Event Notification Report for October 19, 2014
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/18/2014 - 10/19/2014
Power Reactor
Event Number: 50550
Facility: SALEM
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DAN McHUGH
HQ OPS Officer: MARK ABRAMOVITZ
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DAN McHUGH
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 10/20/2014
Notification Time: 01:36 [ET]
Event Date: 10/19/2014
Event Time: 20:51 [EDT]
Last Update Date: 11/24/2014
Notification Time: 01:36 [ET]
Event Date: 10/19/2014
Event Time: 20:51 [EDT]
Last Update Date: 11/24/2014
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):
WILLIAM COOK (R1DO)
WILLIAM COOK (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 19 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP RESULTING IN AUX FEEDWATER ACTUATION
"This is an 8-hour notification being made to report that a valid ESF [emergency safety features] auxiliary feed water system actuation occurred. Salem Unit 1 was in mode 1 at 19% reactor power and executed a planned manual reactor trip to begin a scheduled refueling outage (1R23). The 11, 12, and 13 AFW [auxiliary feedwater] pumps were not in service prior to the reactor trip and as a result, narrow range levels in 12, 13, and 14 steam generators reached 14% before recovering. This resulted in a valid ESF actuation for low steam generator water level. Steam generator water levels were restored to normal post trip values as part of the procedurally directed response to a reactor trip.
"All safety related equipment was operational before the reactor trip and AFW actuation and responded as required. There were no equipment failures that contributed to this event. There were no personnel injuries as a result of' this event.
"The NRC Resident Inspector has been informed."
A transformer alarm caused the operators to trip the reactor prior to reaching the step in the procedure to start the AFW pumps.
* * * UPDATE FROM JOHN OSBORNE TO DONALD NORWOOD AT 1555 EST ON 11/24/2014 * * *
"On October 20, 2014 at 0136 EDT, Salem Unit 1 reported an 8 hour notification of a valid ESF Auxiliary Feed Water System actuation following a manual reactor trip at the start of the S1R23 refueling outage. After additional review it was determined that the manual reactor trip met the criteria for a four hour report in accordance with 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical...'."
The licensee notified the NRC Resident Inspector.
Notified R1DO (Noggle).
"This is an 8-hour notification being made to report that a valid ESF [emergency safety features] auxiliary feed water system actuation occurred. Salem Unit 1 was in mode 1 at 19% reactor power and executed a planned manual reactor trip to begin a scheduled refueling outage (1R23). The 11, 12, and 13 AFW [auxiliary feedwater] pumps were not in service prior to the reactor trip and as a result, narrow range levels in 12, 13, and 14 steam generators reached 14% before recovering. This resulted in a valid ESF actuation for low steam generator water level. Steam generator water levels were restored to normal post trip values as part of the procedurally directed response to a reactor trip.
"All safety related equipment was operational before the reactor trip and AFW actuation and responded as required. There were no equipment failures that contributed to this event. There were no personnel injuries as a result of' this event.
"The NRC Resident Inspector has been informed."
A transformer alarm caused the operators to trip the reactor prior to reaching the step in the procedure to start the AFW pumps.
* * * UPDATE FROM JOHN OSBORNE TO DONALD NORWOOD AT 1555 EST ON 11/24/2014 * * *
"On October 20, 2014 at 0136 EDT, Salem Unit 1 reported an 8 hour notification of a valid ESF Auxiliary Feed Water System actuation following a manual reactor trip at the start of the S1R23 refueling outage. After additional review it was determined that the manual reactor trip met the criteria for a four hour report in accordance with 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical...'."
The licensee notified the NRC Resident Inspector.
Notified R1DO (Noggle).
Part 21
Event Number: 50636
Rep Org: AREVA, INC.
Licensee: AREVA, INC.
Region: 1
City: LYNCHBURG State: VA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: GAYLE ELLIOTT
HQ OPS Officer: JOHN SHOEMAKER
Licensee: AREVA, INC.
Region: 1
City: LYNCHBURG State: VA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: GAYLE ELLIOTT
HQ OPS Officer: JOHN SHOEMAKER
Notification Date: 11/25/2014
Notification Time: 10:10 [ET]
Event Date: 10/19/2014
Event Time: 00:00 [EST]
Last Update Date: 11/25/2014
Notification Time: 10:10 [ET]
Event Date: 10/19/2014
Event Time: 00:00 [EST]
Last Update Date: 11/25/2014
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):
JAMES NOGGLE (R1DO)
FRANK EHRHARDT (R2DO)
ERIC DUNCAN (R3DO)
VIVIAN CAMPBELL (R4DO)
PART 21 GROUP (EMAI)
JAMES NOGGLE (R1DO)
FRANK EHRHARDT (R2DO)
ERIC DUNCAN (R3DO)
VIVIAN CAMPBELL (R4DO)
PART 21 GROUP (EMAI)
PART 21 - DEFECT IN LOCA ANALYSIS FOR B&W PLANTS
The following information was summarized from the report obtained from the vendor via facsimile:
"The defect is related to the thermal conductivity model in the codes TACO3 and GDTACO which are used in the AREVA LOCA model for B&W plants. The thermal conductivity model does not adequately represent the change in conductivity with burnup for the fuel. The correction of the thermal conductivity model results in the peak cladding temperature limit in 10 CFR 50.46 (2200 degrees F) being exceeded.
"The defect exists for the following plants: Oconee Nuclear Station Unit 1, Oconee Nuclear Station Unit 2, Oconee Nuclear Station Unit 3, Arkansas Nuclear One Unit 1, Three Mile Island Unit 1, and Davis-Besse Unit 1"
If there are any technical questions or concerns, please contact:
Gayle Elliott
AREVA, lnc., 3315 Old Forest Road
Lynchburg, VA 24501
Ph. # 434-841-0306.
The following information was summarized from the report obtained from the vendor via facsimile:
"The defect is related to the thermal conductivity model in the codes TACO3 and GDTACO which are used in the AREVA LOCA model for B&W plants. The thermal conductivity model does not adequately represent the change in conductivity with burnup for the fuel. The correction of the thermal conductivity model results in the peak cladding temperature limit in 10 CFR 50.46 (2200 degrees F) being exceeded.
"The defect exists for the following plants: Oconee Nuclear Station Unit 1, Oconee Nuclear Station Unit 2, Oconee Nuclear Station Unit 3, Arkansas Nuclear One Unit 1, Three Mile Island Unit 1, and Davis-Besse Unit 1"
If there are any technical questions or concerns, please contact:
Gayle Elliott
AREVA, lnc., 3315 Old Forest Road
Lynchburg, VA 24501
Ph. # 434-841-0306.