| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/18/2004 - 11/19/2004 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 41108 | Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [ ] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: KEVIN MOSES
HQ OPS Officer: HOWIE CROUCH | Notification Date: 10/10/2004
Notification Time: 00:51 [ET]
Event Date: 10/09/2004
Event Time: 19:45 [EDT]
Last Update Date: 11/18/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text REACTOR BUILDING PRESSURE DECREASED BELOW SPECIFIED LIMITS
The following information was obtained from the licensee via facsimile:
"On October 9, 2004 at 1945 hrs. EST, Operators discovered that Oconee Nuclear Station Unit 3 Reactor Building pressure decreased to less than the limit specified in Oconee Selected Licensee Commitment 16.6.13, 'Additional Requirements to Support Low Pressure Injection (LPI) Operability'. This initial Reactor Building Pressure is used in the NPSH Analysis for the LPI Pumps in the Sump Recirculation phase of post-LOCA operation. In addition to effecting LPl, this condition also affects the Reactor Building Spray system. Engineering Evaluation performed on May 25, 2004 identified that guidance contained within the Selected Licensee Commitment may be inadequate and as a result, the Reactor Building Spray and Low Pressure Injection systems were determined to be Operable But Degraded/Nonconforming. For the interim, until appropriate changes are made to the Selected Licensee Commitment, Engineering recommended that Oconee Operations enter Technical Specification 3.0.3 any time that the limits of Selected Licensee Commitment 16.6.13 are exceeded.
"At the time of discovery, Oconee Unit 3 was cooling down for refueling outage. Unit 3 was in Mode 4 at approximately 235 degrees, 275 psig with one train of Reactor Building Spray deactivated per the Controlling Procedure for Unit Shutdown. When it was identified that the limits of Selected Licensee Commitment 16.6.13 were exceeded, Operations began increasing Oconee Unit 3 Reactor Building pressure. Oconee Unit 3 Reactor Building pressure was restored to within the limits of Selected Licensee Commitment 16.6.13 on October 9, 2004 at 2023 hrs. EST.
"Initial Safety Significance:
"The NPSH analysis for the Low Pressure Injection pumps in the sump recirculation phase of post-LOCA operation credit reactor building overpressure of 2.2 psig as permitted by a license amendment granted July 19, 1999 and supplemented August 19, 1999. Operation with Reactor Building pressure less than the limits specified in Selected Licensee Commitment 16.6.13 cannot ensure that 2.2 psig overpressure will always be available.
"Corrective Action(s):
"Actions were taken by Operations to restore Unit 3 Reactor Building pressure to within the limits of Selected Licensee Commitment 16.6.13. These actions were successful in restoring Reactor Building Pressure to within limits of Selected Licensee Commitment 16.6.13 approximately 38 minutes from time of discovery."
The licensee has informed the NRC Resident Inspector.
* * * UPDATE 1849 EST ON 11/18/04 FROM R.P. TODD TO HOWIE CROUCH * * *
This report is retracted based on the following:
"Event: Withdrawal of Event Number 41108
"On October 9, 2004 at 1945 hrs. EST, Operators discovered that Oconee Nuclear Station Unit 3 Reactor Building (RB) pressure decreased to less than the limit specified in Oconee Selected Licensee Commitment (SLC) 16.6.13, 'Additional Requirements to Support Low Pressure Injection (LPI) Operability'. This initial RB Pressure (Containment overpressure) is credited in the NPSH Analysis for the LPI and RB Spray Pumps in the Sump Recirculation phase of post-LOCA operation. Engineering recommended that Oconee Operations enter Technical Specification 3.0.3 any time that the limits of SLC 16.6.13 are exceeded. Operations exited the condition after taking actions which raised RB pressure within the SLC limit. The ENS was reported as a loss of safety function under 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION.
"Justification for conclusion that there was no loss of safety functions:
"An engineering analysis determined that, at the time of this event, the worst case post-LOCA RB sump temperature would be approximately 170F. From a July 19, 1999 SER for the license amendment which credited containment overpressure, the NRC Staff concurred that containment overpressure credit was only needed for sump temperatures above 208F. Therefore, for the conditions which existed during this event, there was adequate NPSH, and no reportable loss of safety function existed."
The licensee informed the NRC Resident Inspector. Notified R2DO (Tom Decker). |
| General Information or Other | Event Number: 41196 | Rep Org: NV DIV OF RAD HEALTH
Licensee: LANDMARK TESTING
Region: 4
City: LAS VEGAS State: NV
County:
License #: 00-11-0518-01
Agreement: Y
Docket:
NRC Notified By: STAN MARSHALL
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/15/2004
Notification Time: 13:54 [ET]
Event Date: 11/11/2004
Event Time: 17:00 [PST]
Last Update Date: 11/15/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
KRISS KENNEDY (R4)
JOHN HICKEY (NMSS)
JOE FOSTER (TAS)
| Event Text STATE OF NEVADA AGREEMENT STATE REPORT: STOLEN TROXLER MOISTURE DENSITY GAUGE.
On November 11, 2004, at 1700 hours PST, a Troxler Model number 3440 (serial number 30887) moisture density gauge was stolen from a transport vehicle in Las Vegas, Nevada. Landmark Testing of Las Vegas, NV is the license holder of the gauge. The gauge contains 8 millicures of Cesium-137 and 40 millicuries of Americium-241:Be. Las Vegas Metro Police Department was notified of the stolen gauge.
Cause and Contributing Factors: The security measures for securing the gauge to the transport vehicle were defeated.
Corrective Actions: Licensee conducted additional training regarding transportation requirements.
Nevada Event Report ID No. NV-04-008. |
| General Information or Other | Event Number: 41202 | Rep Org: NV DIV OF RAD HEALTH
Licensee: KRAZAN & ASSOCIATES, INC.
Region: 4
City: RENO State: NV
County:
License #: CA 4247-10
Agreement: Y
Docket:
NRC Notified By: STAN MARSHALL
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/16/2004
Notification Time: 12:55 [ET]
Event Date: 11/14/2004
Event Time: 03:30 [PST]
Last Update Date: 11/16/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
DALE POWERS (R4)
MICHELE BURGESS (NMSS)
JOE FOSTER (TAS)
| Event Text STATE OF NEVADA AGREEMENT STATE REPORT: TROXLER MOISTURE DENSITY GAUGE STOLEN.
On November 16, 2004, Krazan & Associates, Inc., California Radioactive Material License # 4247-10, reported to the Nevada Health Division Radiological Health Section that one of their Troxler Moisture Density Gauges (Model # 3430, serial # 21922) was stolen. The gauge was stolen on November 14, 2004, between 0330 am and 0700 am PST from a job site in Reno, Nevada. Reno, Nevada Police Department was notified.
Cause & contributing factors: Security measures defeated.
Troxler Moisture Density Gauge Model # 3430 contains 9 millicuries of cesium-137 and 44 millicuries of Am:Be.
State of Nevada Event Report ID No. NV-04-009. |
| Power Reactor | Event Number: 41204 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [ ] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: TODD STRAIN
HQ OPS Officer: BILL HUFFMAN | Notification Date: 11/17/2004
Notification Time: 23:46 [ET]
Event Date: 11/17/2004
Event Time: 20:50 [CST]
Last Update Date: 11/18/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
JAMES CREED (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 0 | Hot Standby | Event Text TECHNICAL SPECIFICATION REQUIRED SHUT DOWN DUE TO BOTH TRAINS OF CONTAINMENT COOLING SYSTEM FAN COIL UNITS INOPERABLE
"Initiation of a Unit 2 shutdown commenced at 2050 on November 17, 2004, as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 because both trains of containment cooling were declared inoperable as of 1615. Each train of Prairie Island's Containment Cooling System includes two containment fan coil units (CFCUs), for a total of four CFCUs. TS 3.6.5 requires two trains of containment cooling to be operable. Leakage from the 23 CFCU was determined to render one train of containment cooling inoperable at 1551. At 1615, the 22 CFCU (one of two CFCUs in the opposite train of containment cooling) was also determined to be leaking rendering the second train of containment cooling inoperable. Two CFCUs are fully capable of removing the design basis heat load; however, as the two remaining operable CFCUs are in different trains, neither train of containment cooling could be considered operable. Both inoperable CFCUs have been isolated to restore containment integrity and facilitate repairs. TS LCO 3.6.5 does not have a Condition for both containment cooling trains inoperable, therefore TS LCO 3.0.3 was entered."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 41205 | Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: BRUCE BREDEMAN
HQ OPS Officer: BILL GOTT | Notification Date: 11/18/2004
Notification Time: 10:52 [ET]
Event Date: 09/18/2004
Event Time: 17:35 [CDT]
Last Update Date: 11/18/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
DALE POWERS (R4)
TERRY REIS (NRR)
HO NIEH (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 | Event Text DISCOVERY OF AFTER-THE-FACT EMERGENCY CONDITION (UNUSUAL EVENT)
"At 1735, 9/18/04, Callaway Control Room received an alarm on K0008 (Fire Protection Alarm and Control Workstation) for the Communication Corridor Lobby 2047 elevation. Initial investigation found significant smoke in the area of the Communication Corridor Elevator Room. During the investigation, a report was received in the Control Room that a small fire had occurred on the roof and had been extinguished. At 1743, the Callaway Fire Brigade was activated to investigate the source of the smoke. Upon arriving at the scene, no flames were observed and the Brigade began removing visquene [plastic] to facilitate observing additional areas within the room. At this time, small flames were observed and plant Emergency Alarm Level (EAL) 3E, 'Fire within the Protected Area boundary not extinguished within 15 minutes of verification' 15-minute time limit was entered. The fire was extinguished using a single powder type extinguisher and declared out at 1754.
At 1910, an Operator reported that the Communication Corridor fire had reflashed. The Fire Brigade was reactivated and the fire was extinguished at 1919 using a single powder type extinguisher. After determining the temperature of the elevator room ceiling was 250 degrees F, water was applied to the ceiling from inside the room and from the roof above. After verification that the fire was extinguished, a continuous fire watch was established for both the affected room and roof area. Initial evaluation of the fire event concluded that the fire was extinguished within 15 minutes of discovery, and due to the size and location of the fire, the fire did not pose any safety significance. No safety systems had been impacted, Control Room habitability was not impacted, no safety system functions were required, and no safety functions were lost. Based upon this evaluation, it was decided that in accordance with guidance provided by EAL 3E, declaration of an Unusual Event was not required.
Upon further review of the event associated with NRC Integrated Inspection Report 05000483/2004004, at 0925, 11/18/04, it was determined that a fire existed from the initial alarm at 1735 and lasted longer than 15 minutes. Because the initial fire lasted longer than 15 minutes, an Unusual Event should have been declared in accordance with plant Emergency Alarm Level 3E, 'Fire within the Protected Area boundary not extinguished within 15 minutes of verification.' Per guidance given in NUREG 1022, Revision 2, Section 3.1.1, 'Immediate Notifications', an ENS notification is being made within one hour of discovery of the event. No actual declaration of the emergency class will be made since the event has been concluded and the basis for the emergency no longer exists."
The licensee will inform both state and local agencies and has informed the NRC Resident Inspector. |
| Power Reactor | Event Number: 41207 | Facility: KEWAUNEE
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: DAVE KARST
HQ OPS Officer: HOWIE CROUCH | Notification Date: 11/18/2004
Notification Time: 18:59 [ET]
Event Date: 11/18/2004
Event Time: 16:30 [CST]
Last Update Date: 11/18/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
JAMES CREED (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text BOTH TRAINS OF RADIATION MONITORING AUTOMATIC ISOLATION FEATURES DISABLED DURING REFUELING OPERATIONS
The following information was obtained from the licensee via facsimile:
"At 1630 [hrs CST] the control room was performing head lift and pre-rod drive unlatching checkout and discovered R-12 (Containment Radiation Monitor) would not isolate reactor ventilation isolation. Investigation found that work on Engineered Safeguards Features pre-startup logic test had defeated both Train A and B automatic isolation signals to [Reactor Building Ventilation]. This test was started and signals defeated at 0803 [hrs.] and 0804 [hrs.]. Refueling evolutions, upper internals lift and rod latching were performed. Technical Specifications require auto-isolation capability for [Reactor Building Ventilation] during refueling ops.
"Both R-12 and R-21 would respond to high radiation and manual isolation of Reactor Building Ventilation was still available.
"Currently, refueling operations are stopped until auto-isolation of Reactor Building Ventilation is restored."
The licensee has determined that no increased gaseous or particulate radiation levels existed while the auto-isolation features were defeated.
The licensee has notified the NRC Resident Inspector. | |