Event Notification Report for September 01, 1999
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
08/31/1999 - 09/01/1999
EVENT NUMBERS
3611936111361123610836109
Power Reactor
Event Number: 36119
Facility: SOUTH TEXAS
Region: 4 State: TX
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JOHN MILLIFF
HQ OPS Officer: WILLIAM POERTNER
Region: 4 State: TX
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JOHN MILLIFF
HQ OPS Officer: WILLIAM POERTNER
Notification Date: 09/02/1999
Notification Time: 19:51 [ET]
Event Date: 09/01/1999
Event Time: 19:49 [CDT]
Last Update Date: 09/02/1999
Notification Time: 19:51 [ET]
Event Date: 09/01/1999
Event Time: 19:49 [CDT]
Last Update Date: 09/02/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
CHARLES CAIN (R4)
CHARLES CAIN (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
LICENSEE 24 HOUR REPORT FOR VIOLATION OF OPERATING LICENSE NPF-80 (2G) Section 2.C(2)
Steam generator level channel L-0549 failed on 9/1/99 at 1529 CDT placing the unit in a one hour Limiting Condition for Operation per Technical Specifications. Action 20 of TS 3.3.2.6.d was entered and the channel bistables were tripped. It was determined the cause of the level channel failure could be manifested in the solid state protection system. At 1949 CDT the bistables were untripped to verify the operability of the solid state protection system. Unit 2 entered Technical Specification 3.0.3 for this condition. The input relay to the 'S' logic train for steam generator level channel L-0549 was determined to be the cause of the failed channel. Action 9 of Technical Specification 3.3.1.21 and action 22 of Technical Specification 3.3.2.6.b were entered at 2017 CDT to replace the failed relay. The relay was replaced and steam generator level channel L-0549 was declared operable at 2055 CDT.
The NRC resident inspector has been notified.
Steam generator level channel L-0549 failed on 9/1/99 at 1529 CDT placing the unit in a one hour Limiting Condition for Operation per Technical Specifications. Action 20 of TS 3.3.2.6.d was entered and the channel bistables were tripped. It was determined the cause of the level channel failure could be manifested in the solid state protection system. At 1949 CDT the bistables were untripped to verify the operability of the solid state protection system. Unit 2 entered Technical Specification 3.0.3 for this condition. The input relay to the 'S' logic train for steam generator level channel L-0549 was determined to be the cause of the failed channel. Action 9 of Technical Specification 3.3.1.21 and action 22 of Technical Specification 3.3.2.6.b were entered at 2017 CDT to replace the failed relay. The relay was replaced and steam generator level channel L-0549 was declared operable at 2055 CDT.
The NRC resident inspector has been notified.
Power Reactor
Event Number: 36111
Facility: WASHINGTON NUCLEAR
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: MIKE KELLER
HQ OPS Officer: BOB STRANSKY
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: MIKE KELLER
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/01/1999
Notification Time: 17:41 [ET]
Event Date: 09/01/1999
Event Time: 11:00 [PDT]
Last Update Date: 09/01/1999
Notification Time: 17:41 [ET]
Event Date: 09/01/1999
Event Time: 11:00 [PDT]
Last Update Date: 09/01/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
Person (Organization):
CHARLES CAIN (R4)
CHARLES CAIN (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 79 | Power Operation | 79 | Power Operation |
INADVERTENT ACTUATION OF EMERGENCY SIRENS DURING TESTING
"During the routine weekly testing of the emergency response sirens (continuity test of siren circuitry), WNP-2 inadvertently activated the audible offsite sirens. The sirens were activated at 1100 hours PDT, and sounded for three minutes.
"WNP-2 notified the following government agencies of the inadvertent activation of the emergency response sirens:
"Benton County Southeast Communications (911 Center)
"Franklin County Emergency Management
"United States Department of Energy (Hanford Operations-Occurrence Notification Center)
"Washington State Department of Emergency Management
"In addition, a notification was made to the local radio station responsible for emergency-broadcasts, with a media/press release to local newspapers.
"All offsite agencies and media were advised that the siren activation was inadvertent and not in response to an actual emergency. In addition, offsite agencies were instructed that no emergency response actions were required."
The NRC resident inspector has been informed of this event by the licensee.
"During the routine weekly testing of the emergency response sirens (continuity test of siren circuitry), WNP-2 inadvertently activated the audible offsite sirens. The sirens were activated at 1100 hours PDT, and sounded for three minutes.
"WNP-2 notified the following government agencies of the inadvertent activation of the emergency response sirens:
"Benton County Southeast Communications (911 Center)
"Franklin County Emergency Management
"United States Department of Energy (Hanford Operations-Occurrence Notification Center)
"Washington State Department of Emergency Management
"In addition, a notification was made to the local radio station responsible for emergency-broadcasts, with a media/press release to local newspapers.
"All offsite agencies and media were advised that the siren activation was inadvertent and not in response to an actual emergency. In addition, offsite agencies were instructed that no emergency response actions were required."
The NRC resident inspector has been informed of this event by the licensee.
Power Reactor
Event Number: 36112
Facility: SAN ONOFRE
Region: 4 State: CA
Unit: [] [2] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: CLAY WILLIAMS
HQ OPS Officer: BOB STRANSKY
Region: 4 State: CA
Unit: [] [2] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: CLAY WILLIAMS
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/01/1999
Notification Time: 19:24 [ET]
Event Date: 09/01/1999
Event Time: 15:26 [PDT]
Last Update Date: 09/01/1999
Notification Time: 19:24 [ET]
Event Date: 09/01/1999
Event Time: 15:26 [PDT]
Last Update Date: 09/01/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
CHARLES CAIN (R4)
CHARLES CAIN (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
RADIATION MONITOR CABLES DO NOT MEET 10 CFR 50 APPENDIX R SEPARATION REQUIREMENTS
"This notification is for San Onofre Units 2 and 3 and is being made In accordance with 10CFR50.72(b)(1)(ii)(B).
"Southern California Edison is in the process of implementing a change to the radiation monitoring system at Units 2 and 3. To complete this change, SCE installed a temporary panel in the control room area and removed the original panels and some cable wraps. On August 31,1999, shortly before midnight, licensee personnel noticed that several cables for Train A, Train B, and Train X (e.g., various low voltage electrical power cables and instrument signal lines) appeared to not meet the separation criteria of Regulatory Guide 1.75, 'Physical Independence of Electrical Systems.' Several of the affected cables provide actuation for Engineered Safety Features (e.g., control room isolation and containment purge isolation). The Updated Final Safety Analysis Report for Units 2 and 3 indicates these systems comply with IEEE-297, 1971, 'Criteria for Protection System for Nuclear Power Generating Stations' which is referenced by Regulatory Guide 1.75. Consequently, at 1526 PDT, SCE concluded this condition was reportable in accordance with 10CFR50.72(b)(1)(ii)(B) as a condition 'outside the design basis of the plant'.
"Shortly after identifying this condition, control room operators (1) placed Train A of the Control Room Emergency Air Cleanup System in service, and (2) closed and removed power from the containment minipurge valves. SCE determined that this condition has only existed since August 27, 1999. SCE has also confirmed that the affected components remained capable of performing their intended safety functions and, therefore concludes the safety significance of this occurrence is minimal."
The NRC resident inspector has been informed of this notification by the licensee.
"This notification is for San Onofre Units 2 and 3 and is being made In accordance with 10CFR50.72(b)(1)(ii)(B).
"Southern California Edison is in the process of implementing a change to the radiation monitoring system at Units 2 and 3. To complete this change, SCE installed a temporary panel in the control room area and removed the original panels and some cable wraps. On August 31,1999, shortly before midnight, licensee personnel noticed that several cables for Train A, Train B, and Train X (e.g., various low voltage electrical power cables and instrument signal lines) appeared to not meet the separation criteria of Regulatory Guide 1.75, 'Physical Independence of Electrical Systems.' Several of the affected cables provide actuation for Engineered Safety Features (e.g., control room isolation and containment purge isolation). The Updated Final Safety Analysis Report for Units 2 and 3 indicates these systems comply with IEEE-297, 1971, 'Criteria for Protection System for Nuclear Power Generating Stations' which is referenced by Regulatory Guide 1.75. Consequently, at 1526 PDT, SCE concluded this condition was reportable in accordance with 10CFR50.72(b)(1)(ii)(B) as a condition 'outside the design basis of the plant'.
"Shortly after identifying this condition, control room operators (1) placed Train A of the Control Room Emergency Air Cleanup System in service, and (2) closed and removed power from the containment minipurge valves. SCE determined that this condition has only existed since August 27, 1999. SCE has also confirmed that the affected components remained capable of performing their intended safety functions and, therefore concludes the safety significance of this occurrence is minimal."
The NRC resident inspector has been informed of this notification by the licensee.
Power Reactor
Event Number: 36108
Facility: BIG ROCK POINT
Region: 3 State: MI
Unit: [1] [] []
RX Type: [1] GE-1
NRC Notified By: LARRY MONSHOR
HQ OPS Officer: FANGIE JONES
Region: 3 State: MI
Unit: [1] [] []
RX Type: [1] GE-1
NRC Notified By: LARRY MONSHOR
HQ OPS Officer: FANGIE JONES
Notification Date: 09/01/1999
Notification Time: 11:20 [ET]
Event Date: 09/01/1999
Event Time: 10:44 [EDT]
Last Update Date: 09/01/1999
Notification Time: 11:20 [ET]
Event Date: 09/01/1999
Event Time: 10:44 [EDT]
Last Update Date: 09/01/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
Person (Organization):
DAVID HILLS (R3)
TAD MARSH (NRR)
DAVID HILLS (R3)
TAD MARSH (NRR)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Decommissioned | 0 | Decommissioned |
OFFSITE NOTIFICATION DUE TO SMOLDERING EMBERS
At the Big Rock Point Restoration Project, some smoldering embers in a pipeway of the recirculation pump room generated smoke. The Charlevoix Township fire department responded to site, but the on site fire brigade was able to take care of the problem. The problem was due to main steam pipe cutting activities heating the pipe insulation cloth coverings to the point of generating embers. There was no fire and no entry into emergency classification.
The licensee notified the NRC Resident Inspector.
At the Big Rock Point Restoration Project, some smoldering embers in a pipeway of the recirculation pump room generated smoke. The Charlevoix Township fire department responded to site, but the on site fire brigade was able to take care of the problem. The problem was due to main steam pipe cutting activities heating the pipe insulation cloth coverings to the point of generating embers. There was no fire and no entry into emergency classification.
The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 36109
Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [] [3] []
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: SCOTT PABIS
HQ OPS Officer: BOB STRANSKY
Region: 1 State: PA
Unit: [] [3] []
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: SCOTT PABIS
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/01/1999
Notification Time: 11:52 [ET]
Event Date: 09/01/1999
Event Time: 08:46 [EDT]
Last Update Date: 09/01/1999
Notification Time: 11:52 [ET]
Event Date: 09/01/1999
Event Time: 08:46 [EDT]
Last Update Date: 09/01/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
Person (Organization):
JOHN WHITE (R1)
JOHN WHITE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 89 | Power Operation | 89 | Power Operation |
REACTOR BUILDING VENTILATION ISOLATION
An inboard primary containment isolation system (PCIS) Group III (reactor building ventilation) isolation occurred due to an inadvertent loss of power to the 'A' reactor protection system (RPS) bus during a plant modification. All affected components responded as designed. The 'A' RPS bus was reenergized, and all components were restored. The NRC resident inspector has been informed of this event by the licensee.
An inboard primary containment isolation system (PCIS) Group III (reactor building ventilation) isolation occurred due to an inadvertent loss of power to the 'A' reactor protection system (RPS) bus during a plant modification. All affected components responded as designed. The 'A' RPS bus was reenergized, and all components were restored. The NRC resident inspector has been informed of this event by the licensee.