SG 1-2 were defective. Thirty-six (36) defective tubes in SG 1-2 were detected and are being plugged. Most of the pluggable indications are due to outside diameter stress corrosion cracking at the tube support plate intersections. Review of the eddy current inspections are complete. Results of the SG tube inspection fall into Category C-3, which requires a 4-hour non-emergency report in accordance with Technical Specification Table 5.5.9-2 and 10 CFR 50.72(b)(3)(ii) an 8-hour requirement.
The other three SGs' tubes have been checked.
An NRC teleconference in accordance with Technical Specification 5.6.10.d to report the results of the voltage-based repair criteria implemented for the tube support plate intersections is currently scheduled for Monday May 20, 2002, at 1230 hours EDT.
The NRC Resident Inspector will be notified of this event by the licensee.
Power Reactor
Event Number: 39034
Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: CLARE BLEAU
HQ OPS Officer: JOHN MacKINNON
Region: 3 State: IA
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: CLARE BLEAU
HQ OPS Officer: JOHN MacKINNON
Notification Date: 07/02/2002
Notification Time: 10:47 [ET]
Event Date: 05/19/2002
Event Time: 11:00 [CDT]
Last Update Date: 07/02/2002
Notification Time: 10:47 [ET]
Event Date: 05/19/2002
Event Time: 11:00 [CDT]
Last Update Date: 07/02/2002
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):
BRUCE JORGENSEN (R3)
BRUCE JORGENSEN (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 21 | Power Operation | 0 | Cold Shutdown |
RADIATION MONITOR DECLARED INOPERABLE
While operating in Mode 1, in the process of a controlled shutdown to perform an elective repair of a condenser tube leak, Division "A" Refuel Pool Radiation Monitor lost power resulting in a Group 3A Primary Containment Isolation System (PCIS) actuation. The Group 3A isolation function as designed. Radiation levels in the area of the monitor were normal and the actuation was considered invalid.
The Radiation Monitor was declared inoperable for maintenance. The monitor was declared operable at 1049 CST on 05/20/02 at which time the Group III isolation was also reset.
The NRC Resident Inspector was notified of this event by the licensee.
While operating in Mode 1, in the process of a controlled shutdown to perform an elective repair of a condenser tube leak, Division "A" Refuel Pool Radiation Monitor lost power resulting in a Group 3A Primary Containment Isolation System (PCIS) actuation. The Group 3A isolation function as designed. Radiation levels in the area of the monitor were normal and the actuation was considered invalid.
The Radiation Monitor was declared inoperable for maintenance. The monitor was declared operable at 1049 CST on 05/20/02 at which time the Group III isolation was also reset.
The NRC Resident Inspector was notified of this event by the licensee.