Event Notification Report for March 03, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/02/2000 - 03/03/2000
EVENT NUMBERS
367503675136752367533675436755
Power Reactor
Event Number: 36750
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: KEVIN SMALL / KEN HILL
HQ OPS Officer: DICK JOLLIFFE
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: KEVIN SMALL / KEN HILL
HQ OPS Officer: DICK JOLLIFFE
Notification Date: 03/03/2000
Notification Time: 10:38 [ET]
Event Date: 03/03/2000
Event Time: 09:50 [EST]
Last Update Date: 03/03/2000
Notification Time: 10:38 [ET]
Event Date: 03/03/2000
Event Time: 09:50 [EST]
Last Update Date: 03/03/2000
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a)(1)(i) - EMERGENCY DECLARED
10 CFR Section:
50.72(a)(1)(i) - EMERGENCY DECLARED
Person (Organization):
LEN WERT (R2)
TAD MARSH (NRR)
FRANK CONGEL (IRO)
KEN CIBOCH (FEMA)
LEN WERT (R2)
TAD MARSH (NRR)
FRANK CONGEL (IRO)
KEN CIBOCH (FEMA)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
UNUSUAL EVENT DECLARED DUE TO LOSS OF OFFSITE POWER TO 4 kV EMERGENCY BUSES - (REFER TO EVENT #36751 AND EVENT #36752 FOR ADDITIONAL INFORMATION.)
At 0950 on 03/03/00, with Unit 1 in a refueling outage, the licensee declared an Unusual Event due to a loss of offsite power to 4 kV emergency buses E-1 and E-2 due to a mispositioned switch during relay testing. All four emergency diesel generators (two for each unit) started; both Unit 1 EDGs are in operation powering buses E-1 and E-2. Shutdown cooling was lost but has been restored; spent fuel pool temperature is normal at 83°F. The licensee has suspended all fuel movement and is restoring power to systems that lost power. Unit 1 is stable. Unit 2, which is at 100% power, is unaffected by this event.
The licensee notified the NRC Resident Inspector and state and local officials.
* * * UPDATE ON 3/3/00 @ 1226 BY HILL TO GOULD * * *
WHILE ATTEMPTING TO COME OUT OF THE LOSS OF OFF SITE POWER CONDITION, THE PLANT HAD AN OVERCURRENT ALARM ON THE D/G-2, AND THEY LOST POWER FROM THE NORMAL AND EMERGENCY POWER SOURCES TO THE E-2 BUS AT 1200. THIS IN TURN RESULTED IN THE LOSS OF SHUT DOWN COOLING. SHUTDOWN COOLING WAS RESTORED AT 1218 WITH MINIMAL TEMPERATURE INCREASE. THEY ALSO HAD THE HALON SYSTEM ACTUATION IN THE D/G BASEMENT AREA. THE CAUSE OF THIS MAY HAVE BEEN RELATED TO THE GROUND ON THE E-2 BUS.
THE RESIDENT INSPECTOR WAS NOTIFIED. THE REG 2 RDO(WERT), NRR EO(HANNON), AND IRO(CONGEL) WERE NOTIFIED.
* * * UPDATE ON 3/3/00 @ 1615 * * * COMMISSIONER'S ASSISTANTS BRIEF
MALLETT(BRIEFER), MARSH(NRR EO), REYES(RA), EASLICK(RI), COLLINS(NRR), MILLER(IRO), WERT(RDO), WIENS(NRR), BEECHER(PA), LOHAS(SP), FAULKNER(IP), SATORIUS(EDO), BATES(SECY), CHAN(CHAIRMAN MESERVES' ASST), OLENCZ(COMM. DICUS' ASST), SHARFEY(COMM. MCGAFFIGAN'S ASST), CASTLEMAN(COMM. DIAZ'S ASST), AND McCABE(COMM. MERRIFIELD"S ASST).
* * * UPDATE ON 3/3/00 @ 1844 FROM JOLICOEUR TO GOULD * * *
NOUE TERMINATED AT 1840 DUE TO RESTORING POWER FROM NORMAL OFF-SITE SOURCES TO THE E-1 AND E-2 BUSSES.
The NRC Resident Inspector was notified.
REG 2 RDO(WERT), NRR EO(HANNON), IRO(CONGEL), AND FEMA(STEINDURF) WERE NOTIFIED.
At 0950 on 03/03/00, with Unit 1 in a refueling outage, the licensee declared an Unusual Event due to a loss of offsite power to 4 kV emergency buses E-1 and E-2 due to a mispositioned switch during relay testing. All four emergency diesel generators (two for each unit) started; both Unit 1 EDGs are in operation powering buses E-1 and E-2. Shutdown cooling was lost but has been restored; spent fuel pool temperature is normal at 83°F. The licensee has suspended all fuel movement and is restoring power to systems that lost power. Unit 1 is stable. Unit 2, which is at 100% power, is unaffected by this event.
The licensee notified the NRC Resident Inspector and state and local officials.
* * * UPDATE ON 3/3/00 @ 1226 BY HILL TO GOULD * * *
WHILE ATTEMPTING TO COME OUT OF THE LOSS OF OFF SITE POWER CONDITION, THE PLANT HAD AN OVERCURRENT ALARM ON THE D/G-2, AND THEY LOST POWER FROM THE NORMAL AND EMERGENCY POWER SOURCES TO THE E-2 BUS AT 1200. THIS IN TURN RESULTED IN THE LOSS OF SHUT DOWN COOLING. SHUTDOWN COOLING WAS RESTORED AT 1218 WITH MINIMAL TEMPERATURE INCREASE. THEY ALSO HAD THE HALON SYSTEM ACTUATION IN THE D/G BASEMENT AREA. THE CAUSE OF THIS MAY HAVE BEEN RELATED TO THE GROUND ON THE E-2 BUS.
THE RESIDENT INSPECTOR WAS NOTIFIED. THE REG 2 RDO(WERT), NRR EO(HANNON), AND IRO(CONGEL) WERE NOTIFIED.
* * * UPDATE ON 3/3/00 @ 1615 * * * COMMISSIONER'S ASSISTANTS BRIEF
MALLETT(BRIEFER), MARSH(NRR EO), REYES(RA), EASLICK(RI), COLLINS(NRR), MILLER(IRO), WERT(RDO), WIENS(NRR), BEECHER(PA), LOHAS(SP), FAULKNER(IP), SATORIUS(EDO), BATES(SECY), CHAN(CHAIRMAN MESERVES' ASST), OLENCZ(COMM. DICUS' ASST), SHARFEY(COMM. MCGAFFIGAN'S ASST), CASTLEMAN(COMM. DIAZ'S ASST), AND McCABE(COMM. MERRIFIELD"S ASST).
* * * UPDATE ON 3/3/00 @ 1844 FROM JOLICOEUR TO GOULD * * *
NOUE TERMINATED AT 1840 DUE TO RESTORING POWER FROM NORMAL OFF-SITE SOURCES TO THE E-1 AND E-2 BUSSES.
The NRC Resident Inspector was notified.
REG 2 RDO(WERT), NRR EO(HANNON), IRO(CONGEL), AND FEMA(STEINDURF) WERE NOTIFIED.
Power Reactor
Event Number: 36751
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JESTER
HQ OPS Officer: CHAUNCEY GOULD
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JESTER
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/03/2000
Notification Time: 13:25 [ET]
Event Date: 03/03/2000
Event Time: 09:31 [EST]
Last Update Date: 03/03/2000
Notification Time: 13:25 [ET]
Event Date: 03/03/2000
Event Time: 09:31 [EST]
Last Update Date: 03/03/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION 50.72(b)(2)(iii)(B) - POT RHR INOP 50.72(b)(2)(vi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION 50.72(b)(2)(iii)(B) - POT RHR INOP 50.72(b)(2)(vi) - OFFSITE NOTIFICATION
Person (Organization):
LEN WERT (R2)
LEN WERT (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
THE PLANT HAD ESF ACTUATION DURING THE LOSS OF OFFSITE POWER EVENT -
(REFER TO EVENT #36750 AND EVENT #36752 FOR ADDITIONAL INFORMATION.)
With Brunswick Unit 1 shutdown for refueling at approximately 09:31 on 3/3/00 while performing procedure TM-TTBUS1A (Trip Testing of 230KV Bus 1A) a loss of power to the Unit 1 BOP 4160 and emergency busses occurred. This caused an automatic start of 4 diesel generators with diesels number 1 and 2 loading and restoring power to their respective emergency busses. Unit 1 received a group 8 isolation signal that caused a loss of shutdown cooling. Core alterations were secured and shutdown cooling was reestablished at 09:52. A small rise in reactor coolant temperature occurred (89°F was the maximum) and the coolant system did not approach boiling temperatures. Six group isolation signals occurred with various valves closing depending on initial (pre-event) position.
The following actuations occurred: 4 Diesel generators started, Reactor Building HVAC isolated, Standby
Gas trains started, and an RPS trip occurred (no rod motion occurred due to all rods being inserted).
Systems and components performed as designed based on initial assessment of the event. An Unusual Event
was declared at 09:50 (event #36750) due to a loss of offsite power to the emergency busses.
At approximately 12:00 an actuation of the diesel building fire protection halon system occurred with an isolation of the control building ventilation system. A subsequent loss of emergency bus E2 occurred due to a trip of DG#2, which resulted in a second loss of shutdown cooling on Unit 1. Group isolations similar to those occurring initially also occurred. Shutdown cooling was reestablished at 12:18. The plant is stable and investigation into the cause of the DG trip is underway. The plant fire brigade responded and found no immediate evidence of a fire.
The NRC Resident Inspector was notified along with state and local officials.
(REFER TO EVENT #36750 AND EVENT #36752 FOR ADDITIONAL INFORMATION.)
With Brunswick Unit 1 shutdown for refueling at approximately 09:31 on 3/3/00 while performing procedure TM-TTBUS1A (Trip Testing of 230KV Bus 1A) a loss of power to the Unit 1 BOP 4160 and emergency busses occurred. This caused an automatic start of 4 diesel generators with diesels number 1 and 2 loading and restoring power to their respective emergency busses. Unit 1 received a group 8 isolation signal that caused a loss of shutdown cooling. Core alterations were secured and shutdown cooling was reestablished at 09:52. A small rise in reactor coolant temperature occurred (89°F was the maximum) and the coolant system did not approach boiling temperatures. Six group isolation signals occurred with various valves closing depending on initial (pre-event) position.
The following actuations occurred: 4 Diesel generators started, Reactor Building HVAC isolated, Standby
Gas trains started, and an RPS trip occurred (no rod motion occurred due to all rods being inserted).
Systems and components performed as designed based on initial assessment of the event. An Unusual Event
was declared at 09:50 (event #36750) due to a loss of offsite power to the emergency busses.
At approximately 12:00 an actuation of the diesel building fire protection halon system occurred with an isolation of the control building ventilation system. A subsequent loss of emergency bus E2 occurred due to a trip of DG#2, which resulted in a second loss of shutdown cooling on Unit 1. Group isolations similar to those occurring initially also occurred. Shutdown cooling was reestablished at 12:18. The plant is stable and investigation into the cause of the DG trip is underway. The plant fire brigade responded and found no immediate evidence of a fire.
The NRC Resident Inspector was notified along with state and local officials.
Power Reactor
Event Number: 36752
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: HILL
HQ OPS Officer: CHAUNCEY GOULD
Region: 2 State: NC
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: HILL
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/03/2000
Notification Time: 14:29 [ET]
Event Date: 03/03/2000
Event Time: 13:57 [EST]
Last Update Date: 03/03/2000
Notification Time: 14:29 [ET]
Event Date: 03/03/2000
Event Time: 13:57 [EST]
Last Update Date: 03/03/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS
Person (Organization):
LEN WERT (R2)
LEN WERT (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 99 | Power Operation |
REACTOR SHUT DOWN DUE TO ENTERING TS 3.0.3. LCO ACTION STATEMENT -
(REFER TO EVENT #36750 AND EVENT #36751 FOR ADDITIONAL INFORMATION.)
At 1300 hours, Unit 2 entered Technical Specification 3.0.3 due to the loss of diesel generator #2 while two off-site AC circuits were inoperable as required by Limiting Condition of Operation 3.8.1, Condition I. The loss of AC power and the diesel generator failure were previously reported in ENS #36751. At 1357 hours, Unit 2 reactor shutdown commenced from 100% power.
INITIAL SAFETY SIGNIFICANCE EVALUATION: The significance of this occurrence is minimal. Sufficient redundant equipment remains operable such that no safety functions have been lost. Plant conditions are stable at this time.
CORRECTIVE ACTION(S): Efforts to restore the inoperable equipment are in progress. Reference ENS # 36751
The NRC Resident Inspector was notified.
* * * UPDATE ON 3/3/00 @ 2008 BY BELLER TO GOULD * * *
PLANT EXITED THE 3.0.3 LCO AT 1756 WHEN THE UNIT 1 STARTUP AUX TRANSFORMER WAS USED TO POWER THE EMERGENCY BUSSES. THEY ARE STILL IN A 12-HOUR LCO ACTION STATEMENT TO RESTORE UNIT 1 UNIT AUXILLARY TRANSFORMER TO SERVICE OR BE IN HOT SHUTDOWN IN THE FOLLOWING 12 HOURS. POWER HAD BEEN REDUCED TO APPROXIMATELY 60%.
THE NRC RESIDENT WAS INFORMED BY THE LICENSEE. THE NRC OPERATIONS OFFICER NOTIFIED THE R2DO (WERT).
(REFER TO EVENT #36750 AND EVENT #36751 FOR ADDITIONAL INFORMATION.)
At 1300 hours, Unit 2 entered Technical Specification 3.0.3 due to the loss of diesel generator #2 while two off-site AC circuits were inoperable as required by Limiting Condition of Operation 3.8.1, Condition I. The loss of AC power and the diesel generator failure were previously reported in ENS #36751. At 1357 hours, Unit 2 reactor shutdown commenced from 100% power.
INITIAL SAFETY SIGNIFICANCE EVALUATION: The significance of this occurrence is minimal. Sufficient redundant equipment remains operable such that no safety functions have been lost. Plant conditions are stable at this time.
CORRECTIVE ACTION(S): Efforts to restore the inoperable equipment are in progress. Reference ENS # 36751
The NRC Resident Inspector was notified.
* * * UPDATE ON 3/3/00 @ 2008 BY BELLER TO GOULD * * *
PLANT EXITED THE 3.0.3 LCO AT 1756 WHEN THE UNIT 1 STARTUP AUX TRANSFORMER WAS USED TO POWER THE EMERGENCY BUSSES. THEY ARE STILL IN A 12-HOUR LCO ACTION STATEMENT TO RESTORE UNIT 1 UNIT AUXILLARY TRANSFORMER TO SERVICE OR BE IN HOT SHUTDOWN IN THE FOLLOWING 12 HOURS. POWER HAD BEEN REDUCED TO APPROXIMATELY 60%.
THE NRC RESIDENT WAS INFORMED BY THE LICENSEE. THE NRC OPERATIONS OFFICER NOTIFIED THE R2DO (WERT).
Power Reactor
Event Number: 36753
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [] [2] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: ROY GREEN
HQ OPS Officer: STEVE SANDIN
Region: 1 State: NY
Unit: [] [2] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: ROY GREEN
HQ OPS Officer: STEVE SANDIN
Notification Date: 03/03/2000
Notification Time: 15:56 [ET]
Event Date: 03/03/2000
Event Time: 14:17 [EST]
Last Update Date: 03/03/2000
Notification Time: 15:56 [ET]
Event Date: 03/03/2000
Event Time: 14:17 [EST]
Last Update Date: 03/03/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION 50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION 50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
Person (Organization):
WAYNE LANNING (R1)
JIM TRAPP (R1)
JOHN HANNON (NRR)
WAYNE LANNING (R1)
JIM TRAPP (R1)
JOHN HANNON (NRR)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 28 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR SCRAM FOLLOWING A LOSS OF FEEDWATER SEAL INJECTION FLOW DURING PLANT SHUTDOWN
"WHILE PERFORMING A NORMAL SHUTDOWN, THE REACTOR WAS MANUALLY SCRAMMED AT 1417EST DUE TO A LOSS OF SEAL WATER TO THE OPERATING FEEDWATER PUMP AND ITS POTENTIAL FAILURE. ALL RODS FULLY INSERTED. SUBSEQUENT TO THE MANUAL SCRAM, AN AUTOMATIC SCRAM WAS RECEIVED ON REACTOR WATER LEVEL 3. AN ATTEMPT WAS MADE TO MANUALLY INITIATE REACTOR CORE ISOLATION COOLING TO ASSIST WITH REACTOR WATER LEVEL CONTROL. THE RCIC SYSTEM FAILED TO INITIATE DUE TO A TRIPPED TRIP THROTTLE VALVE. AT 1420EST THE RCIC SYSTEM WAS DECLARED INOPERABLE. THE RCIC SYSTEM HAS BEEN RESTORED TO STANDBY. THE OPERATING FEEDWATER PUMP CURRENTLY REMAINS IN SERVICE AND SCRAM RECOVERY PROCEDURE AND REACTOR COOLDOWN ARE IN PROGRESS."
FOLLOWING THE SCRAM MINIMAL SEAL INJECTION FLOW WAS AVAILABLE TO THE OPERATING FEEDWATER PUMP. DECAY HEAT IS CURRENTLY BEING REJECTED TO THE MAIN CONDENSER.
THE NRC RESIDENT INSPECTOR WAS INFORMED.
"WHILE PERFORMING A NORMAL SHUTDOWN, THE REACTOR WAS MANUALLY SCRAMMED AT 1417EST DUE TO A LOSS OF SEAL WATER TO THE OPERATING FEEDWATER PUMP AND ITS POTENTIAL FAILURE. ALL RODS FULLY INSERTED. SUBSEQUENT TO THE MANUAL SCRAM, AN AUTOMATIC SCRAM WAS RECEIVED ON REACTOR WATER LEVEL 3. AN ATTEMPT WAS MADE TO MANUALLY INITIATE REACTOR CORE ISOLATION COOLING TO ASSIST WITH REACTOR WATER LEVEL CONTROL. THE RCIC SYSTEM FAILED TO INITIATE DUE TO A TRIPPED TRIP THROTTLE VALVE. AT 1420EST THE RCIC SYSTEM WAS DECLARED INOPERABLE. THE RCIC SYSTEM HAS BEEN RESTORED TO STANDBY. THE OPERATING FEEDWATER PUMP CURRENTLY REMAINS IN SERVICE AND SCRAM RECOVERY PROCEDURE AND REACTOR COOLDOWN ARE IN PROGRESS."
FOLLOWING THE SCRAM MINIMAL SEAL INJECTION FLOW WAS AVAILABLE TO THE OPERATING FEEDWATER PUMP. DECAY HEAT IS CURRENTLY BEING REJECTED TO THE MAIN CONDENSER.
THE NRC RESIDENT INSPECTOR WAS INFORMED.
Power Reactor
Event Number: 36754
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: PFEFFER
HQ OPS Officer: CHAUNCEY GOULD
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: PFEFFER
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/03/2000
Notification Time: 17:34 [ET]
Event Date: 03/03/2000
Event Time: 11:15 [CST]
Last Update Date: 03/03/2000
Notification Time: 17:34 [ET]
Event Date: 03/03/2000
Event Time: 11:15 [CST]
Last Update Date: 03/03/2000
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):
MIKE JORDAN (R3)
MIKE JORDAN (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
ISOLATION CONTROLS FOR THE TIP GUIDE TUBE ARE NOT DESIGNED/QUALIFIED FOR SAFETY RELATED SERVICE.
The Transverse Incore Probe (TIP) guide tube primary containment isolation controls are not designed or qualified for safety-related service. This could result in a condition outside the Design Basis of the plant because multiple failures could cause the undesirable opening of the TIP ball valves during a Design Basis Accident. However, the TIP valve and TIP Shear Valve Actuation circuits remain operable and continue to be capable of operating as containment isolation valves. The TIP Ball Valve power source is through the TIP Drive Mechanism. The TIP Ball Valves will be disabled by removing the power source from the drive Mechanism when the TIP system is not in use. The drive Mechanism power will be restored, under administrative controls, when the TIP system is required to be used.
The NRC Resident Inspector will be notified.
The Transverse Incore Probe (TIP) guide tube primary containment isolation controls are not designed or qualified for safety-related service. This could result in a condition outside the Design Basis of the plant because multiple failures could cause the undesirable opening of the TIP ball valves during a Design Basis Accident. However, the TIP valve and TIP Shear Valve Actuation circuits remain operable and continue to be capable of operating as containment isolation valves. The TIP Ball Valve power source is through the TIP Drive Mechanism. The TIP Ball Valves will be disabled by removing the power source from the drive Mechanism when the TIP system is not in use. The drive Mechanism power will be restored, under administrative controls, when the TIP system is required to be used.
The NRC Resident Inspector will be notified.
Power Reactor
Event Number: 36755
Facility: COOK
Region: 3 State: MI
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: NANCE
HQ OPS Officer: CHAUNCEY GOULD
Region: 3 State: MI
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: NANCE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/03/2000
Notification Time: 17:42 [ET]
Event Date: 03/03/2000
Event Time: 16:30 [EST]
Last Update Date: 03/03/2000
Notification Time: 17:42 [ET]
Event Date: 03/03/2000
Event Time: 16:30 [EST]
Last Update Date: 03/03/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - DEG/UNANALYZED COND
10 CFR Section:
50.72(b)(2)(i) - DEG/UNANALYZED COND
Person (Organization):
MIKE JORDAN (R3)
MIKE JORDAN (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
| 2 | N | N | 0 | Refueling | 0 | Refueling |
THE LICENSEE DETERMINED A CONDITION FOUND WHICH COULD HAVE RESULTED IN THE PLANT BEING SERIOUSLY DEGRADED.
As a result of an Engineering evaluation of all safety-related large bore piping, a number of as-built discrepancies were found with portions of the Unit 2 Residual Heat Removal, Safety Injection and Containment Spray piping systems. This analysis has concluded that these piping systems deviate from the design criteria USAS 831.1-1967, which is the code of record for the Cook Nuclear Plant.
As the design and construction of the corresponding systems in Unit 1 are similar, this notification is being made for both units despite the fact that the Unit 1 review effort has not been completed.
Though informal evaluations have concluded that the piping is unlikely to fail in a postulated seismic event, no formal analysis exists to support this conclusion. The engineering efforts to evaluate and resolve these issues are ongoing.
Both units are currently in no Technical Specification Operating mode with both cores unloaded.
The Resident Inspector will be notified.
As a result of an Engineering evaluation of all safety-related large bore piping, a number of as-built discrepancies were found with portions of the Unit 2 Residual Heat Removal, Safety Injection and Containment Spray piping systems. This analysis has concluded that these piping systems deviate from the design criteria USAS 831.1-1967, which is the code of record for the Cook Nuclear Plant.
As the design and construction of the corresponding systems in Unit 1 are similar, this notification is being made for both units despite the fact that the Unit 1 review effort has not been completed.
Though informal evaluations have concluded that the piping is unlikely to fail in a postulated seismic event, no formal analysis exists to support this conclusion. The engineering efforts to evaluate and resolve these issues are ongoing.
Both units are currently in no Technical Specification Operating mode with both cores unloaded.
The Resident Inspector will be notified.