Event Notification Report for September 18, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
09/17/2000 - 09/18/2000
EVENT NUMBERS
37341373423733837339373443734537346
Power Reactor
Event Number: 37341
Facility: WATERFORD
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: OSCAR PIPKINS
HQ OPS Officer: BOB STRANSKY
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: OSCAR PIPKINS
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/18/2000
Notification Time: 13:41 [ET]
Event Date: 09/18/2000
Event Time: 11:50 [CDT]
Last Update Date: 09/18/2000
Notification Time: 13:41 [ET]
Event Date: 09/18/2000
Event Time: 11:50 [CDT]
Last Update Date: 09/18/2000
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):
JEFF SHACKELFORD (R4)
JEFF SHACKELFORD (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
UNIT OUTSIDE DESIGN BASIS OF FIRE PROTECTION PROGRAM
"During review of the plant Cable and Conduit List (CCL) database, it was observed that six essential cables were not listed as requiring fire wrap and the cables are not wrapped in the field. The cable, with one exception, is listed in the safe shutdown analysis as required for hot and cold safe shutdown, but are not provided with Appendix R separation per 10CFR50 Appendix R Section IIIG. The cables are located in Room 299 on the +46 ft elevation of the Reactor Auxiliary Bldg (RAB) in Fire Area RAB 2. The condition has existed since plant original construction. At the time of discovery, the plant is operating at 100% Reactor Power. The cables were immediately listed as impaired, and compensatory actions were established (hourly fire watches). The condition has been entered into the plant corrective action program (CR-WF3-2000-1088). The equipment would be affected by a hypothetical fire in that fire area include the A/B Essential Chillers, RAB HVAG Exhaust Fan A, RAB HVAC Equipment Room Supply Fan AH-13 (3A-SA) Dampers. These components should be the protected components in the fire area. The condition affects the ability to achieve safe shutdown. Therefore, the condition is considered to be outside the design basis of the fire protection program."
The NRC resident inspector has been informed of this event by the licensee.
"During review of the plant Cable and Conduit List (CCL) database, it was observed that six essential cables were not listed as requiring fire wrap and the cables are not wrapped in the field. The cable, with one exception, is listed in the safe shutdown analysis as required for hot and cold safe shutdown, but are not provided with Appendix R separation per 10CFR50 Appendix R Section IIIG. The cables are located in Room 299 on the +46 ft elevation of the Reactor Auxiliary Bldg (RAB) in Fire Area RAB 2. The condition has existed since plant original construction. At the time of discovery, the plant is operating at 100% Reactor Power. The cables were immediately listed as impaired, and compensatory actions were established (hourly fire watches). The condition has been entered into the plant corrective action program (CR-WF3-2000-1088). The equipment would be affected by a hypothetical fire in that fire area include the A/B Essential Chillers, RAB HVAG Exhaust Fan A, RAB HVAC Equipment Room Supply Fan AH-13 (3A-SA) Dampers. These components should be the protected components in the fire area. The condition affects the ability to achieve safe shutdown. Therefore, the condition is considered to be outside the design basis of the fire protection program."
The NRC resident inspector has been informed of this event by the licensee.
Power Reactor
Event Number: 37342
Facility: GRAND GULF
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: CARL EHRHARDT
HQ OPS Officer: BOB STRANSKY
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: CARL EHRHARDT
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/18/2000
Notification Time: 15:26 [ET]
Event Date: 09/18/2000
Event Time: 14:15 [CDT]
Last Update Date: 09/18/2000
Notification Time: 15:26 [ET]
Event Date: 09/18/2000
Event Time: 14:15 [CDT]
Last Update Date: 09/18/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(A) - UNANALYZED COND OP
10 CFR Section:
50.72(b)(1)(ii)(A) - UNANALYZED COND OP
Person (Organization):
JEFF SHACKELFORD (R4)
JEFF SHACKELFORD (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 13 | Power Operation | 13 | Power Operation |
MAIN GENERATOR PARTIAL LOAD REJECT MAY NOT CAUSE REACTOR SCRAM AS ASSUMED
"During evaluation of the data from a recent Reactor scram (Event #37325 on 9/15/00) a possible unanalyzed condition was found. Specifically, a Main Generator partial Load Reject can actuate a circuit contained in the turbine controls that may not always activate a reactor scram/recirculation pump down shift as required by analysis. Theoretically, with the partial load reject that fails to actuate the Reactor Scram/Recirculation pump trip and assuming failure of the bypass valves, MCPR limits could be affected. Immediate corrective action is to change the setpoint of the EHC trip to ensure that a partial load reject will result in a scram and Recirculation pump transfer to slow speed."
The NRC resident inspector has been informed of this notification by the licensee.
"During evaluation of the data from a recent Reactor scram (Event #37325 on 9/15/00) a possible unanalyzed condition was found. Specifically, a Main Generator partial Load Reject can actuate a circuit contained in the turbine controls that may not always activate a reactor scram/recirculation pump down shift as required by analysis. Theoretically, with the partial load reject that fails to actuate the Reactor Scram/Recirculation pump trip and assuming failure of the bypass valves, MCPR limits could be affected. Immediate corrective action is to change the setpoint of the EHC trip to ensure that a partial load reject will result in a scram and Recirculation pump transfer to slow speed."
The NRC resident inspector has been informed of this notification by the licensee.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37338
Facility: VOGTLE
Region: 2 State: GA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEITH POPE
HQ OPS Officer: DOUG WEAVER
Region: 2 State: GA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEITH POPE
HQ OPS Officer: DOUG WEAVER
Notification Date: 09/18/2000
Notification Time: 06:16 [ET]
Event Date: 09/18/2000
Event Time: 04:50 [EDT]
Last Update Date: 10/11/2000
Notification Time: 06:16 [ET]
Event Date: 09/18/2000
Event Time: 04:50 [EDT]
Last Update Date: 10/11/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
Person (Organization):
STEPHEN CAHILL (R2)
STEPHEN CAHILL (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
BOTH TRAINS OF RHR (RECIRC CAPABILITY) INOPERABLE FOR APPROXIMATELY SEVEN HOURS
"While placing RHR Train 'B' in shutdown cooling during a refueling outage the pump minimum flow valve did not close as expected. Investigation revealed that the mini-flow transmitter 1 FIS-0611 was isolated causing the valve 1 FV0611 to remain open with RHR flow in excess of setpoint that would cause the valve to close. Work was performed on August 29, 2000, to calibrate this transmitter. On September 9, 2000, RHR Train 'A' seal cooler was isolated for maintenance rendering both trains of RHR inoperable for 6 hours and 54 minutes. The Cold leg injection capability from RHR Train 'A' remained available during this time. The Cold leg recirculation capability during these inoperabilities was not available. This would have required entry into T. S. 3.0.3.
"RHR Train 'B' mini-flow transmitter 1FIS-0611 was discovered isolated causing the valve 1FV0611 to remain open with RHR flow in excess of setpoint that would cause the valve to close. The transmitter has been returned to service and the valve 1 FV0611 is operating properly."
The licensee notified the NRC resident inspector.
* * * RETRACTED AT 1524 EDT ON 10/11/00 BY BILL DUNN TO FANGIE JONES * * *
"This retracts the 9-18-2000 notification made per 10 CFR 50.72 (b)(2)(iii)(D). That notification described an event where both trains of RHR were inoperable for 6 hours and 54 minutes when one train was in a system outage and the inservice train was found to be inoperable. Full flow to the RCS cold legs was apparently not available because it was determined that the inservice train's miniflow line valve would not close had a design basis accident occurred. Design engineering reviews have revealed that, with the mini-flow valve open, the inservice RHR train remained capable of providing minimum flow to the RCS for analyzed accident scenarios. Therefore, because one train of RHR remained operable, a condition did not exist that would have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident."
The licensee notified the NRC Resident Inspector. The R2DO (Ken Barr) has been notified.
"While placing RHR Train 'B' in shutdown cooling during a refueling outage the pump minimum flow valve did not close as expected. Investigation revealed that the mini-flow transmitter 1 FIS-0611 was isolated causing the valve 1 FV0611 to remain open with RHR flow in excess of setpoint that would cause the valve to close. Work was performed on August 29, 2000, to calibrate this transmitter. On September 9, 2000, RHR Train 'A' seal cooler was isolated for maintenance rendering both trains of RHR inoperable for 6 hours and 54 minutes. The Cold leg injection capability from RHR Train 'A' remained available during this time. The Cold leg recirculation capability during these inoperabilities was not available. This would have required entry into T. S. 3.0.3.
"RHR Train 'B' mini-flow transmitter 1FIS-0611 was discovered isolated causing the valve 1FV0611 to remain open with RHR flow in excess of setpoint that would cause the valve to close. The transmitter has been returned to service and the valve 1 FV0611 is operating properly."
The licensee notified the NRC resident inspector.
* * * RETRACTED AT 1524 EDT ON 10/11/00 BY BILL DUNN TO FANGIE JONES * * *
"This retracts the 9-18-2000 notification made per 10 CFR 50.72 (b)(2)(iii)(D). That notification described an event where both trains of RHR were inoperable for 6 hours and 54 minutes when one train was in a system outage and the inservice train was found to be inoperable. Full flow to the RCS cold legs was apparently not available because it was determined that the inservice train's miniflow line valve would not close had a design basis accident occurred. Design engineering reviews have revealed that, with the mini-flow valve open, the inservice RHR train remained capable of providing minimum flow to the RCS for analyzed accident scenarios. Therefore, because one train of RHR remained operable, a condition did not exist that would have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident."
The licensee notified the NRC Resident Inspector. The R2DO (Ken Barr) has been notified.
Power Reactor
Event Number: 37339
Facility: GINNA
Region: 1 State: NY
Unit: [1] [] []
RX Type: [1] W-2-LP
NRC Notified By: MICKLOW
HQ OPS Officer: DOUG WEAVER
Region: 1 State: NY
Unit: [1] [] []
RX Type: [1] W-2-LP
NRC Notified By: MICKLOW
HQ OPS Officer: DOUG WEAVER
Notification Date: 09/18/2000
Notification Time: 06:47 [ET]
Event Date: 09/18/2000
Event Time: 05:03 [EDT]
Last Update Date: 09/18/2000
Notification Time: 06:47 [ET]
Event Date: 09/18/2000
Event Time: 05:03 [EDT]
Last Update Date: 09/18/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
JAMES TRAPP (R1)
JAMES TRAPP (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | N | 0 | Startup | 0 | Hot Standby |
AUTOMATIC REACTOR TRIP DURING PLANT SHUTDOWN
"A plant shutdown was in progress for a refueling outage, RCS pressure was at 2236 psig, Tavg at 541F. Power as indicated on intermediate range channels N35 & N36 was 4 x10 minus 12 amps and 7.5 x 10 minus 11 amps, respectively. Control room operator was in the process of driving control rods to 5 steps on Bank A, when P-6 permissive reset. Coincident with P-6 reset, a control power fuse blew on IR channel N-36, causing a Rx trip from 1/2 IR channels high. All rods fully inserted on the trip. Cause of the blown fuse is under investigation and plant shutdown will continue using normal operating procedures."
The licensee notified the NRC resident inspector.
"A plant shutdown was in progress for a refueling outage, RCS pressure was at 2236 psig, Tavg at 541F. Power as indicated on intermediate range channels N35 & N36 was 4 x10 minus 12 amps and 7.5 x 10 minus 11 amps, respectively. Control room operator was in the process of driving control rods to 5 steps on Bank A, when P-6 permissive reset. Coincident with P-6 reset, a control power fuse blew on IR channel N-36, causing a Rx trip from 1/2 IR channels high. All rods fully inserted on the trip. Cause of the blown fuse is under investigation and plant shutdown will continue using normal operating procedures."
The licensee notified the NRC resident inspector.
Power Reactor
Event Number: 37344
Facility: COOK
Region: 3 State: MI
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: SCOTT RICHARDSON
HQ OPS Officer: BOB STRANSKY
Region: 3 State: MI
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: SCOTT RICHARDSON
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/18/2000
Notification Time: 19:51 [ET]
Event Date: 09/18/2000
Event Time: 17:10 [EDT]
Last Update Date: 09/18/2000
Notification Time: 19:51 [ET]
Event Date: 09/18/2000
Event Time: 17:10 [EDT]
Last Update Date: 09/18/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
Person (Organization):
BRUCE BURGESS (R3)
BRUCE BURGESS (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
ECCS PERFORMANCE POTENTIALLY IMPAIRED DUE TO VALVE CLOSURE DURING TESTING
"During review of emergency core cooling system (ECCS) valve operability test procedures, it was determined that, on one occasion in 1995 and again in 1997, one safety injection (SI) system containment isolation valve (CIV) was closed for valve operability testing with reactor power at or near 100%. With a CIV closed, injection from the two SI pumps following a small break loss of coolant accident (SBLOCA) would be directed to only two of the four reactor coolant system loops. At this power level, the high head SI system function of maintaining peak clad temperature (PCT) below 2200°F following a SBLOCA may not be satisfied. The condition becomes worse if a single failure is considered to affect one of the two SI pumps. Further evaluation of the condition is in progress."
The NRC resident inspector has been informed of this notification by the licensee.
"During review of emergency core cooling system (ECCS) valve operability test procedures, it was determined that, on one occasion in 1995 and again in 1997, one safety injection (SI) system containment isolation valve (CIV) was closed for valve operability testing with reactor power at or near 100%. With a CIV closed, injection from the two SI pumps following a small break loss of coolant accident (SBLOCA) would be directed to only two of the four reactor coolant system loops. At this power level, the high head SI system function of maintaining peak clad temperature (PCT) below 2200°F following a SBLOCA may not be satisfied. The condition becomes worse if a single failure is considered to affect one of the two SI pumps. Further evaluation of the condition is in progress."
The NRC resident inspector has been informed of this notification by the licensee.
Power Reactor
Event Number: 37345
Facility: WNP-2
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: J. W. GASTON
HQ OPS Officer: BOB STRANSKY
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: J. W. GASTON
HQ OPS Officer: BOB STRANSKY
Notification Date: 09/18/2000
Notification Time: 19:56 [ET]
Event Date: 09/18/2000
Event Time: 15:09 [PDT]
Last Update Date: 09/18/2000
Notification Time: 19:56 [ET]
Event Date: 09/18/2000
Event Time: 15:09 [PDT]
Last Update Date: 09/18/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
JEFF SHACKELFORD (R4)
JEFF SHACKELFORD (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 65 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR SCRAM DUE TO DECREASING CONDENSER VACUUM
"At 1425 PDT the plant experienced a sudden increase in offgas flow and decreasing condenser vacuum. Reactor power was reduced from 100% to 65% in accordance with the plant abnormal procedures for the loss of condenser vacuum. When the condenser back pressure continued to rise, a manual scram was inserted. All safety systems functioned as designed. The plant is currently stable in mode 3. The cause of the increase in off-gas flow is currently under investigation."
The NRC resident inspector has been informed of this event by the licensee.
"At 1425 PDT the plant experienced a sudden increase in offgas flow and decreasing condenser vacuum. Reactor power was reduced from 100% to 65% in accordance with the plant abnormal procedures for the loss of condenser vacuum. When the condenser back pressure continued to rise, a manual scram was inserted. All safety systems functioned as designed. The plant is currently stable in mode 3. The cause of the increase in off-gas flow is currently under investigation."
The NRC resident inspector has been informed of this event by the licensee.
Power Reactor
Event Number: 37346
Facility: DRESDEN
Region: 3 State: IL
Unit: [] [] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: BRIAN GRANT
HQ OPS Officer: DOUG WEAVER
Region: 3 State: IL
Unit: [] [] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: BRIAN GRANT
HQ OPS Officer: DOUG WEAVER
Notification Date: 09/19/2000
Notification Time: 01:26 [ET]
Event Date: 09/18/2000
Event Time: 22:30 [CDT]
Last Update Date: 09/20/2000
Notification Time: 01:26 [ET]
Event Date: 09/18/2000
Event Time: 22:30 [CDT]
Last Update Date: 09/20/2000
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):
BRUCE BURGESS (R3)
BRUCE BURGESS (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | N | 0 | Refueling | 0 | Refueling |
DEGRADED WHILE SHUTDOWN - FAILED CONTAINMENT LEAK RATE TEST
System Engineering reports the local leak rate test results for 3A Feedwater Header Inboard and Outboard Drywell Check Valves (3-0220-58A & 3-0220-62A) are indeterminate due to excessive leakage. This exceeds the containment pathway leakage values allowed by Technical Specifications Primary Containment Leakage Rate Testing Program.
The licensee will notify the NRC resident inspector.
* * * UPDATE 2216 9/20/2000 FROM BRIAN GRANT TAKEN BY BOB STRANSKY * * *
The "B" main steam inboard and outboard isolation valves also exceeded their maximum allowable leakage. The NRC resident inspector has been informed by the licensee. Notified R3DO (Burgess).
System Engineering reports the local leak rate test results for 3A Feedwater Header Inboard and Outboard Drywell Check Valves (3-0220-58A & 3-0220-62A) are indeterminate due to excessive leakage. This exceeds the containment pathway leakage values allowed by Technical Specifications Primary Containment Leakage Rate Testing Program.
The licensee will notify the NRC resident inspector.
* * * UPDATE 2216 9/20/2000 FROM BRIAN GRANT TAKEN BY BOB STRANSKY * * *
The "B" main steam inboard and outboard isolation valves also exceeded their maximum allowable leakage. The NRC resident inspector has been informed by the licensee. Notified R3DO (Burgess).