Event Notification Report for September 04, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
09/03/2001 - 09/04/2001
Power Reactor
Event Number: 38263
Facility: WATTS BAR
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: RODEN
HQ OPS Officer: CHAUNCEY GOULD
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: RODEN
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 09/04/2001
Notification Time: 16:54 [ET]
Event Date: 09/04/2001
Event Time: 13:04 [EDT]
Last Update Date: 09/20/2001
Notification Time: 16:54 [ET]
Event Date: 09/04/2001
Event Time: 13:04 [EDT]
Last Update Date: 09/20/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
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)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
JOHN MONNINGER (R2)
JOHN MONNINGER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 100 | Power Operation | 0 | Hot Standby |
REACTOR MANUALLY TRIPPED FROM 100% POWER DUE TO STEAM GENERATOR LEVEL APPROACHING THE LO-LO LEVEL SETPOINT
While operating at 100% power, the Watts Bar Unit 1 reactor was manually tripped at approximately 1304 EDT on September 4, 2001 due to steam generator (SG) level approaching the Lo-Lo level setpoint. Just prior to the manual trip of the reactor, an abnormal voltage alarm was received on 125V Vital Battery Board 1-1. The immediate cause of the trip is considered to be the loss of the 1-1 Vital AC Inverter which caused a feedwater regulating valve to close. Subsequently, the maintenance power supply was used to restore power to the 1-1 vital AC bus.
All control rods inserted properly in response to the reactor trip. The Auxiliary Feedwater (AFW) System actuated as expected. However, subsequent to the initial actuation signal, additional operator action was required to[manually] start 1B motor driven pump. After the 1B pump was successfully started, the motor driven AFW pump 1A was secured based on instrument indications available to the operating staff. After power was restored to the vital AC bus, the 1A pump was subsequently started and performed properly. Inspections are currently underway to determine the exact cause of the loss of the 1-1 Vital Inverter and basis for the 1B pump failing to [auto] start.
No ECCS injection occurred, but the steam generator atmospheric relief valves lifted and fully reseated.
The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv).
The plant is stable in mode 3.
The NRC Resident Inspector was notified.
* * * UPDATE 1244EDT ON 9/20/01 FROM CHARLES TOUCHSTONE TO S. SANDIN * * *
The following information was received as an update:
"This report supplements the report made by TVA at approximately 1654 on September 4, 2001 - NRC Event Report 38263.
"Description of Event:
"While operating at 100% power, the Watts Bar Unit 1 reactor was manually tripped at approximately 1305 EDT on September 4, 2001, due to the level of Steam Generator (SG) Number 1 approaching the low-low level setpoint. The immediate cause of the trip is considered to be the loss of the 1-I Vital AC Inverter which caused the Number 1 SG feedwater regulating valve to close. Subsequently, the maintenance power supply was used to restore power to the 1-I Vital AC board.
"10 CFR 50.72 Notification:
"Initial notification to NRC of the reactor trip was made by WBN Operations at 1654 EDT The manual actuation of the Reactor Protection System (RPS) was reported as a four hour report under 10 CFR 50.72 (b)(2)(iv). The actuation of the Auxiliary Feedwater System (an engineered safety feature) was reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv). This report supplements the initial report and is being reported under the same criteria.
"Cause of the Event:
"The root cause of the reactor trip was the failure of the sensing/current limiting card installed in vital battery charger 6-S or its associated edge connector. The failure of the component resulted in the DC output of the charger rising to 209V. The high DC output voltage resulted in a high input to inverters 1-I and 2-I. Inverter 1-I input and output fuses were blown, while inverter 2-I continued without failure. This occurred because inverter 1-I is equipped with parallel 5 and 15 KVA constant voltage transformers (CVTs), which are more sensitive to high input voltage than the type used in inverter 2-1 which utilizes a single 20 KVA CVT. The inverters are rated for a maximum input of 140 volts.
"Equipment Response:
"All control rods inserted properly in response to the reactor trip. For WBN, the AFW system consist of a turbine driven AFW (TDAFW) pump and two trains (Train A and Train B) of motor driven AFW (MDAFW) pumps. The design calls for the MDAFW arid TDAFW pumps to start due to the loss of both Main Feedwater (MFW) pumps or an AMSAC actuation, In addition, the MDAFW pumps will actuate on low-low S/G level in any S/G and the TDAFW pump will actuate on low-low level in any two S/Gs. For this event, actuation of the TDAFW pump resulted from the required logic being met for SGs Number 1 and 4. Based on data obtained subsequent to the reactor trip, an auto start for the TDAFW train was initiated immediately after S/G low-low level setpoints were met.
"Subsequent to the initial AFW actuation signal, additional operator action was required to start the 1A-A and 1B-B motor driven pumps. For the Train A MDAFW, the pump did not receive an auto start signal due to the loss of Vital Board 1-I. This resulted from the power for the signal generating circuit being lost with the Vital Board. The evaluation performed to establish the cause for the Train B MDAFW not starting found that the breaker for the pump did not close as a result of the automatic signal initiated by the low SG level. The breaker was subsequently tested and performed as designed and therefore, no conclusive cause was found for the breaker not closing. However, it was established that the most likely cause of the problem was dirty relay contacts on the 1X and 30RX relays."
The licensee informed the NRC resident inspector. Notified R2DO(Sparks).
While operating at 100% power, the Watts Bar Unit 1 reactor was manually tripped at approximately 1304 EDT on September 4, 2001 due to steam generator (SG) level approaching the Lo-Lo level setpoint. Just prior to the manual trip of the reactor, an abnormal voltage alarm was received on 125V Vital Battery Board 1-1. The immediate cause of the trip is considered to be the loss of the 1-1 Vital AC Inverter which caused a feedwater regulating valve to close. Subsequently, the maintenance power supply was used to restore power to the 1-1 vital AC bus.
All control rods inserted properly in response to the reactor trip. The Auxiliary Feedwater (AFW) System actuated as expected. However, subsequent to the initial actuation signal, additional operator action was required to[manually] start 1B motor driven pump. After the 1B pump was successfully started, the motor driven AFW pump 1A was secured based on instrument indications available to the operating staff. After power was restored to the vital AC bus, the 1A pump was subsequently started and performed properly. Inspections are currently underway to determine the exact cause of the loss of the 1-1 Vital Inverter and basis for the 1B pump failing to [auto] start.
No ECCS injection occurred, but the steam generator atmospheric relief valves lifted and fully reseated.
The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv).
The plant is stable in mode 3.
The NRC Resident Inspector was notified.
* * * UPDATE 1244EDT ON 9/20/01 FROM CHARLES TOUCHSTONE TO S. SANDIN * * *
The following information was received as an update:
"This report supplements the report made by TVA at approximately 1654 on September 4, 2001 - NRC Event Report 38263.
"Description of Event:
"While operating at 100% power, the Watts Bar Unit 1 reactor was manually tripped at approximately 1305 EDT on September 4, 2001, due to the level of Steam Generator (SG) Number 1 approaching the low-low level setpoint. The immediate cause of the trip is considered to be the loss of the 1-I Vital AC Inverter which caused the Number 1 SG feedwater regulating valve to close. Subsequently, the maintenance power supply was used to restore power to the 1-I Vital AC board.
"10 CFR 50.72 Notification:
"Initial notification to NRC of the reactor trip was made by WBN Operations at 1654 EDT The manual actuation of the Reactor Protection System (RPS) was reported as a four hour report under 10 CFR 50.72 (b)(2)(iv). The actuation of the Auxiliary Feedwater System (an engineered safety feature) was reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv). This report supplements the initial report and is being reported under the same criteria.
"Cause of the Event:
"The root cause of the reactor trip was the failure of the sensing/current limiting card installed in vital battery charger 6-S or its associated edge connector. The failure of the component resulted in the DC output of the charger rising to 209V. The high DC output voltage resulted in a high input to inverters 1-I and 2-I. Inverter 1-I input and output fuses were blown, while inverter 2-I continued without failure. This occurred because inverter 1-I is equipped with parallel 5 and 15 KVA constant voltage transformers (CVTs), which are more sensitive to high input voltage than the type used in inverter 2-1 which utilizes a single 20 KVA CVT. The inverters are rated for a maximum input of 140 volts.
"Equipment Response:
"All control rods inserted properly in response to the reactor trip. For WBN, the AFW system consist of a turbine driven AFW (TDAFW) pump and two trains (Train A and Train B) of motor driven AFW (MDAFW) pumps. The design calls for the MDAFW arid TDAFW pumps to start due to the loss of both Main Feedwater (MFW) pumps or an AMSAC actuation, In addition, the MDAFW pumps will actuate on low-low S/G level in any S/G and the TDAFW pump will actuate on low-low level in any two S/Gs. For this event, actuation of the TDAFW pump resulted from the required logic being met for SGs Number 1 and 4. Based on data obtained subsequent to the reactor trip, an auto start for the TDAFW train was initiated immediately after S/G low-low level setpoints were met.
"Subsequent to the initial AFW actuation signal, additional operator action was required to start the 1A-A and 1B-B motor driven pumps. For the Train A MDAFW, the pump did not receive an auto start signal due to the loss of Vital Board 1-I. This resulted from the power for the signal generating circuit being lost with the Vital Board. The evaluation performed to establish the cause for the Train B MDAFW not starting found that the breaker for the pump did not close as a result of the automatic signal initiated by the low SG level. The breaker was subsequently tested and performed as designed and therefore, no conclusive cause was found for the breaker not closing. However, it was established that the most likely cause of the problem was dirty relay contacts on the 1X and 30RX relays."
The licensee informed the NRC resident inspector. Notified R2DO(Sparks).
Power Reactor
Event Number: 38264
Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: LAVATE
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: MD
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: LAVATE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 09/04/2001
Notification Time: 17:00 [ET]
Event Date: 09/04/2001
Event Time: 15:00 [EDT]
Last Update Date: 09/04/2001
Notification Time: 17:00 [ET]
Event Date: 09/04/2001
Event Time: 15:00 [EDT]
Last Update Date: 09/04/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
HAROLD GRAY (R1)
HAROLD GRAY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
FAILURE OF 17 OUT OF THE 49 EMERGENCY NOTIFICATION SYSTEM SIRENS FOR CALVERT COUNTY
The licensee reported that 17 out of 49 sirens in Calvert County failed to respond to the scheduled county test.. However, all sirens did respond to a mobile siren test kit which is used locally at the siren. Calvert County Emergency Management was notified to perform route alerting in the affected areas in the event of an emergency. The cause is believed to be a transmission problem.
The NRC Resident Inspector will be notified. The county has been notified.
The licensee reported that 17 out of 49 sirens in Calvert County failed to respond to the scheduled county test.. However, all sirens did respond to a mobile siren test kit which is used locally at the siren. Calvert County Emergency Management was notified to perform route alerting in the affected areas in the event of an emergency. The cause is believed to be a transmission problem.
The NRC Resident Inspector will be notified. The county has been notified.
Fuel Cycle Facility
Event Number: 38265
Facility: PADUCAH GASEOUS DIFFUSION PLANT
Region: 3 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: UNDERWOOD
HQ OPS Officer: CHAUNCEY GOULD
Region: 3 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: UNDERWOOD
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 09/04/2001
Notification Time: 18:09 [ET]
Event Date: 09/04/2001
Event Time: 14:00 [CDT]
Last Update Date: 09/12/2001
Notification Time: 18:09 [ET]
Event Date: 09/04/2001
Event Time: 14:00 [CDT]
Last Update Date: 09/12/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
MARK RING (R3)
M. WAYNE HODGES (NMSS)
MARK RING (R3)
M. WAYNE HODGES (NMSS)
24-HOUR 91-01 BULLETIN RESPONSE
At 1400, on 9/4/01, the Plant Shift Superintendent (PSS) was notified of an NCSA violation that had occurred at the C-355 air plant. Immediately after switching drying units at the air plant, a high-high moisture alarm was received. The alarm indicated air moisture content greater than 1300 ppm water, violating NCSA GEN-10-01. NCSA GEN-010 credits the dry air system for producing dry air with moisture content of less than 1300 ppm. The purpose of this requirement is to prevent exposure of fissile uranium deposits to a moderating environment. No equipment containing fissile material was exposed to high moisture content plant air at the time of the alarm or following receipt of the alarm.
Since one leg of double contingency was lost, this is being reported to the NRC as a 24-hr. event report.
SAFETY SIGNIFICANCE OF EVENTS:
There was no purging of equipment containing fissile material in progress at the time of the moisture excursion. Although a parameter was exceeded, the proper response to the alarm was taken prior to resuming buffering of purging equipment containing fissile material.
POTENTIAL CRITICALITY PATHWAYS INVOLVED(BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR:
In order for a criticality to be possible. Operations personnel would have had to fail to respond to the alarm. In addition, a fissile deposit containing greater than a critical mass and absorbing greater than 10 kg of water would have had to be present.
CONTROLLED PARAMETERS (MASS, MODERATION. GEOMETRY, CONCENTRATION. ETC:
Double contingency for this scenario is established by implementing two controls on moderation.
ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS):
n/a
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
Double contingency for this scenario is established by implementing two controls on moderation.
The first leg of double contingency Is based on receiving a high-high moisture alarm indicating a moisture content greater than 1300 ppm water and ceasing buffering and purging operations in response to the alarm. This control was not violated.
The second leg of double contingency is based on the dry air system producing plant air with a moisture content less than 1300 ppm water. Since the moisture content was confirmed to be greater than 1300 ppm, this process condition was exceeded and double contingency was not maintained.
Since double contingency is based on two controls on moderation, double contingency was not maintained.
CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:
No corrective actions implemented. Proper procedure response was performed.
The NRC Resident Inspector was notified.
* * * UPDATE ON 09/12/01 AT 1455 ET BY PITTMAN TAKEN BY MACKINNON * * *
Criticality Safety Incident Report NCS-INC-01-022 has been revised to include information from EN-C-826-01-039 that documented that the moisture content did not exceed the NCS limit of 1300 ppm. Since the moisture limit for the plant air system was not exceeded, no violation occurred. R3DO (Sonia Burgess) notified.
The NRC Resident Inspector was notified of this update by the certificate holder.
At 1400, on 9/4/01, the Plant Shift Superintendent (PSS) was notified of an NCSA violation that had occurred at the C-355 air plant. Immediately after switching drying units at the air plant, a high-high moisture alarm was received. The alarm indicated air moisture content greater than 1300 ppm water, violating NCSA GEN-10-01. NCSA GEN-010 credits the dry air system for producing dry air with moisture content of less than 1300 ppm. The purpose of this requirement is to prevent exposure of fissile uranium deposits to a moderating environment. No equipment containing fissile material was exposed to high moisture content plant air at the time of the alarm or following receipt of the alarm.
Since one leg of double contingency was lost, this is being reported to the NRC as a 24-hr. event report.
SAFETY SIGNIFICANCE OF EVENTS:
There was no purging of equipment containing fissile material in progress at the time of the moisture excursion. Although a parameter was exceeded, the proper response to the alarm was taken prior to resuming buffering of purging equipment containing fissile material.
POTENTIAL CRITICALITY PATHWAYS INVOLVED(BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR:
In order for a criticality to be possible. Operations personnel would have had to fail to respond to the alarm. In addition, a fissile deposit containing greater than a critical mass and absorbing greater than 10 kg of water would have had to be present.
CONTROLLED PARAMETERS (MASS, MODERATION. GEOMETRY, CONCENTRATION. ETC:
Double contingency for this scenario is established by implementing two controls on moderation.
ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS):
n/a
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
Double contingency for this scenario is established by implementing two controls on moderation.
The first leg of double contingency Is based on receiving a high-high moisture alarm indicating a moisture content greater than 1300 ppm water and ceasing buffering and purging operations in response to the alarm. This control was not violated.
The second leg of double contingency is based on the dry air system producing plant air with a moisture content less than 1300 ppm water. Since the moisture content was confirmed to be greater than 1300 ppm, this process condition was exceeded and double contingency was not maintained.
Since double contingency is based on two controls on moderation, double contingency was not maintained.
CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:
No corrective actions implemented. Proper procedure response was performed.
The NRC Resident Inspector was notified.
* * * UPDATE ON 09/12/01 AT 1455 ET BY PITTMAN TAKEN BY MACKINNON * * *
Criticality Safety Incident Report NCS-INC-01-022 has been revised to include information from EN-C-826-01-039 that documented that the moisture content did not exceed the NCS limit of 1300 ppm. Since the moisture limit for the plant air system was not exceeded, no violation occurred. R3DO (Sonia Burgess) notified.
The NRC Resident Inspector was notified of this update by the certificate holder.