Event Notification Report for December 29, 2012
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
12/28/2012 - 12/29/2012
EVENT NUMBERS
48637486384865048797
Power Reactor
Event Number: 48637
Facility: GRAND GULF
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: CHAD ROGERS
HQ OPS Officer: VINCE KLCO
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: CHAD ROGERS
HQ OPS Officer: VINCE KLCO
Notification Date: 12/29/2012
Notification Time: 05:05 [ET]
Event Date: 12/29/2012
Event Time: 00:18 [CST]
Last Update Date: 12/29/2012
Notification Time: 05:05 [ET]
Event Date: 12/29/2012
Event Time: 00:18 [CST]
Last Update Date: 12/29/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
HEATHER GEPFORD (R4DO)
HEATHER GEPFORD (R4DO)
| 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 |
AUTOMATIC REACTOR SCRAM DUE TO A TURBINE/GENERATOR TRIP
"Actuation of RPS [Reactor Protection System] with reactor critical. Reactor Scram occurred at 0018 [CST], 12/29/12, from 100% CTP [Core Thermal Power]. The cause of scram appears to be a Generator/Turbine trip.
"Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF [engineered safety feature] power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred.
"MSIVs [Main Steam Isolation Valves] remained open and SRVs [Safety Relief Valves] lifted. Currently, reactor water level is being maintained by the condensate and feedwater system in normal band and reactor pressure is being controlled to limit cooldown."
All control rods inserted. The plant is in hot shutdown with decay heat removal to the condenser and the electrical line-up is in a normal configuration. The cause of the turbine/generator trip is under investigation.
The licensee notified the NRC Resident Inspector.
"Actuation of RPS [Reactor Protection System] with reactor critical. Reactor Scram occurred at 0018 [CST], 12/29/12, from 100% CTP [Core Thermal Power]. The cause of scram appears to be a Generator/Turbine trip.
"Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF [engineered safety feature] power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred.
"MSIVs [Main Steam Isolation Valves] remained open and SRVs [Safety Relief Valves] lifted. Currently, reactor water level is being maintained by the condensate and feedwater system in normal band and reactor pressure is being controlled to limit cooldown."
All control rods inserted. The plant is in hot shutdown with decay heat removal to the condenser and the electrical line-up is in a normal configuration. The cause of the turbine/generator trip is under investigation.
The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 48638
Facility: SURRY
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BRET RICKERT
HQ OPS Officer: BILL HUFFMAN
Region: 2 State: VA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BRET RICKERT
HQ OPS Officer: BILL HUFFMAN
Notification Date: 12/29/2012
Notification Time: 15:21 [ET]
Event Date: 12/29/2012
Event Time: 08:03 [EST]
Last Update Date: 12/29/2012
Notification Time: 15:21 [ET]
Event Date: 12/29/2012
Event Time: 08:03 [EST]
Last Update Date: 12/29/2012
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):
BINOY DESAI (R2DO)
BINOY DESAI (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
EMERGENCY DIESEL GENERATOR AUTO-START AND LOAD UPON TRIP OF RESERVE STATION TRANSFORMER
"At 0803 EST on 12/29/2012 With Unit 1 and Unit 2 operating at 100%, the Supply breaker to 'B' Reserve Station Transformer (RSST), 0-EP-BKR-252 tripped on an instantaneous over current of the B-C phases due to a Pelican contacting the lines. This resulted in a lockout of the 'B' RSST and loss of power to the Unit 2 'H' Emergency Bus from its normal off site power supply. #2 Emergency Diesel Generator auto started on under voltage and loaded onto the Unit 2 'H' Emergency Bus. The Operations staff entered abnormal operating procedure 2-AP-10.07, Loss of Unit 2 Power, to recover from the electrical transient. All equipment performed as expected during the event and there were no unexplained occurrences. Maintenance has verified there is no damage to the line, breaker or 'B' RSST and at 1242 EST offsite power was restored to Unit 2 'H' Emergency Bus. At 1501, #2 Emergency Diesel Generator was secured in accordance with Station Operating procedures.
"Currently Unit 1 is operating at 100%, 922 MWE.
"Currently Unit 2 is operating at 100%, 915 MWE.
"All electrical distribution systems were restored to a normal alignment.
"No adverse radiological consequences resulted from this event. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events.
"This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to the auto start of #2 Emergency Diesel Generator."
The licensee has notified the NRC Resident Inspector.
"At 0803 EST on 12/29/2012 With Unit 1 and Unit 2 operating at 100%, the Supply breaker to 'B' Reserve Station Transformer (RSST), 0-EP-BKR-252 tripped on an instantaneous over current of the B-C phases due to a Pelican contacting the lines. This resulted in a lockout of the 'B' RSST and loss of power to the Unit 2 'H' Emergency Bus from its normal off site power supply. #2 Emergency Diesel Generator auto started on under voltage and loaded onto the Unit 2 'H' Emergency Bus. The Operations staff entered abnormal operating procedure 2-AP-10.07, Loss of Unit 2 Power, to recover from the electrical transient. All equipment performed as expected during the event and there were no unexplained occurrences. Maintenance has verified there is no damage to the line, breaker or 'B' RSST and at 1242 EST offsite power was restored to Unit 2 'H' Emergency Bus. At 1501, #2 Emergency Diesel Generator was secured in accordance with Station Operating procedures.
"Currently Unit 1 is operating at 100%, 922 MWE.
"Currently Unit 2 is operating at 100%, 915 MWE.
"All electrical distribution systems were restored to a normal alignment.
"No adverse radiological consequences resulted from this event. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events.
"This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to the auto start of #2 Emergency Diesel Generator."
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 48650
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: PAUL HERMANN
HQ OPS Officer: HOWIE CROUCH
Region: 2 State: AL
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: PAUL HERMANN
HQ OPS Officer: HOWIE CROUCH
Notification Date: 01/04/2013
Notification Time: 13:20 [ET]
Event Date: 12/29/2012
Event Time: 00:00 [CST]
Last Update Date: 01/04/2013
Notification Time: 13:20 [ET]
Event Date: 12/29/2012
Event Time: 00:00 [CST]
Last Update Date: 01/04/2013
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):
BINOY DESAI (R2DO)
PART 21 GRP (EMAI)
BINOY DESAI (R2DO)
PART 21 GRP (EMAI)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
PART 21 - ANTI-ROTATION PIN FAILURE IN 10-INCH ANCHOR DARLING (FLOWSERVE) DOUBLE DISC GATE VALVE
The following is a summary of the information received from TVA - Browns Ferry via facsimile:
On December 29, 2012 it was determined that sufficient data existed to determine a defect existed in a 10-inch Anchor Darling double disc gate valve installed in Browns Ferry Unit 1.
In November, 2012 during a scheduled refueling, the High Pressure Coolant Injection valve, 1-FCV-073-0002 failed its local leak rate test with gross leakage identified on the upstream (reactor side) disc. The downstream side of the discs showed acceptable leakage.
When maintenance was performed on the valve, the disc retainer bolt was found sheared. Further investigation revealed that the anti-rotation pin had failed. A review of work order history determined that this valve was installed during the Unit 1 recovery in 2007.
The cause of the anti-rotation pin failure was determined to be that the stem was not adequately torqued to the upper wedge at the manufacturing plant. This cause was confirmed by internal inspections of the valve that were performed at Browns Ferry. Flowserve Corporation is also conducting an investigation.
The licensee identified 15 other applications of this particular 10-inch valve in all three Browns Ferry units. For all those valves except 2-FCV-073-0002, documentation exists demonstrating that the stem to upper wedge was torqued to manufacturer requirements, determined through MOVATS [Motor-Operated Valve Analysis and Test System] testing that the anti-rotation pin is not sheared, or determined through visual inspection that the anti-rotation pin remains intact. 2-FCV-073-0002 is considered to be non-conforming and will be inspected during the next scheduled Unit 2 refueling.
The licensee has notified the NRC Resident Inspector of this notification.
The following is a summary of the information received from TVA - Browns Ferry via facsimile:
On December 29, 2012 it was determined that sufficient data existed to determine a defect existed in a 10-inch Anchor Darling double disc gate valve installed in Browns Ferry Unit 1.
In November, 2012 during a scheduled refueling, the High Pressure Coolant Injection valve, 1-FCV-073-0002 failed its local leak rate test with gross leakage identified on the upstream (reactor side) disc. The downstream side of the discs showed acceptable leakage.
When maintenance was performed on the valve, the disc retainer bolt was found sheared. Further investigation revealed that the anti-rotation pin had failed. A review of work order history determined that this valve was installed during the Unit 1 recovery in 2007.
The cause of the anti-rotation pin failure was determined to be that the stem was not adequately torqued to the upper wedge at the manufacturing plant. This cause was confirmed by internal inspections of the valve that were performed at Browns Ferry. Flowserve Corporation is also conducting an investigation.
The licensee identified 15 other applications of this particular 10-inch valve in all three Browns Ferry units. For all those valves except 2-FCV-073-0002, documentation exists demonstrating that the stem to upper wedge was torqued to manufacturer requirements, determined through MOVATS [Motor-Operated Valve Analysis and Test System] testing that the anti-rotation pin is not sheared, or determined through visual inspection that the anti-rotation pin remains intact. 2-FCV-073-0002 is considered to be non-conforming and will be inspected during the next scheduled Unit 2 refueling.
The licensee has notified the NRC Resident Inspector of this notification.
Part 21
Event Number: 48797
Rep Org: FLOWSERVE
Licensee: ANCHOR DARLING
Region: 1
City: RALEIGH State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAMES TUCKER
HQ OPS Officer: BILL HUFFMAN
Licensee: ANCHOR DARLING
Region: 1
City: RALEIGH State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAMES TUCKER
HQ OPS Officer: BILL HUFFMAN
Notification Date: 03/01/2013
Notification Time: 11:12 [ET]
Event Date: 12/29/2012
Event Time: 00:00 [EST]
Last Update Date: 07/11/2017
Notification Time: 11:12 [ET]
Event Date: 12/29/2012
Event Time: 00:00 [EST]
Last Update Date: 07/11/2017
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):
ART BURRITT (R1DO)
RANDY MUSSER (R2DO)
JAMNES CAMERON (R3DO)
DON ALLEN (R4DO)
NRC HQ PART 21 GROUP (EMAI)
ART BURRITT (R1DO)
RANDY MUSSER (R2DO)
JAMNES CAMERON (R3DO)
DON ALLEN (R4DO)
NRC HQ PART 21 GROUP (EMAI)
PART 21 - WEDGE PIN FAILURE IN ANCHOR DARLING MOTOR OPERATED DOUBLE DISC GATE VALVES WITH THREADED STEM TO UPPER WEDGE CONNECTIONS
The following is a summary of information received from Flowserve via facsimile:
"This is to notify the US Nuclear Regulatory Commission that, in accordance with the provisions of 10CFR Part 21, we have identified a potential issue and are submitting our evaluation of the event.
"Flowserve has been working with the Tennessee Valley Authority's (TVA) Browns Ferry Nuclear Plant to investigate the failure of a Size 10, Class 900 Anchor/Darling motor-operated double-disc gate valve. The failure was due to the shearing of the wedge pin which serves a joint locking function at the threaded interface between the valve stem and upper wedge. The pin is designed to ensure that the joint does not loosen due to vibration and other secondary loads. On some valve designs, the pin also is used to attach the disc retainers to the upper wedge. The pin shearing allowed rotation of the stem during the closing stroke when the valve was seating and ultimately resulted in loss of the stem to upper wedge joint integrity.
"Flowserve has completed an evaluation of the failure and concluded the root cause of the wedge pin failure was excessive load on the pin. The stem operating torque exceeded the torque to tighten the stem into the upper wedge before installation of the wedge pin. The additional stem torque produced a load on the wedge pin creating a stress which exceeded the pin shear strength causing the failure. The recommended assembly stem torque did not envelope the operating torque for the TVA application providing the potential for an over load situation and ultimate failure. The operating torque for the TVA valve was unusually high due to the fast closing time of the actuator and very conservative closing thrust margin.
"This situation can potentially occur on any Anchor/Darling type double-disc gate valve with a threaded stem to upper wedge connection, typically size 2.5" and larger, operated by an actuator that applies torque on the stem to produce the required valve operating thrust. An operating stem torque greater than the assembly stem torque can provide the opportunity for excessive pin load and potentially failure.
"We have reviewed our records, and the only similar wedge pin failure that we can identify, in addition to the Browns Ferry problems, is a sheared wedge pin at LaSalle Nuclear Station in 1993. Our investigation of the LaSalle failure concluded that the wedge pin failed due to excessive torque in the opening direction due to bonnet over pressurization.
"Flowserve recommends that all critical Anchor/Darling Double-Disc Gate valves with threaded stem to upper wedge connections and actuators that produce a torque on the stem be evaluated for potential wedge pin failure. Valves with electric motor actuators which produce high output torques are the most susceptible to failure. Valves which were assembled with stem torques that exceed the operating torque are not candidates for failure.
"Below is a list, based on our records, of customers, utilities and nuclear plants which were supplied with Anchor/Darling Double-Disc Gate valves with motor actuators on contracts with ASME Section III and/or 10 CFR 21 imposed.
"Flowserve plans to provide each of the customers identified [below] with a copy of this notification letter."
The following facilities in the United States may be affected:
ANO 1, Browns Ferry, Brunswick, Callaway, Catawba, Clinton, Columbia, Cook, Cooper, Crystal River, Dresden, Diablo Canyon, Duane Arnold, Fitzpatrick, Fort Calhoun, Grand Gulf, Hatch, Indian Point, Kewaunee, LaSalle, Limerick, Maine Yankee, Millstone, Monticello, Nine Mile, North Anna, Oconee, Oyster Creek, Peach Bottom, Perry, Pilgrim, Prairie Island, Quad Cities, River Bend, Robinson, San Onofre, St. Lucie, Surry, Three Mile Island 2, Waterford, VC Summer, Vermont Yankee, Wolf Creek.
See Related Part 21 EN #48650.
* * * UPDATE AT 1537 EDT ON 7/11/17 FROM AMY OATHOUT TO JEFF HERRERA * * *
The following is a summary of a report provided by Flowserve via email:
This is to notify the US Nuclear Regulatory Commission that, in accordance with the provisions of 10 CFR Part 21, Flowserve has gained additional insight and information concerning the referenced previously reported issue based on a recent incident at the LaSalle County Station, Unit 2 involving a similar valve.
Evaluation of a similar incident at LaSalle added an element not addressed in the previous evaluation regarding the limitation of a pressed-on stem collar to support the actuator thrust and maintain the stem-wedge preload.
Valve evaluations and actions resulting from the previous notification are applicable and still apply. This notification includes additional information for maintaining the stem preload that was not addressed previously. The actuator thrust as well as the torque must be reviewed to insure the preload is maintained.
A list of customers, utilities and nuclear plants which were supplied with Anchor/Darling DD Gate valves with motor actuators on contracts with Section 111 and/or 10CFR21 was provided. This list added a few sites not included on the list provided with the original notification.
For questions or additional information please contact:
Joseph Carter
Manager, Quality Assurance
Flowserve Corporation, FCD
Raleigh, NC
919-831-3220
Notified R1DO(Dimitriadis), R2DO(Bonser), R3DO(Peterson), R4DO(Proulx), Part 21 group (via email).
The following is a summary of information received from Flowserve via facsimile:
"This is to notify the US Nuclear Regulatory Commission that, in accordance with the provisions of 10CFR Part 21, we have identified a potential issue and are submitting our evaluation of the event.
"Flowserve has been working with the Tennessee Valley Authority's (TVA) Browns Ferry Nuclear Plant to investigate the failure of a Size 10, Class 900 Anchor/Darling motor-operated double-disc gate valve. The failure was due to the shearing of the wedge pin which serves a joint locking function at the threaded interface between the valve stem and upper wedge. The pin is designed to ensure that the joint does not loosen due to vibration and other secondary loads. On some valve designs, the pin also is used to attach the disc retainers to the upper wedge. The pin shearing allowed rotation of the stem during the closing stroke when the valve was seating and ultimately resulted in loss of the stem to upper wedge joint integrity.
"Flowserve has completed an evaluation of the failure and concluded the root cause of the wedge pin failure was excessive load on the pin. The stem operating torque exceeded the torque to tighten the stem into the upper wedge before installation of the wedge pin. The additional stem torque produced a load on the wedge pin creating a stress which exceeded the pin shear strength causing the failure. The recommended assembly stem torque did not envelope the operating torque for the TVA application providing the potential for an over load situation and ultimate failure. The operating torque for the TVA valve was unusually high due to the fast closing time of the actuator and very conservative closing thrust margin.
"This situation can potentially occur on any Anchor/Darling type double-disc gate valve with a threaded stem to upper wedge connection, typically size 2.5" and larger, operated by an actuator that applies torque on the stem to produce the required valve operating thrust. An operating stem torque greater than the assembly stem torque can provide the opportunity for excessive pin load and potentially failure.
"We have reviewed our records, and the only similar wedge pin failure that we can identify, in addition to the Browns Ferry problems, is a sheared wedge pin at LaSalle Nuclear Station in 1993. Our investigation of the LaSalle failure concluded that the wedge pin failed due to excessive torque in the opening direction due to bonnet over pressurization.
"Flowserve recommends that all critical Anchor/Darling Double-Disc Gate valves with threaded stem to upper wedge connections and actuators that produce a torque on the stem be evaluated for potential wedge pin failure. Valves with electric motor actuators which produce high output torques are the most susceptible to failure. Valves which were assembled with stem torques that exceed the operating torque are not candidates for failure.
"Below is a list, based on our records, of customers, utilities and nuclear plants which were supplied with Anchor/Darling Double-Disc Gate valves with motor actuators on contracts with ASME Section III and/or 10 CFR 21 imposed.
"Flowserve plans to provide each of the customers identified [below] with a copy of this notification letter."
The following facilities in the United States may be affected:
ANO 1, Browns Ferry, Brunswick, Callaway, Catawba, Clinton, Columbia, Cook, Cooper, Crystal River, Dresden, Diablo Canyon, Duane Arnold, Fitzpatrick, Fort Calhoun, Grand Gulf, Hatch, Indian Point, Kewaunee, LaSalle, Limerick, Maine Yankee, Millstone, Monticello, Nine Mile, North Anna, Oconee, Oyster Creek, Peach Bottom, Perry, Pilgrim, Prairie Island, Quad Cities, River Bend, Robinson, San Onofre, St. Lucie, Surry, Three Mile Island 2, Waterford, VC Summer, Vermont Yankee, Wolf Creek.
See Related Part 21 EN #48650.
* * * UPDATE AT 1537 EDT ON 7/11/17 FROM AMY OATHOUT TO JEFF HERRERA * * *
The following is a summary of a report provided by Flowserve via email:
This is to notify the US Nuclear Regulatory Commission that, in accordance with the provisions of 10 CFR Part 21, Flowserve has gained additional insight and information concerning the referenced previously reported issue based on a recent incident at the LaSalle County Station, Unit 2 involving a similar valve.
Evaluation of a similar incident at LaSalle added an element not addressed in the previous evaluation regarding the limitation of a pressed-on stem collar to support the actuator thrust and maintain the stem-wedge preload.
Valve evaluations and actions resulting from the previous notification are applicable and still apply. This notification includes additional information for maintaining the stem preload that was not addressed previously. The actuator thrust as well as the torque must be reviewed to insure the preload is maintained.
A list of customers, utilities and nuclear plants which were supplied with Anchor/Darling DD Gate valves with motor actuators on contracts with Section 111 and/or 10CFR21 was provided. This list added a few sites not included on the list provided with the original notification.
For questions or additional information please contact:
Joseph Carter
Manager, Quality Assurance
Flowserve Corporation, FCD
Raleigh, NC
919-831-3220
Notified R1DO(Dimitriadis), R2DO(Bonser), R3DO(Peterson), R4DO(Proulx), Part 21 group (via email).