Event Notification Report for April 03, 2015
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
04/02/2015 - 04/03/2015
EVENT NUMBERS
5106850950509515095250953
Power Reactor
Event Number: 51068
Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: DON GREGOIRE
HQ OPS Officer: DANIEL MILLS
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: DON GREGOIRE
HQ OPS Officer: DANIEL MILLS
Notification Date: 05/12/2015
Notification Time: 22:13 [ET]
Event Date: 04/03/2015
Event Time: 00:00 [PDT]
Last Update Date: 05/12/2015
Notification Time: 22:13 [ET]
Event Date: 04/03/2015
Event Time: 00:00 [PDT]
Last Update Date: 05/12/2015
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):
MICHAEL VASQUEZ (R4DO)
MICHAEL VASQUEZ (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
LOSS OF RADIOLOGICAL ASSESSMENT CAPABILITY DUE TO NON-FUNCTIONAL RADIATION MONITOR
"On 4/21/2015, during performance of source check surveillance on the liquid effluent radiation monitor for the Plant Service Water (TSW), a non-radioactive system, it was discovered that the instrument was determined to be nonfunctional. It was determined on 4/25/15 that the failure was due to an incorrect 'as left' setting from testing conducted on 4/3/2015. The instrument was determined to be non-functional from the period 4/03/15 to 4/25/15 when the setting was corrected.
"On 5/12/15 it was recognized that because no compensatory measures were implemented during the time the instrument was non-functional that this condition constituted a major loss of radiation assessment capability which is reportable in accordance with 10 CFR 50.72(b)(3)(xiii).
"The NRC Resident Inspector will be notified."
"On 4/21/2015, during performance of source check surveillance on the liquid effluent radiation monitor for the Plant Service Water (TSW), a non-radioactive system, it was discovered that the instrument was determined to be nonfunctional. It was determined on 4/25/15 that the failure was due to an incorrect 'as left' setting from testing conducted on 4/3/2015. The instrument was determined to be non-functional from the period 4/03/15 to 4/25/15 when the setting was corrected.
"On 5/12/15 it was recognized that because no compensatory measures were implemented during the time the instrument was non-functional that this condition constituted a major loss of radiation assessment capability which is reportable in accordance with 10 CFR 50.72(b)(3)(xiii).
"The NRC Resident Inspector will be notified."
Power Reactor
Event Number: 50950
Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFF HUMAN
HQ OPS Officer: DANIEL MILLS
Region: 3 State: MN
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFF HUMAN
HQ OPS Officer: DANIEL MILLS
Notification Date: 04/03/2015
Notification Time: 10:12 [ET]
Event Date: 04/03/2015
Event Time: 06:52 [CDT]
Last Update Date: 04/03/2015
Notification Time: 10:12 [ET]
Event Date: 04/03/2015
Event Time: 06:52 [CDT]
Last Update Date: 04/03/2015
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):
ERIC DUNCAN (R3DO)
ERIC DUNCAN (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 100 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR SCRAM DUE TO LOSS OF MAIN FEEDWATER PUMP
"On April 3, 2015 at 0652 CDT, the Unit 2 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 21 Main Feedwater Pump as required by the annunciator response procedure for the lockout alarm. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. The auxiliary feedwater pumps have subsequently been secured and returned to automatic. Steam generators are being supplied by 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system.
"The cause of 21 Main Feedwater Pump trip has been determined to be a failed suction pressure switch.
"There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event.
"The NRC Senior Resident Inspector has been notified."
The licensee plans to issue a press release.
"On April 3, 2015 at 0652 CDT, the Unit 2 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 21 Main Feedwater Pump as required by the annunciator response procedure for the lockout alarm. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. The auxiliary feedwater pumps have subsequently been secured and returned to automatic. Steam generators are being supplied by 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system.
"The cause of 21 Main Feedwater Pump trip has been determined to be a failed suction pressure switch.
"There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event.
"The NRC Senior Resident Inspector has been notified."
The licensee plans to issue a press release.
Power Reactor
Event Number: 50951
Facility: LIMERICK
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JOHN HARKINS
HQ OPS Officer: STEVE SANDIN
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JOHN HARKINS
HQ OPS Officer: STEVE SANDIN
Notification Date: 04/03/2015
Notification Time: 11:54 [ET]
Event Date: 04/03/2015
Event Time: 10:00 [EDT]
Last Update Date: 04/03/2015
Notification Time: 11:54 [ET]
Event Date: 04/03/2015
Event Time: 10:00 [EDT]
Last Update Date: 04/03/2015
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):
SILAS KENNEDY (R1DO)
SILAS KENNEDY (R1DO)
| 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 | 82 | Power Operation | 82 | Power Operation |
INOPERABLE SEISMIC INSTRUMENTATION IDENTIFIED DURING HISTORICAL REVIEW
"Limerick Generating Station (LGS) has completed a review of recent seismic monitor performance. The seismic monitor is currently capable of fulfilling its emergency assessment support function. The review of the system performance over the last three years identified two occasions when the system was degraded and incapable of performing its emergency assessment support function such that emergency classification at the UNUSUAL EVENT and ALERT levels could not be declared within fifteen minutes as required with site instrumentation. The seismic monitor was determined to be degraded and incapable of performing its emergency assessment support function on the following dates:
1) January 17, 2014 - Unit 1 100% power, Unit 2 97% power
2) March 22, 2014 - Unit 1 Refueling, Unit 2 100% power
"These unplanned inoperable conditions of the seismic monitor were entered into the LGS Corrective Action Program (CAP) when they occurred.
"While Exelon procedural direction allowed the use of offsite sources to obtain seismic data when the seismic monitor is incapable of assessing emergency plan Emergency Action Levels (EALs), this was not explicitly referenced in the approved EALs. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery."
For the two dates above, the 1/17/14 failure was related to function of the system CPU (central processing unit) and the 3/22/14 failure was due to a failed instrument backup battery. There was no seismic event on either of these two dates.
The licensee informed the NRC Resident Inspector.
"Limerick Generating Station (LGS) has completed a review of recent seismic monitor performance. The seismic monitor is currently capable of fulfilling its emergency assessment support function. The review of the system performance over the last three years identified two occasions when the system was degraded and incapable of performing its emergency assessment support function such that emergency classification at the UNUSUAL EVENT and ALERT levels could not be declared within fifteen minutes as required with site instrumentation. The seismic monitor was determined to be degraded and incapable of performing its emergency assessment support function on the following dates:
1) January 17, 2014 - Unit 1 100% power, Unit 2 97% power
2) March 22, 2014 - Unit 1 Refueling, Unit 2 100% power
"These unplanned inoperable conditions of the seismic monitor were entered into the LGS Corrective Action Program (CAP) when they occurred.
"While Exelon procedural direction allowed the use of offsite sources to obtain seismic data when the seismic monitor is incapable of assessing emergency plan Emergency Action Levels (EALs), this was not explicitly referenced in the approved EALs. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery."
For the two dates above, the 1/17/14 failure was related to function of the system CPU (central processing unit) and the 3/22/14 failure was due to a failed instrument backup battery. There was no seismic event on either of these two dates.
The licensee informed the NRC Resident Inspector.
Power Reactor
Event Number: 50952
Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: JACK BREEN
HQ OPS Officer: JEFF ROTTON
Region: 2 State: FL
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: JACK BREEN
HQ OPS Officer: JEFF ROTTON
Notification Date: 04/03/2015
Notification Time: 14:17 [ET]
Event Date: 04/03/2015
Event Time: 15:00 [EDT]
Last Update Date: 04/03/2015
Notification Time: 14:17 [ET]
Event Date: 04/03/2015
Event Time: 15:00 [EDT]
Last Update Date: 04/03/2015
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):
SHAKUR WALKER (R2DO)
SHAKUR WALKER (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Defueled | 0 | Defueled |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
TECHNICAL SUPPORT CENTER VENTILATION NONFUNCTIONAL DURING PREPLANNED MAINTENANCE
"On April 3, 2015, at 1500 EDT, the Control Room Emergency Ventilation System on St. Lucie Unit 1 will be declared inoperable due to pre-planned maintenance during the current refueling outage. The Technical Support Center (TSC) ventilation system is part of the Unit 1 Control Room Emergency Ventilation System, therefore, the TSC ventilation system has been rendered non-functional during the course of the work activities. The TSC ventilation is expected to be returned to service in approximately 24 hours.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. Should the TSC become uninhabitable, the TSC staff will relocate to an alternate TSC location in accordance with applicable site procedures.
"This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the potential loss of an emergency response facility. An update will be provided once the TSC ventilation system has been restored to normal operation.
"The NRC Resident Inspector has been notified."
* * * UPDATE PROVIDED BY DAVE FIELDS TO JEFF ROTTON AT 2240 EDT ON 4/03/2015 * * *
"The TSC ventilation system has been restored to normal operation as of 2230 EDT on April 3, 2015. The NRC Resident Inspector has been notified."
Notified R2DO (Walker).
"On April 3, 2015, at 1500 EDT, the Control Room Emergency Ventilation System on St. Lucie Unit 1 will be declared inoperable due to pre-planned maintenance during the current refueling outage. The Technical Support Center (TSC) ventilation system is part of the Unit 1 Control Room Emergency Ventilation System, therefore, the TSC ventilation system has been rendered non-functional during the course of the work activities. The TSC ventilation is expected to be returned to service in approximately 24 hours.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. Should the TSC become uninhabitable, the TSC staff will relocate to an alternate TSC location in accordance with applicable site procedures.
"This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the potential loss of an emergency response facility. An update will be provided once the TSC ventilation system has been restored to normal operation.
"The NRC Resident Inspector has been notified."
* * * UPDATE PROVIDED BY DAVE FIELDS TO JEFF ROTTON AT 2240 EDT ON 4/03/2015 * * *
"The TSC ventilation system has been restored to normal operation as of 2230 EDT on April 3, 2015. The NRC Resident Inspector has been notified."
Notified R2DO (Walker).
Power Reactor
Event Number: 50953
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MURTAZA ABBAS
HQ OPS Officer: STEVE SANDIN
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MURTAZA ABBAS
HQ OPS Officer: STEVE SANDIN
Notification Date: 04/03/2015
Notification Time: 19:37 [ET]
Event Date: 04/03/2015
Event Time: 16:31 [CDT]
Last Update Date: 04/03/2015
Notification Time: 19:37 [ET]
Event Date: 04/03/2015
Event Time: 16:31 [CDT]
Last Update Date: 04/03/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
ERIC DUNCAN (R3DO)
ERIC DUNCAN (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
LIQUID PENETRATION EXAMINATION RESULTS IN INDICATION ON REACTOR VESSEL HEAD PENETRATION
"On April 3, 2015, during the Braidwood Station Unit 1 refueling outage (A1R18), an in-service Liquid Penetration examination was performed on the previously repaired control rod drive mechanism (CRDM) penetration 69. During the examination on the weld build up for CRDM penetration 69, a 3/8 inch rounded indication was discovered located at 0 degrees on the reactor head portion of the weld build up, and it is 4 inches from the transition of the head to penetration. 0 degrees is located at the outermost portion of the penetration on the flange side. The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head.
"Rounded indications that exceed 3/16 inch are rejectable per ASME Code Case N-729-1.
"This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded' since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair.
"The NRC Resident Inspector has been informed."
"On April 3, 2015, during the Braidwood Station Unit 1 refueling outage (A1R18), an in-service Liquid Penetration examination was performed on the previously repaired control rod drive mechanism (CRDM) penetration 69. During the examination on the weld build up for CRDM penetration 69, a 3/8 inch rounded indication was discovered located at 0 degrees on the reactor head portion of the weld build up, and it is 4 inches from the transition of the head to penetration. 0 degrees is located at the outermost portion of the penetration on the flange side. The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head.
"Rounded indications that exceed 3/16 inch are rejectable per ASME Code Case N-729-1.
"This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded' since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair.
"The NRC Resident Inspector has been informed."