Event Notification Report for March 29, 2003
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/28/2003 - 03/29/2003
EVENT NUMBERS
39712397133971439797
Power Reactor
Event Number: 39712
Facility: HATCH
Region: 2 State: GA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: EDWEN URQUHART
HQ OPS Officer: MIKE RIPLEY
Region: 2 State: GA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: EDWEN URQUHART
HQ OPS Officer: MIKE RIPLEY
Notification Date: 03/29/2003
Notification Time: 04:44 [ET]
Event Date: 03/29/2003
Event Time: 02:55 [EST]
Last Update Date: 03/29/2003
Notification Time: 04:44 [ET]
Event Date: 03/29/2003
Event Time: 02:55 [EST]
Last Update Date: 03/29/2003
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
Person (Organization):
RUDOLPH BERNHARD (R2)
RUDOLPH BERNHARD (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 1 | Startup | 1 | Startup |
HPCI TURBINE FAILS TO COME TO SPEED DURING SURVEILLANCE TEST
"While starting up Unit Two reactor from a planned refueling/maintenance outage, the HPCI system would not pass the required tech spec surveillance when tested at 160 psig reactor pressure. The HPCI turbine failed to come up to speed due to its turbine steam control valve did not open after the turbine stop valve was opened per system operability surveillance. The system engineer and maintenance are investigating the speed control circuit at this time."
The licensee has notified the NRC Resident Inspector.
"While starting up Unit Two reactor from a planned refueling/maintenance outage, the HPCI system would not pass the required tech spec surveillance when tested at 160 psig reactor pressure. The HPCI turbine failed to come up to speed due to its turbine steam control valve did not open after the turbine stop valve was opened per system operability surveillance. The system engineer and maintenance are investigating the speed control circuit at this time."
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 39713
Facility: SALEM
Region: 1 State: NJ
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JOHN GARECHT
HQ OPS Officer: MIKE RIPLEY
Region: 1 State: NJ
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JOHN GARECHT
HQ OPS Officer: MIKE RIPLEY
Notification Date: 03/29/2003
Notification Time: 07:15 [ET]
Event Date: 03/29/2003
Event Time: 04:42 [EST]
Last Update Date: 03/29/2003
Notification Time: 07:15 [ET]
Event Date: 03/29/2003
Event Time: 04:42 [EST]
Last Update Date: 03/29/2003
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):
PAMELA HENDERSON (R1)
PAMELA HENDERSON (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 80 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP DUE TO LOSS OF SECONDARY CIRCULATING WATER PUMPS
"Salem Unit 2 was manually tripped in response to loss of circulating water pumps. Loss of circulators was caused by heavy grass intrusion from the Delaware River. At time of trip, a load reduction was in progress and power level was approximately 80%. All safety systems operated as designed during and after the reactor trip. Unit 2 is currently stable in Mode 3 at normal operating temperature and pressure. All reactor cooling pumps are in service and core cooling is being supported by auxiliary feedwater pumps and main steam dumps to the condenser. No safety systems or equipment was out of service at the time of the trip. Salem Unit 1 was also affected by the heavy river grass conditions and is currently in Mode 1 and stable at 55% reactor power."
The licensee has notified the NRC Resident Inspector.
"Salem Unit 2 was manually tripped in response to loss of circulating water pumps. Loss of circulators was caused by heavy grass intrusion from the Delaware River. At time of trip, a load reduction was in progress and power level was approximately 80%. All safety systems operated as designed during and after the reactor trip. Unit 2 is currently stable in Mode 3 at normal operating temperature and pressure. All reactor cooling pumps are in service and core cooling is being supported by auxiliary feedwater pumps and main steam dumps to the condenser. No safety systems or equipment was out of service at the time of the trip. Salem Unit 1 was also affected by the heavy river grass conditions and is currently in Mode 1 and stable at 55% reactor power."
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 39714
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [] [] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: AZ
Unit: [] [] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 03/29/2003
Notification Time: 13:41 [ET]
Event Date: 03/29/2003
Event Time: 04:00 [MST]
Last Update Date: 03/29/2003
Notification Time: 13:41 [ET]
Event Date: 03/29/2003
Event Time: 04:00 [MST]
Last Update Date: 03/29/2003
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):
BLAIR SPITZBERG (R4)
BLAIR SPITZBERG (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown |
DEGRADED CONDITION DUE TO LEAKAGE AT SMALL BORE INCONEL 600 PENETRATIONS
The following information was received from the licensee via facsimile:
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On March 29, 2003, at approximately 04:00 Mountain Standard Time (MST) engineering personnel performing preplanned visual examinations of reactor coolant System (RCS) piping in accordance with procedure requirements discovered boric acid residue on three RCS small bore Inconel 600 penetrations. One location was the RCS hot leg in-service thermowell 3JRCBTW0112HB. The visual observation was characterized as a small white trail of boron residue running down the hot leg approximately 2 inches. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or hot leg. The residue appeared dry. The other two were on pressurizer heater sleeves (nozzles) A-4 and A-18. The visual observation at these two locations was characterized as a small white buildup of boron residue around the heater sleeve as the sleeve enters the pressurizer bottom head. There does not appear to be residue running down the outside of the sleeves. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or [pressurizer]. The residue appeared dry.
"Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.14 permits no reactor coolant system (RCS) pressure boundary leakage and therefore, the discovery of leakage (presumed boric acid residue) from the hot leg thermowell and the pressurizer heater sleeves was a degradation of a principal safety barrier. Therefore, the ENS notification of this event is in accordance with 10CFR50.72(b)(3)(ii). The control room personnel entered LCO 3.4.14 Condition B and are continuing to place the plant in Mode 5. The RCS was being cooled down in preparation for Unit 3's tenth refueling outage. At the time of discovery, the RCS was at approximately 520 degrees F and 2218 psia.
"An investigation of this event will be conducted in accordance with the PVNGS [Palo Verde Nuclear Generating Station] corrective action program. The cracking of Alloy 600 components both at Palo Verde and industry-wide has been attributed to axially oriented, primary water stress corrosion cracking (PWSCC). PWSCC is not considered a significant threat to the structural integrity of the RCS boundary, thermowell, or heater sleeves as this type of cracking typically results only in small leaks.
"The bases for this conclusion is that if PWSCC occurred at Palo Verde, the cracks would be predominately axial in orientation. As in this case, the cracks would result in visibly detectable leakage that would be apparent during visual examinations, performed as part of walkdown inspections, before significant damage to the reactor coolant boundary occurred.
"Palo Verde has a program for replacing the Alloy 600 hot leg thermowells. This thermowell was scheduled for replacement during this refueling outage. The plans to replace the thermowell remain unchanged.
"A mechanical nozzle sleeve assembly (MNSA) will be installed on each of the pressurizer heater sleeves.
"No ESF actuations occurred and none were required. There were no structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were involved. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public."
The licensee has informed the NRC Resident Inspector.
The following information was received from the licensee via facsimile:
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On March 29, 2003, at approximately 04:00 Mountain Standard Time (MST) engineering personnel performing preplanned visual examinations of reactor coolant System (RCS) piping in accordance with procedure requirements discovered boric acid residue on three RCS small bore Inconel 600 penetrations. One location was the RCS hot leg in-service thermowell 3JRCBTW0112HB. The visual observation was characterized as a small white trail of boron residue running down the hot leg approximately 2 inches. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or hot leg. The residue appeared dry. The other two were on pressurizer heater sleeves (nozzles) A-4 and A-18. The visual observation at these two locations was characterized as a small white buildup of boron residue around the heater sleeve as the sleeve enters the pressurizer bottom head. There does not appear to be residue running down the outside of the sleeves. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or [pressurizer]. The residue appeared dry.
"Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.14 permits no reactor coolant system (RCS) pressure boundary leakage and therefore, the discovery of leakage (presumed boric acid residue) from the hot leg thermowell and the pressurizer heater sleeves was a degradation of a principal safety barrier. Therefore, the ENS notification of this event is in accordance with 10CFR50.72(b)(3)(ii). The control room personnel entered LCO 3.4.14 Condition B and are continuing to place the plant in Mode 5. The RCS was being cooled down in preparation for Unit 3's tenth refueling outage. At the time of discovery, the RCS was at approximately 520 degrees F and 2218 psia.
"An investigation of this event will be conducted in accordance with the PVNGS [Palo Verde Nuclear Generating Station] corrective action program. The cracking of Alloy 600 components both at Palo Verde and industry-wide has been attributed to axially oriented, primary water stress corrosion cracking (PWSCC). PWSCC is not considered a significant threat to the structural integrity of the RCS boundary, thermowell, or heater sleeves as this type of cracking typically results only in small leaks.
"The bases for this conclusion is that if PWSCC occurred at Palo Verde, the cracks would be predominately axial in orientation. As in this case, the cracks would result in visibly detectable leakage that would be apparent during visual examinations, performed as part of walkdown inspections, before significant damage to the reactor coolant boundary occurred.
"Palo Verde has a program for replacing the Alloy 600 hot leg thermowells. This thermowell was scheduled for replacement during this refueling outage. The plans to replace the thermowell remain unchanged.
"A mechanical nozzle sleeve assembly (MNSA) will be installed on each of the pressurizer heater sleeves.
"No ESF actuations occurred and none were required. There were no structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were involved. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public."
The licensee has informed the NRC Resident Inspector.
Power Reactor
Event Number: 39797
Facility: COOPER
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: COY L. BLAIR
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: COY L. BLAIR
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/25/2003
Notification Time: 10:24 [ET]
Event Date: 03/29/2003
Event Time: 01:14 [CDT]
Last Update Date: 04/25/2003
Notification Time: 10:24 [ET]
Event Date: 03/29/2003
Event Time: 01:14 [CDT]
Last Update Date: 04/25/2003
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):
CHUCK CAIN (R4)
CHUCK CAIN (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
60-DAY OPTIONAL REPORT FOR INVALID SPECIFIED SYSTEM ACTUATION
The following information was obtained from the licensee via facsimile:
"THIS IS BEING REPORTED PER 10 CFR 50.73(a)(2)(iv)(A).
"The following systems inadvertently actuated:
"Core Spray A
"Core Spray B
"#2 Diesel Generator started and came to speed
"Core Spray A and B were complete actuations.
"#2 Diesel Generator was a partial actuation since it did not load. There was no loss of offsite power or requirement to load.
"Affected equipment started and functioned successfully."
The cause of the actuation was human error when the operator performing a Reactor Protection System surveillance test mechanically actuated the wrong relay. The reactor was in refueling and flooded-up at the time of the core spray actuation. There was a minor change in skimmer surge tank level before Operations personnel were able to manually isolate core spray.
The licensee still has a Significant Condition Report (SCR) open that should address ways to minimize re-occurrence.
The licensee will be notifying the NRC Resident Inspector.
The following information was obtained from the licensee via facsimile:
"THIS IS BEING REPORTED PER 10 CFR 50.73(a)(2)(iv)(A).
"The following systems inadvertently actuated:
"Core Spray A
"Core Spray B
"#2 Diesel Generator started and came to speed
"Core Spray A and B were complete actuations.
"#2 Diesel Generator was a partial actuation since it did not load. There was no loss of offsite power or requirement to load.
"Affected equipment started and functioned successfully."
The cause of the actuation was human error when the operator performing a Reactor Protection System surveillance test mechanically actuated the wrong relay. The reactor was in refueling and flooded-up at the time of the core spray actuation. There was a minor change in skimmer surge tank level before Operations personnel were able to manually isolate core spray.
The licensee still has a Significant Condition Report (SCR) open that should address ways to minimize re-occurrence.
The licensee will be notifying the NRC Resident Inspector.