Event Notification Report for December 15, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
12/14/2001 - 12/15/2001
Power Reactor
Event Number: 38574
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [] [2] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: TOM RESTUCCIO
HQ OPS Officer: JOHN MacKINNON
Region: 1 State: NY
Unit: [] [2] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: TOM RESTUCCIO
HQ OPS Officer: JOHN MacKINNON
Notification Date: 12/15/2001
Notification Time: 23:00 [ET]
Event Date: 12/15/2001
Event Time: 19:00 [EST]
Last Update Date: 12/15/2001
Notification Time: 23:00 [ET]
Event Date: 12/15/2001
Event Time: 19:00 [EST]
Last Update Date: 12/15/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
CLIFFORD ANDERSON (R1)
CLIFFORD ANDERSON (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 60 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR SCRAM FROM 60% POWER DUE TO UNIDENTIFIED LEAK RATE EXCEEDING 5 GPM
At 0431 on December 15, 2001, Unit 2 noted a step increase in unidentified drywell leakage from 0.2 gpm (normal) to 0.4 gpm. Coincident with this step increase, containment atmosphere radiation levels rose and alarmed on the particulate channels. Throughout the day, investigation continued into possible causes, and the trend in leakage showed no appreciable increase. At 1658, containment atmosphere radiation levels rose again, and a large step increase in leakage was noted from approximately 0.5 gpm to 1.7 gpm. The technical specification limit is less than 5 gpm unidentified leakage and less than a 2-gpm rise in unidentified leakage during the previous the 24 hours. At 1745, plant management ordered an orderly plant shutdown to identify the source of leakage in the drywell. At 1900, a normal reactor shutdown was commenced, original reactor power was 100%, in preparation for repairing the leak in the drywell. At 1919, drywell unidentified leakage rose to 2.27 gpm, and the technical specification action statement was entered for greater than a 2-gpm rise in unidentified leakage during the previous 24 hours.
At 2040, the drywell unidentified leak rate exceeded 5 gpm (5.87 gpm), and the decision was made to scram the plant. At 2046, the reactor mode switch was placed in shutdown, all systems functioned as required, all rods fully inserted into the core. The scram signal received was Average Power Range Monitor (APRM) neutron upscale due to setpoint setdown with the mode switch out of run. A secondary scram signal was received when reactor water level dropped to less than 159.3" (Level 3) (expected due to level shrink on the scram). Level 3 also provided a Primary Containment Isolation System (PCIS) signal to group 4 (Residual Heat Removal System, [RHS] sampling valves) and group 5 (RHS shutdown cooling valves). At the time of the signal, all group 4 and 5 valves were closed. At 2043, technical specification action statement associated with exceeding 5 gpm was exited. At 2149, technical specification action statement associated with greater than 2 gpm rise over the previous 24 hours was exited.
At this time, drywell unidentified leakage 1.5 gpm and lowering, plant pressure is being reduced. Reactor water level is being maintained with condensate/feedwater. Reactor pressure control is being maintained via turbine bypass valves with cooldown in progress. All Emergency Core Cooling Systems and the Emergency Diesel Generator are fully operable, and the electrical grid is stable.
The NRC resident inspector was notified of this event by the licensee.
At 0431 on December 15, 2001, Unit 2 noted a step increase in unidentified drywell leakage from 0.2 gpm (normal) to 0.4 gpm. Coincident with this step increase, containment atmosphere radiation levels rose and alarmed on the particulate channels. Throughout the day, investigation continued into possible causes, and the trend in leakage showed no appreciable increase. At 1658, containment atmosphere radiation levels rose again, and a large step increase in leakage was noted from approximately 0.5 gpm to 1.7 gpm. The technical specification limit is less than 5 gpm unidentified leakage and less than a 2-gpm rise in unidentified leakage during the previous the 24 hours. At 1745, plant management ordered an orderly plant shutdown to identify the source of leakage in the drywell. At 1900, a normal reactor shutdown was commenced, original reactor power was 100%, in preparation for repairing the leak in the drywell. At 1919, drywell unidentified leakage rose to 2.27 gpm, and the technical specification action statement was entered for greater than a 2-gpm rise in unidentified leakage during the previous 24 hours.
At 2040, the drywell unidentified leak rate exceeded 5 gpm (5.87 gpm), and the decision was made to scram the plant. At 2046, the reactor mode switch was placed in shutdown, all systems functioned as required, all rods fully inserted into the core. The scram signal received was Average Power Range Monitor (APRM) neutron upscale due to setpoint setdown with the mode switch out of run. A secondary scram signal was received when reactor water level dropped to less than 159.3" (Level 3) (expected due to level shrink on the scram). Level 3 also provided a Primary Containment Isolation System (PCIS) signal to group 4 (Residual Heat Removal System, [RHS] sampling valves) and group 5 (RHS shutdown cooling valves). At the time of the signal, all group 4 and 5 valves were closed. At 2043, technical specification action statement associated with exceeding 5 gpm was exited. At 2149, technical specification action statement associated with greater than 2 gpm rise over the previous 24 hours was exited.
At this time, drywell unidentified leakage 1.5 gpm and lowering, plant pressure is being reduced. Reactor water level is being maintained with condensate/feedwater. Reactor pressure control is being maintained via turbine bypass valves with cooldown in progress. All Emergency Core Cooling Systems and the Emergency Diesel Generator are fully operable, and the electrical grid is stable.
The NRC resident inspector was notified of this event by the licensee.
Power Reactor
Event Number: 38575
Facility: PERRY
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: ALAN RABENOLD
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: ALAN RABENOLD
HQ OPS Officer: LEIGH TROCINE
Notification Date: 12/16/2001
Notification Time: 00:30 [ET]
Event Date: 12/15/2001
Event Time: 22:28 [EST]
Last Update Date: 12/17/2001
Notification Time: 00:30 [ET]
Event Date: 12/15/2001
Event Time: 22:28 [EST]
Last Update Date: 12/17/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
ANNE MARIE STONE (R3)
ANNE MARIE STONE (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown |
THE REACTOR RECIRCULATION PUMPS DOWN SHIFTED TO SLOW SPEED FOR UNKNOWN REASONS RESULTING IN AN AUTOMATIC REACTOR SCRAM, BALANCE-OF-PLANT ISOLATIONS, AND EMERGENCY CORE COOLING SYSTEM ACTUATIONS.
At 2228 EST on 12/15/01, the reactor recirculation pumps down shifted to slow speed for unknown reasons. This caused reactor pressure vessel (RPV) water level to swell, and an automatic reactor scram from 100% power occurred when Level 8 was reached. All control rods fully inserted. Following the scram, RPV water level started to decrease because the feedwater pumps also tripped (as designed) when Level 8 was reached. When RPV water level dropped to Level 2 (130 inches), balance-of-plant isolations occurred, and all of the applicable valves properly isolated. In addition, the high pressure core spray (HPCS) and reactor core isolation cooling (RCIC) systems automatically actuated when Level 2 was reached. The HPCS and RCIC systems were utilized to restore RPV water level to the normal range.
The unit is currently stable in Mode 3 with RPV pressure at 900 psi and RPV water level at 210 inches. Normal feedwater is being utilized to maintain RPV water level within the proper band, and the HPCS and RCIC systems have been secured. The main steam isolation valves remained open, and the condenser is being utilized as the heat sink. Containment parameters are currently normal, the electrical grid is stable, and the emergency diesel generators are available.
The licensee stated that all systems functioned as required and that there was nothing unusual or misunderstood other than the cause of the initiating event (the recirculation pumps down shifting to slow speed). The licensee's investigation is underway.
The licensee notified the NRC resident inspector.
* * * UPDATE AT 1844 EST ON 12/17/01 FROM DAVID GUDGER TO S. SANDIN * * *
"This notification is a follow-up to the notification provided on 12/16/01."
"The following is a description of the sequence of plant events that occurred as determined by the failure analysis."
"Initial conditions: Feedwater level control was on Master Level controller selected to the 'B' narrow range channel and the plant was operating at 100% power. Due to the level summer card 1C34K657 failure, the level signal from the selected channel rapidly decreased to less than 178 inches, Level 3 (L3) signal, which caused the reactor recirculation pumps to receive a fast-to-slow speed downshift. Simultaneously, feedwater flow rapidly increased in response to the low level on the 'B' channel. The feedwater pumps tripped at 219 inches, Level 8 (L8), which occurred due to swell from the recirculation pumps downshift and increased feedwater flow. The [motor feed pump (MFP)] did not automatically start due to the L8 signal (as designed) and level decreased to 130 inches, Level 2 (L2). The recirculation pumps tripped, and [the] HPCS and RCIC systems [automatically] started and restored level. The MFP L8 signals were reset, and level was controlled on the startup controller."
"The Reactor Water Cleanup Inboard Containment Isolation Valve failed to close as designed upon [the] L2 isolation signal. The valve was manually closed during containment valve isolation verification following the scram. The cause of the failure to close was determined to be relay failure. Two relays were subsequently replaced and retested satisfactorily."
The licensee informed the NRC resident inspector. The NRC operations officer notified the R3DO (Phillips).
At 2228 EST on 12/15/01, the reactor recirculation pumps down shifted to slow speed for unknown reasons. This caused reactor pressure vessel (RPV) water level to swell, and an automatic reactor scram from 100% power occurred when Level 8 was reached. All control rods fully inserted. Following the scram, RPV water level started to decrease because the feedwater pumps also tripped (as designed) when Level 8 was reached. When RPV water level dropped to Level 2 (130 inches), balance-of-plant isolations occurred, and all of the applicable valves properly isolated. In addition, the high pressure core spray (HPCS) and reactor core isolation cooling (RCIC) systems automatically actuated when Level 2 was reached. The HPCS and RCIC systems were utilized to restore RPV water level to the normal range.
The unit is currently stable in Mode 3 with RPV pressure at 900 psi and RPV water level at 210 inches. Normal feedwater is being utilized to maintain RPV water level within the proper band, and the HPCS and RCIC systems have been secured. The main steam isolation valves remained open, and the condenser is being utilized as the heat sink. Containment parameters are currently normal, the electrical grid is stable, and the emergency diesel generators are available.
The licensee stated that all systems functioned as required and that there was nothing unusual or misunderstood other than the cause of the initiating event (the recirculation pumps down shifting to slow speed). The licensee's investigation is underway.
The licensee notified the NRC resident inspector.
* * * UPDATE AT 1844 EST ON 12/17/01 FROM DAVID GUDGER TO S. SANDIN * * *
"This notification is a follow-up to the notification provided on 12/16/01."
"The following is a description of the sequence of plant events that occurred as determined by the failure analysis."
"Initial conditions: Feedwater level control was on Master Level controller selected to the 'B' narrow range channel and the plant was operating at 100% power. Due to the level summer card 1C34K657 failure, the level signal from the selected channel rapidly decreased to less than 178 inches, Level 3 (L3) signal, which caused the reactor recirculation pumps to receive a fast-to-slow speed downshift. Simultaneously, feedwater flow rapidly increased in response to the low level on the 'B' channel. The feedwater pumps tripped at 219 inches, Level 8 (L8), which occurred due to swell from the recirculation pumps downshift and increased feedwater flow. The [motor feed pump (MFP)] did not automatically start due to the L8 signal (as designed) and level decreased to 130 inches, Level 2 (L2). The recirculation pumps tripped, and [the] HPCS and RCIC systems [automatically] started and restored level. The MFP L8 signals were reset, and level was controlled on the startup controller."
"The Reactor Water Cleanup Inboard Containment Isolation Valve failed to close as designed upon [the] L2 isolation signal. The valve was manually closed during containment valve isolation verification following the scram. The cause of the failure to close was determined to be relay failure. Two relays were subsequently replaced and retested satisfactorily."
The licensee informed the NRC resident inspector. The NRC operations officer notified the R3DO (Phillips).