| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/20/2013 - 05/21/2013 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 48845 | Facility: DRESDEN
Region: 3 State: IL
Unit: [ ] [ ] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: PATRICK HAARHOFF
HQ OPS Officer: PETE SNYDER | Notification Date: 03/22/2013
Notification Time: 19:58 [ET]
Event Date: 11/19/2012
Event Time: 19:00 [CDT]
Last Update Date: 05/20/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
DAVE PASSEHL (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text CONTROL ROD DRIVE CHECK VALVE IN SERVICE LEAK TEST FAILURE
"Control Rod Drive (CRD) Check valves 3-0399-593 and 3-0399-594 are installed on the CRD supply header inside the reactor building to prevent unfiltered radiological release from primary containment past secondary containment through the CRD supply header. The [Design Basis Accident] DBA [Loss of Coolant Accident] LOCA dose analysis does not postulate the CRD supply piping as a release path. These valves were added in response to NRC IN 90-78. Their function is described in the Dresden UFSAR section 4.6.4.6.
"CRD check valves 3-0399-593 and 3-0399-594 failed their as found In Service Testing (IST) seat leakage test during refuel outage D3R22.
"Based on the condition of as-found failed seat leakage tests, Engineering determined that CRD check valves 3-0399-593 and 3-0399-594 would not have been able to perform their design functions as described in the UFSAR to maintain the offsite dose and control room dose within regulatory limits.
"The valves were repaired in D3R22 and successfully retested."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION FROM RILEY RUFFIN TO HOWIE CROUCH ON 5/20/13 AT 1140 EDT * * *
"Based on an evaluation of the current plant design and the overly conservative assumptions made in the 1996 dose estimate, the dose that would be experienced as a result of a DBA LOCA with LOOP [Loss of Offsite Power] due to back leakage through the control rod drive system to the condensate storage tanks [CST] has been reassessed. The initial estimations of dose did not consider the as-built piping configuration of the CRD system. The backflow would be greatly reduced due to the lower differential pressure between the Drywell and CSTs that would exist following a DBA-LOOP. This differential pressure experienced following a DBA-LOOP would not result in a significant backflow due to excess flow check valves and restricting orifices in the current CRD piping configuration. Additionally, in the event that any backflow did occur, the volume of water in the CST would provide a high decontamination factor and dilution. Therefore backflows that could be experienced following a DBA-LOOP would not produce doses that would exceed current limits.
"Based on this assessment, the unanalyzed condition that was reported on March 22 (EN 48845) is being retracted."
The Licensee has notified the NRC Resident Inspector.
Notified R3DO (Kozak). |
| Power Reactor | Event Number: 49053 | Facility: PILGRIM
Region: 1 State: MA
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: DAVE NOYES
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/20/2013
Notification Time: 07:54 [ET]
Event Date: 05/20/2013
Event Time: 03:53 [EDT]
Last Update Date: 05/20/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
DAN SCHROEDER (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 0 | Startup | 0 | Refueling | Event Text OFF-SITE NOTIFICATION DUE TO FIRE IN TURBINE BUILDING
"At 0353 hours [EDT] on Monday, May 20, 2013 Pilgrim Station responded to indications of a fire in the Turbine Building (TB) Lubricating Oil Room. The Pilgrim Fire Brigade responded to the fire and was able to extinguish a small fire associated with the 'A' Auxiliary Oil Pump Motor. The plant was in start-up at the time of the event with the reactor critical and reactor coolant system temperature approximately 180 degrees F.
"The Plymouth Fire Department was contacted and responded to the site. The event did not require entry into the Emergency Action Levels (EALs). The plant is in a safe condition and plant personnel are investigating the cause. The plant will be restarted after a thorough evaluation and any necessary repairs are completed.
"This informational notification is being made in accordance with 10 CFR 50.72(b)(2)(xi). The licensee has notified the Massachusetts Emergency Management Agency (MEMA) .
"The Resident Inspector staff has been informed of this notification."
This was an electrical fire confined to the aux oil pump motor and was extinguished using hand held CO2 and dry chemical extinguishers. | |