Event Notification Report for October 05, 2015
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/04/2015 - 10/05/2015
EVENT NUMBERS
514585145351454514555145651450
Part 21
Event Number: 51458
Rep Org: DRESDEN
Licensee: GENERAL ELECTRIC HITACHI
Region: 3
City: MORRIS State: IL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: AARON THOMPSON
HQ OPS Officer: JEFF HERRERA
Licensee: GENERAL ELECTRIC HITACHI
Region: 3
City: MORRIS State: IL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: AARON THOMPSON
HQ OPS Officer: JEFF HERRERA
Notification Date: 10/06/2015
Notification Time: 22:06 [ET]
Event Date: 10/05/2015
Event Time: 12:00 [CDT]
Last Update Date: 10/06/2015
Notification Time: 22:06 [ET]
Event Date: 10/05/2015
Event Time: 12:00 [CDT]
Last Update Date: 10/06/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
ANN MARIE STONE (R3DO)
PART 21/50.55 REACT (EMAI)
ANN MARIE STONE (R3DO)
PART 21/50.55 REACT (EMAI)
PART 21 REPORT - ELECTROMATIC RELIEF VALVE (EMRV) CUTOUT SWITCH
"Following the return of the actuator that failed bench testing to GEH, on 6/12/15 at 1804 [CDT], General Electric Hitachi (GEH) notified Dresden Station of a potential parts quality Potential Failure of the EMRV Cutout Switch. It has been determined the notification is applicable to DNPS [Dresden Nuclear Power Station], Units 2 and 3.
"The GEH investigation concluded that the EMRV actuator assemblies failed to change state because of the failure of the cutout switch to fully close and provide the appropriate current path. Multiple contributing factors were discovered which could have led to the presence of the gaps in the cutout switch. The most significant of these factors is a change in lever arm positioning causing increased forces in the tension spring which prevent proper closure of the cutout switch. Design changes to reduce wear caused by vibration on the actuators changed lever arm position and also allowed for additional dimensional tolerance which tended to increase force in the tension spring.
"Identification of Facility and Component: DNPS / EMRV Actuator, GEH Part Number 352B2632G001
"Safety Significance (e.g., substantial hazard that is or could be created): Identified condition is a Potential Substantial Safety Hazard since it could cause affected EMRVs to fail to operate as designed, which could result in a loss of safety function. Potential to affect the Minimum Critical Power Ratio (MCPR), Reactor Coolant System (RCS), Automatic Depressurization System (ADS), and Low Set Relief Function
"Plants with similar GEH cutout switches: Quad Cities Nuclear Power Station."
Part 21 Reference: EN #51386
The licensee notified the NRC Resident Inspector.
"Following the return of the actuator that failed bench testing to GEH, on 6/12/15 at 1804 [CDT], General Electric Hitachi (GEH) notified Dresden Station of a potential parts quality Potential Failure of the EMRV Cutout Switch. It has been determined the notification is applicable to DNPS [Dresden Nuclear Power Station], Units 2 and 3.
"The GEH investigation concluded that the EMRV actuator assemblies failed to change state because of the failure of the cutout switch to fully close and provide the appropriate current path. Multiple contributing factors were discovered which could have led to the presence of the gaps in the cutout switch. The most significant of these factors is a change in lever arm positioning causing increased forces in the tension spring which prevent proper closure of the cutout switch. Design changes to reduce wear caused by vibration on the actuators changed lever arm position and also allowed for additional dimensional tolerance which tended to increase force in the tension spring.
"Identification of Facility and Component: DNPS / EMRV Actuator, GEH Part Number 352B2632G001
"Safety Significance (e.g., substantial hazard that is or could be created): Identified condition is a Potential Substantial Safety Hazard since it could cause affected EMRVs to fail to operate as designed, which could result in a loss of safety function. Potential to affect the Minimum Critical Power Ratio (MCPR), Reactor Coolant System (RCS), Automatic Depressurization System (ADS), and Low Set Relief Function
"Plants with similar GEH cutout switches: Quad Cities Nuclear Power Station."
Part 21 Reference: EN #51386
The licensee notified the NRC Resident Inspector.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 51453
Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: JUSTIN E. CROCKER
HQ OPS Officer: JEFF HERRERA
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: JUSTIN E. CROCKER
HQ OPS Officer: JEFF HERRERA
Notification Date: 10/05/2015
Notification Time: 15:01 [ET]
Event Date: 10/05/2015
Event Time: 08:15 [EDT]
Last Update Date: 10/22/2015
Notification Time: 15:01 [ET]
Event Date: 10/05/2015
Event Time: 08:15 [EDT]
Last Update Date: 10/22/2015
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):
BRICE BICKETT (R1DO)
BRICE BICKETT (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Refueling | 0 | Refueling |
TWO REACTOR VESSEL HEAD PENETRATIONS COULD NOT BE DISPOSITIONED AS ACCEPTABLE
"On 10/5/2015, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing planned ultrasonic examinations (UT) on the 66 reactor vessel head penetrations, it was determined, that two penetrations could not be dispositioned as acceptable per ASME [American Society of Mechanical Engineers] Code Section XI in a Reactor Coolant System pressure boundary. The indications of a degraded condition, on these two penetrations, are not through wall, as no leak path was identified. The examinations are being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D), and ASME Code Case N-729-1, to find potential flaws/indications well before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. All 66 reactor vessel head penetrations are scheduled to be examined during the current refueling outage.
"The plant is currently shutdown and in Mode 6. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup.
"This is reportable, pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indications 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 notified."
* * * RETRACTION FROM DAN SCHWER TO STEVEN VITTO ON 10/22/2015 AT 1039 EDT * * *
"The Beaver Valley Power Station Unit 2, FirstEnergy Nuclear Operating Company (FENOC) is retracting the 8-hour non-emergency notification made on 10/05/2015 (EN 51453). This retraction is based on additional examinations and subsequent engineering assessments completed that concluded the penetrations meet the requirements of ASME Code Case N-729-1 as amended by 10CFR50.55a(g)(6)(ii)(D) and no repairs are required.
"The two reactor head penetration indications, are therefore not reportable, pursuant to 10CFR50.72(b)(3)(ii)(A).
"The NRC Resident Inspector has been notified."
R1DO(Gray) has been notified.
"On 10/5/2015, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing planned ultrasonic examinations (UT) on the 66 reactor vessel head penetrations, it was determined, that two penetrations could not be dispositioned as acceptable per ASME [American Society of Mechanical Engineers] Code Section XI in a Reactor Coolant System pressure boundary. The indications of a degraded condition, on these two penetrations, are not through wall, as no leak path was identified. The examinations are being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D), and ASME Code Case N-729-1, to find potential flaws/indications well before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. All 66 reactor vessel head penetrations are scheduled to be examined during the current refueling outage.
"The plant is currently shutdown and in Mode 6. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup.
"This is reportable, pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indications 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 notified."
* * * RETRACTION FROM DAN SCHWER TO STEVEN VITTO ON 10/22/2015 AT 1039 EDT * * *
"The Beaver Valley Power Station Unit 2, FirstEnergy Nuclear Operating Company (FENOC) is retracting the 8-hour non-emergency notification made on 10/05/2015 (EN 51453). This retraction is based on additional examinations and subsequent engineering assessments completed that concluded the penetrations meet the requirements of ASME Code Case N-729-1 as amended by 10CFR50.55a(g)(6)(ii)(D) and no repairs are required.
"The two reactor head penetration indications, are therefore not reportable, pursuant to 10CFR50.72(b)(3)(ii)(A).
"The NRC Resident Inspector has been notified."
R1DO(Gray) has been notified.
Part 21
Event Number: 51454
Rep Org: AZZ/NUCLEAR LOGISTICS INC
Licensee: AZZ/NUCLEAR LOGISTICS, INC.
Region: 4
City: FORT WORTH State: TX
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TRACY BOLT
HQ OPS Officer: JEFF HERRERA
Licensee: AZZ/NUCLEAR LOGISTICS, INC.
Region: 4
City: FORT WORTH State: TX
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TRACY BOLT
HQ OPS Officer: JEFF HERRERA
Notification Date: 10/05/2015
Notification Time: 17:47 [ET]
Event Date: 10/05/2015
Event Time: 00:00 [CDT]
Last Update Date: 10/05/2015
Notification Time: 17:47 [ET]
Event Date: 10/05/2015
Event Time: 00:00 [CDT]
Last Update Date: 10/05/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
NICK TAYLOR (R4DO)
PART 21/50.55 REACT (EMAI)
NICK TAYLOR (R4DO)
PART 21/50.55 REACT (EMAI)
PART 21 - ELECTRONIC POSITION CONTROLLER FAILED INTERNAL COMPONENT
The following was excerpted from a fax received from the licensee:
"NLI manufactured the reverse engineered units to replace the existing obsolete electronic position controller NLI Model: NLI-10A-LM-P.
"The failed component within the controller was sized based on the original published ratings of the original components installed. The unit had an internal component that was subjected to a current value that was higher than the rating for the component. The actual applied current value was not originally provided to NLI when the units were being designed. NLI and the affected plant has been in communication regarding the component condition since the first reported failure.
"Three of units that initially failed in the plant have already been returned to NLI for failure analysis and design modifications. These units were modified by installing a larger bridge rectifier and wire wound resistor onto the revision 0 circuit board. The modified versions, quantity of three, S/N: NLI-10A-LM-P-1003- 0, NLI-10A-LM-P-1004-0 and NLI-10A-LM-P-1006-0 have been returned to Calloway under purchase orders 689233 SR and 694253 SR and have been accepted and installed into the plant and are working with no reported issues.
The model number NLI-10A-LM-P, has been redesigned with a larger bridge rectifier of a higher rating and wire wound resistor that will withstand the applied current in the intended application. The parent circuit board has also been revised to revision 1 to accept the larger bridge rectifier without impacting other components on the board.
The remaining unmodified revision 0 units that were originally supplied have not yet been returned to NLI for incorporation of the design modification. NLI recommends that at the plant's earliest convenience, the remaining units be returned to NLI to be retrofitted with the latest design configuration (revision 1 circuit board)."
The affected plant is Callaway.
The following was excerpted from a fax received from the licensee:
"NLI manufactured the reverse engineered units to replace the existing obsolete electronic position controller NLI Model: NLI-10A-LM-P.
"The failed component within the controller was sized based on the original published ratings of the original components installed. The unit had an internal component that was subjected to a current value that was higher than the rating for the component. The actual applied current value was not originally provided to NLI when the units were being designed. NLI and the affected plant has been in communication regarding the component condition since the first reported failure.
"Three of units that initially failed in the plant have already been returned to NLI for failure analysis and design modifications. These units were modified by installing a larger bridge rectifier and wire wound resistor onto the revision 0 circuit board. The modified versions, quantity of three, S/N: NLI-10A-LM-P-1003- 0, NLI-10A-LM-P-1004-0 and NLI-10A-LM-P-1006-0 have been returned to Calloway under purchase orders 689233 SR and 694253 SR and have been accepted and installed into the plant and are working with no reported issues.
The model number NLI-10A-LM-P, has been redesigned with a larger bridge rectifier of a higher rating and wire wound resistor that will withstand the applied current in the intended application. The parent circuit board has also been revised to revision 1 to accept the larger bridge rectifier without impacting other components on the board.
The remaining unmodified revision 0 units that were originally supplied have not yet been returned to NLI for incorporation of the design modification. NLI recommends that at the plant's earliest convenience, the remaining units be returned to NLI to be retrofitted with the latest design configuration (revision 1 circuit board)."
The affected plant is Callaway.
Power Reactor
Event Number: 51455
Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: ALBERTO SANCHEZ
HQ OPS Officer: JEFF HERRERA
Region: 1 State: PA
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: ALBERTO SANCHEZ
HQ OPS Officer: JEFF HERRERA
Notification Date: 10/05/2015
Notification Time: 21:47 [ET]
Event Date: 10/05/2015
Event Time: 21:15 [EDT]
Last Update Date: 10/06/2015
Notification Time: 21:47 [ET]
Event Date: 10/05/2015
Event Time: 21:15 [EDT]
Last Update Date: 10/06/2015
Emergency Class: ALERT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
Person (Organization):
BRICE BICKETT (R1DO)
DAN DORMAN (RA)
ALLEN HOWE (NRR)
BILL DEAN (NRR)
WILLIAM GOTT (IRD)
BRICE BICKETT (R1DO)
DAN DORMAN (RA)
ALLEN HOWE (NRR)
BILL DEAN (NRR)
WILLIAM GOTT (IRD)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
ALERT DECLARATION DUE TO FIRE IN THE AUXILIARY BUILDING
"At 2115 [EDT] on 10/05/15, an Alert was declared due to a fire in the Auxiliary building affecting DC-P-1A [Decay closed cooling pump 1A] and A-Train safety equipment. The fire is out."
The licensee reported the fire was extinguished at approximately 2201 EDT. The fire did not hamper operations personnel responding to the fire. Offsite fire assistance was requested.
The NRC Resident Inspector will be notified.
The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin.
Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
* * * UPDATE AT 0015 EDT ON 10/06/15 FROM JAMES CREIGHTON TO S. SANDIN * * *
The licensee is terminating the Alert at 0009 EDT on 10/06/15 based on the following:
"At 2115 [EDT] on 10/05/15, an Alert (HA3) was declared due to a fire in the Decay Closed Cooling Water Pump '1A' motor and breaker. The fire was extinguished at 2201 [EDT].
"Following inspection by electrical maintenance the 'P' 480V bus was re-energized at 2305 [EDT] and restoration of previously running loads is in progress.
"Station (TMI) is in a 72-hour LCO for repairs to the Decay Closed Cooling Water Pump '1A' (DC-P-1A).
"TMI (Three Mile Island) is terminating the event based upon the above information."
The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin. The licensee will issue a press release.
The licensee informed the NRC Resident Inspector. Notified R1DO (Bickett), NRR (Howe) and IRD (Gott).
Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
"At 2115 [EDT] on 10/05/15, an Alert was declared due to a fire in the Auxiliary building affecting DC-P-1A [Decay closed cooling pump 1A] and A-Train safety equipment. The fire is out."
The licensee reported the fire was extinguished at approximately 2201 EDT. The fire did not hamper operations personnel responding to the fire. Offsite fire assistance was requested.
The NRC Resident Inspector will be notified.
The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin.
Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
* * * UPDATE AT 0015 EDT ON 10/06/15 FROM JAMES CREIGHTON TO S. SANDIN * * *
The licensee is terminating the Alert at 0009 EDT on 10/06/15 based on the following:
"At 2115 [EDT] on 10/05/15, an Alert (HA3) was declared due to a fire in the Decay Closed Cooling Water Pump '1A' motor and breaker. The fire was extinguished at 2201 [EDT].
"Following inspection by electrical maintenance the 'P' 480V bus was re-energized at 2305 [EDT] and restoration of previously running loads is in progress.
"Station (TMI) is in a 72-hour LCO for repairs to the Decay Closed Cooling Water Pump '1A' (DC-P-1A).
"TMI (Three Mile Island) is terminating the event based upon the above information."
The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin. The licensee will issue a press release.
The licensee informed the NRC Resident Inspector. Notified R1DO (Bickett), NRR (Howe) and IRD (Gott).
Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
Power Reactor
Event Number: 51456
Facility: PILGRIM
Region: 1 State: MA
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: KENNETH GRACIA
HQ OPS Officer: MARK ABRAMOVITZ
Region: 1 State: MA
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: KENNETH GRACIA
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 10/05/2015
Notification Time: 21:48 [ET]
Event Date: 10/05/2015
Event Time: 17:25 [EDT]
Last Update Date: 10/05/2015
Notification Time: 21:48 [ET]
Event Date: 10/05/2015
Event Time: 17:25 [EDT]
Last Update Date: 10/05/2015
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):
BRICE BICKETT (R1DO)
BRICE BICKETT (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 |
SAFE SHUTDOWN ANALYSIS FOR FIRE EVENT - OPEN ITEM WITH POTENTIAL TO IMPACT CAPABILITY TO OPERATE MOTOR OPERATED VALVES
"At 1725 [EDT] on Monday, October 5, 2015, while updating the site fire safe shutdown analysis report, Pilgrim Nuclear Power Station (PNPS) identified vulnerabilities in two (2) specific fire areas that required a compensatory fire watch. The compensatory fire watches have already been established.
"The specific concern involves the unlikely scenario that fire forces evacuation of the control room. In addition, it is postulated that motor operated valves cannot be operated from their alternate control location due to spurious operation caused by hot shorts in control wiring. The specific failure mode is described in NRC Information Notice (IN) 92-18.
"Engineering modifications are under development and scheduled for implementation to address this specific IN 92-18 concern.
"Based on the above, the condition is reportable to the NRC as an unanalyzed condition that significantly degrades plant safety. As such the condition is reportable under 50.72(b)(3)(ii)(B), and requires an 8 hour notification.
"The licensee has notified the NRC Resident Inspector."
"At 1725 [EDT] on Monday, October 5, 2015, while updating the site fire safe shutdown analysis report, Pilgrim Nuclear Power Station (PNPS) identified vulnerabilities in two (2) specific fire areas that required a compensatory fire watch. The compensatory fire watches have already been established.
"The specific concern involves the unlikely scenario that fire forces evacuation of the control room. In addition, it is postulated that motor operated valves cannot be operated from their alternate control location due to spurious operation caused by hot shorts in control wiring. The specific failure mode is described in NRC Information Notice (IN) 92-18.
"Engineering modifications are under development and scheduled for implementation to address this specific IN 92-18 concern.
"Based on the above, the condition is reportable to the NRC as an unanalyzed condition that significantly degrades plant safety. As such the condition is reportable under 50.72(b)(3)(ii)(B), and requires an 8 hour notification.
"The licensee has notified the NRC Resident Inspector."
Power Reactor
Event Number: 51450
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: LINDSAY GREEN
HQ OPS Officer: VINCE KLCO
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: LINDSAY GREEN
HQ OPS Officer: VINCE KLCO
Notification Date: 10/05/2015
Notification Time: 08:41 [ET]
Event Date: 10/05/2015
Event Time: 01:05 [CDT]
Last Update Date: 10/05/2015
Notification Time: 08:41 [ET]
Event Date: 10/05/2015
Event Time: 01:05 [CDT]
Last Update Date: 10/05/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
STEVE ORTH (R3DO)
STEVE ORTH (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Hot Standby | 0 | Hot Standby |
SPECIFIED SYSTEM ACTUATIONS
"Braidwood Unit 2 was performing a planned plant shutdown for refueling outage A2R18. In accordance with plant shutdown procedures while in Mode 1 (Power Operations) at approximately 15% power, operators attempted to start the Start Up Feedwater (SFWP) pump and the pump immediately tripped on Phase A Overcurrent. The 2A Motor Driven Feedwater pump (MDFWP) was manually started to maintain Steam Generator Water Level during the shutdown and subsequent plant cooldown. While in Mode 3 (Hot Standby) at [550 Degree-F], the 2A MDFWP was manually secured due to pump inboard journal bearing temperature exceeding its [200 Degree-F] operating limit. At 0105 [CDT] an anticipated automatic Auxiliary Feedwater actuation signal was generated on low Steam Generator level (36.3%) and both the 2A and 2B Auxiliary Feedwater pumps (AFP) auto-started. Also at 0105 [CDT] a Reactor Protection System (RPS) Reactor trip signal was received due to low Steam Generator level (36.3%) with the reactor not critical. Both Auxiliary Feedwater trains operated as designed with the Main Steam Dumps in service and the Main Condenser providing the heat sink.
"All systems operated as designed with the exception of the SFWP and the MDFWP described above.
"The plant is currently stable in Mode 5 with both AFPs secured.
"This report is being made per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the (1) RPS Reactor Trip with the reactor not critical and (6) Auxiliary Feedwater System, 8 hour notification."
The licensee notified the NRC Resident Inspector.
"Braidwood Unit 2 was performing a planned plant shutdown for refueling outage A2R18. In accordance with plant shutdown procedures while in Mode 1 (Power Operations) at approximately 15% power, operators attempted to start the Start Up Feedwater (SFWP) pump and the pump immediately tripped on Phase A Overcurrent. The 2A Motor Driven Feedwater pump (MDFWP) was manually started to maintain Steam Generator Water Level during the shutdown and subsequent plant cooldown. While in Mode 3 (Hot Standby) at [550 Degree-F], the 2A MDFWP was manually secured due to pump inboard journal bearing temperature exceeding its [200 Degree-F] operating limit. At 0105 [CDT] an anticipated automatic Auxiliary Feedwater actuation signal was generated on low Steam Generator level (36.3%) and both the 2A and 2B Auxiliary Feedwater pumps (AFP) auto-started. Also at 0105 [CDT] a Reactor Protection System (RPS) Reactor trip signal was received due to low Steam Generator level (36.3%) with the reactor not critical. Both Auxiliary Feedwater trains operated as designed with the Main Steam Dumps in service and the Main Condenser providing the heat sink.
"All systems operated as designed with the exception of the SFWP and the MDFWP described above.
"The plant is currently stable in Mode 5 with both AFPs secured.
"This report is being made per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the (1) RPS Reactor Trip with the reactor not critical and (6) Auxiliary Feedwater System, 8 hour notification."
The licensee notified the NRC Resident Inspector.