Event Notification Report for March 10, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/09/2001 - 03/10/2001
EVENT NUMBERS
37822378233782437934
Power Reactor
Event Number: 37822
Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ARDLE KISSINGER
HQ OPS Officer: FANGIE JONES
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ARDLE KISSINGER
HQ OPS Officer: FANGIE JONES
Notification Date: 03/10/2001
Notification Time: 04:50 [ET]
Event Date: 03/10/2001
Event Time: 04:14 [EST]
Last Update Date: 03/10/2001
Notification Time: 04:50 [ET]
Event Date: 03/10/2001
Event Time: 04:14 [EST]
Last Update Date: 03/10/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
RONALD BELLAMY (R1)
RONALD BELLAMY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 21 | Power Operation | 0 | Hot Shutdown |
COURTESY NOTIFICATION FOR ENTRY INTO 10TH REFUELING
"This is a Courtesy Notification and does not fall within 10CFR50.72 Guidelines.
"Susquehanna Unit 2 was manually shutdown on 3/10/01 to enter its 10th Refueling and Inspection outage. At 1630 hours on 3/9/01, a controlled shutdown was commenced by reducing recirc flow and then inserting control rods. At 03:45 on 3/10/99, the Main Turbine was manually tripped with power at 17.5%. At 04:14 with power at 21% the reactor mode switch was taken to shutdown placing the unit in condition 3. All control rods fully inserted. Rx Water level dropped to approximately -1 inch as recorded on wide range instrumentation, and was restored using Feedwater. There were no ECCS initiations or Diesel Generator starts. No safety/relief valves lifted. There were no challenges to containment or fuel as a result of the shutdown.
"The shutdown is continuing according to schedule and is scheduled to return to service in 35 days."
The licensee notified the NRC Resident Inspector.
"This is a Courtesy Notification and does not fall within 10CFR50.72 Guidelines.
"Susquehanna Unit 2 was manually shutdown on 3/10/01 to enter its 10th Refueling and Inspection outage. At 1630 hours on 3/9/01, a controlled shutdown was commenced by reducing recirc flow and then inserting control rods. At 03:45 on 3/10/99, the Main Turbine was manually tripped with power at 17.5%. At 04:14 with power at 21% the reactor mode switch was taken to shutdown placing the unit in condition 3. All control rods fully inserted. Rx Water level dropped to approximately -1 inch as recorded on wide range instrumentation, and was restored using Feedwater. There were no ECCS initiations or Diesel Generator starts. No safety/relief valves lifted. There were no challenges to containment or fuel as a result of the shutdown.
"The shutdown is continuing according to schedule and is scheduled to return to service in 35 days."
The licensee notified the NRC Resident Inspector.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37823
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: P DECKER
HQ OPS Officer: JOHN MacKINNON
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: P DECKER
HQ OPS Officer: JOHN MacKINNON
Notification Date: 03/10/2001
Notification Time: 18:59 [ET]
Event Date: 03/10/2001
Event Time: 17:05 [CST]
Last Update Date: 04/26/2001
Notification Time: 18:59 [ET]
Event Date: 03/10/2001
Event Time: 17:05 [CST]
Last Update Date: 04/26/2001
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):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
DIFFERENTIAL PRESSURE TO OPEN THE SUPPRESSION CHAMBER TO DRYWELL VACUUM BREAKERS DISCREPANCIES WERE IDENTIFIED.
During review of Inservice Testing requirements, while the plant was in shutdown, discrepancies were identified with acceptance criteria for the torque values used to determine the equivalent differential pressure to open the suppression chamber to drywell vacuum breakers. The measured torque values indicate that the differential pressure required to fully open the suppression chamber to drywell vacuum breakers would be greater than that assumed in containment analyses. The suppression chamber to drywell vacuum breakers insure that the 2 psig external design pressure for the primary containment drywell suppression chamber and vent system is not exceeded and containment integrity is assured.
The NRC Resident Inspector will be notified of this event by the licensee.
* * * RETRACTION ON 04/23/01 AT 1805ET BY JIM MCKAY TAKEN BY MACKINNON * * *
Retraction of notification made on March 10, 2001. The notification stated that discrepancies were identified with acceptance criteria for the torque values used to determine the equivalent differential pressure to open the suppression chamber to drywell vacuum breakers. The measured values indicated a greater differential pressure required to fully open then assumed in containment analysis. Upon further evaluation and review of the containment analyses, the results of the containment analysis were effectively independent and not adversely impacted by the suppression [chamber to drywell vacuum breaker full open differential pressure assumption in the range of concern. Therefore, an unanalyzed condition that significantly degrades the plant safety did not exist.]
R3DO (Tom Kozak) has been notified.
NRC Resident Inspector will notified of this retraction by the licensee
* * * UPDATE 1000EDT ON 4/26/01 TO S. SANDIN * * *
The above retraction was updated to include information shown in brackets which had been submitted but inadvertently omitted from the report.
During review of Inservice Testing requirements, while the plant was in shutdown, discrepancies were identified with acceptance criteria for the torque values used to determine the equivalent differential pressure to open the suppression chamber to drywell vacuum breakers. The measured torque values indicate that the differential pressure required to fully open the suppression chamber to drywell vacuum breakers would be greater than that assumed in containment analyses. The suppression chamber to drywell vacuum breakers insure that the 2 psig external design pressure for the primary containment drywell suppression chamber and vent system is not exceeded and containment integrity is assured.
The NRC Resident Inspector will be notified of this event by the licensee.
* * * RETRACTION ON 04/23/01 AT 1805ET BY JIM MCKAY TAKEN BY MACKINNON * * *
Retraction of notification made on March 10, 2001. The notification stated that discrepancies were identified with acceptance criteria for the torque values used to determine the equivalent differential pressure to open the suppression chamber to drywell vacuum breakers. The measured values indicated a greater differential pressure required to fully open then assumed in containment analysis. Upon further evaluation and review of the containment analyses, the results of the containment analysis were effectively independent and not adversely impacted by the suppression [chamber to drywell vacuum breaker full open differential pressure assumption in the range of concern. Therefore, an unanalyzed condition that significantly degrades the plant safety did not exist.]
R3DO (Tom Kozak) has been notified.
NRC Resident Inspector will notified of this retraction by the licensee
* * * UPDATE 1000EDT ON 4/26/01 TO S. SANDIN * * *
The above retraction was updated to include information shown in brackets which had been submitted but inadvertently omitted from the report.
Power Reactor
Event Number: 37824
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: JAMES JOHNSTONE
HQ OPS Officer: JOHN MacKINNON
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: JAMES JOHNSTONE
HQ OPS Officer: JOHN MacKINNON
Notification Date: 03/10/2001
Notification Time: 22:29 [ET]
Event Date: 03/10/2001
Event Time: 15:15 [EST]
Last Update Date: 03/22/2001
Notification Time: 22:29 [ET]
Event Date: 03/10/2001
Event Time: 15:15 [EST]
Last Update Date: 03/22/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
THOMAS DECKER (R2)
THOMAS DECKER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 72 | Power Operation | 71 | Power Operation |
TECHNICAL SPECIFICATION (TS) SHUTDOWN
"At time 2058 (EST) on 03/10/01, Unit 2 commenced a normal rampdown to cold shutdown for refueling. The unit was in action statement that expires at 1515 on 03/11/01. The action statement of TS 3.6.1.3.b had been entered at 1515 on 03/10/01 when 2-PT-62.1 failed. This test was a routine pre-outage test on the containment equipment hatch personnel access air lock. The unit shutdown was planned for this date and time. Repairs to the failed component will be made with unit in mode 5."
The licensee thinks that the equipment hatch inner air lock door equalizing valve is leaking by.
The NRC Resident Inspector was notified of this event by the licensee.
* * * RETRACTED AT 1301 EST ON 3/22/01 BY JIM CROSSMAN TO FANGIE JONES * * *
The licensee is retracting this event notification. The original notification was conservatively made. At the time the plant had planned to shutdown for a refueling outage and it was unknown at the time if repairs could have been made within the 24 hour time clock required by technical specifications. The problem with the hatch turned out to be minor in nature and easily corrected, thus the plant would not have been required to shutdown per the technical specifications. The problem was excessive amount of hardened grease applied to the seal surface. The grease was removed and the air lock successfully passed leakage testing all within about one hour.
The licensee notified the NRC Resident Inspector. The R2DO (Mark Lesser) was notified.
"At time 2058 (EST) on 03/10/01, Unit 2 commenced a normal rampdown to cold shutdown for refueling. The unit was in action statement that expires at 1515 on 03/11/01. The action statement of TS 3.6.1.3.b had been entered at 1515 on 03/10/01 when 2-PT-62.1 failed. This test was a routine pre-outage test on the containment equipment hatch personnel access air lock. The unit shutdown was planned for this date and time. Repairs to the failed component will be made with unit in mode 5."
The licensee thinks that the equipment hatch inner air lock door equalizing valve is leaking by.
The NRC Resident Inspector was notified of this event by the licensee.
* * * RETRACTED AT 1301 EST ON 3/22/01 BY JIM CROSSMAN TO FANGIE JONES * * *
The licensee is retracting this event notification. The original notification was conservatively made. At the time the plant had planned to shutdown for a refueling outage and it was unknown at the time if repairs could have been made within the 24 hour time clock required by technical specifications. The problem with the hatch turned out to be minor in nature and easily corrected, thus the plant would not have been required to shutdown per the technical specifications. The problem was excessive amount of hardened grease applied to the seal surface. The grease was removed and the air lock successfully passed leakage testing all within about one hour.
The licensee notified the NRC Resident Inspector. The R2DO (Mark Lesser) was notified.
Power Reactor
Event Number: 37934
Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [2] [] []
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: SAARE
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: PA
Unit: [2] [] []
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: SAARE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 04/20/2001
Notification Time: 10:19 [ET]
Event Date: 03/10/2001
Event Time: 13:30 [EDT]
Last Update Date: 04/20/2001
Notification Time: 10:19 [ET]
Event Date: 03/10/2001
Event Time: 13:30 [EDT]
Last Update Date: 04/20/2001
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):
WILLIAM RULAND (R1)
WILLIAM RULAND (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
INVALID ACTUATION OF THE "B" EMERGENCY SERVICE WATER PUMP OCCURRED
This 60-day optional report, as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10CFR50.73(a)(2)(iv)(A) to describe an invalid actuation of a specified system, specifically Emergency Service Water (ESW).
On March 10, 2001, during performance of a surveillance test on the Emergency Diesel Generator Logic, an invalid actuation of the "B" ESW pump occurred. The event was caused by a jumper inadvertently being left installed when the surveillance test required the jumper to be momentarily installed. The jumper simulated an initiation signal on 1 EDG. As the test progressed, the "B" ESW pump received an initiation signal and started.
The "B" ESW pump started as expected for the given conditions and was subsequently returned to standby per plant procedures. No deficiencies were noted during operation of the "B" ESW pump.
This event has been entered into the site-specific corrective action program for resolution.
The resident has been informed.
This 60-day optional report, as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10CFR50.73(a)(2)(iv)(A) to describe an invalid actuation of a specified system, specifically Emergency Service Water (ESW).
On March 10, 2001, during performance of a surveillance test on the Emergency Diesel Generator Logic, an invalid actuation of the "B" ESW pump occurred. The event was caused by a jumper inadvertently being left installed when the surveillance test required the jumper to be momentarily installed. The jumper simulated an initiation signal on 1 EDG. As the test progressed, the "B" ESW pump received an initiation signal and started.
The "B" ESW pump started as expected for the given conditions and was subsequently returned to standby per plant procedures. No deficiencies were noted during operation of the "B" ESW pump.
This event has been entered into the site-specific corrective action program for resolution.
The resident has been informed.