| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/17/2008 - 03/18/2008 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 43940 | Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RICKY GIVENS
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 01/29/2008
Notification Time: 00:51 [ET]
Event Date: 01/28/2008
Event Time: 18:43 [CST]
Last Update Date: 03/17/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
BINOY DESAI (R2)
|
| 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 HIGH PRESSURE CORE INJECTION (HPCI) INOPERABLE
"On 1/28/08 at 1843 CST, Browns Ferry Unit 1 was performing 1-SR-3.3.5.1.3(D) HPCI System Condensate Header Low Level Switch Calibration and Functional Test when 1-LS-73-56A failed to actuate. Per TS 3.3.5.1, 1-LS-73-56A is inoperable. 1-SR-3.3.5.1.3(D) defeats the logic relay normally actuated by 73-56A & B. This causes HPCI to be inoperable per TS 3.3.5.1.D if the relay is defeated for greater than 1 hour. Failure of the 73-56A switch prevented restoration of the relay within the 1 hour time frame.
"This event is reportable under 10CFR 50.72(b)(3)(v)(B) 'any event or condition that at the time of discovery could have prevented the fulfillment of the Safety Function of structures or systems that are needed to: Remove Residual Heat' and 10CFR 50.72(b)(3)(v)(D) 'any event or condition that at the time of discovery could have prevented the fulfillment of the Safety Function of structures or systems that are needed to: mitigate the consequences of an accident.'
"This event also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(v)(B) and 10CFR 50.73(a)(2)(v)(D).
"The defeated relay was restored to normal and the HPCI system returned to operable status at 2330 CST on 1/28/08."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 1359 EDT ON 3/17/08 FROM RASMUSSEN TO HUFFMAN * * *
"On January 28, 2008, Browns Ferry Unit 1 entered an LCO to perform a planned maintenance activity, High Pressure Coolant Injection System Condensate Header Low Level Switch Calibration and Functional Test, 1-SR-3.5.5.1.3(D). During the calibration of 1-LS-073-0056A and 1-LS-073- 0056B, 1-LS-073-0056A was found inoperable. The removal of both level switches from service (and as a result the HPCI transfer on low condensate header level function) was a planned maintenance activity, performed in accordance with an approved procedure and in accordance with the plants TSs. During this time no condition was discovered that could have prevented HPCI from performing its intended function because 1-LS-073-056B was considered OPERABLE. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(B) or 10 CFR 50.72(b)(3)(v)(D)."
The licensee notified the NRC Resident Inspector. R2DO(Lesser) notified. |
| General Information or Other | Event Number: 44058 | Rep Org: OR DEPT OF HEALTH RAD PROTECTION
Licensee: UNKNOWN
Region: 4
City: PORTLAND State: OR
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TERRY LINDSEY
HQ OPS Officer: BILL HUFFMAN | Notification Date: 03/12/2008
Notification Time: 18:22 [ET]
Event Date: 03/12/2008
Event Time: 11:30 [PDT]
Last Update Date: 03/13/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
TROY PRUETT (R4)
ANNA BRADFORD (FSME)
| Event Text AGREEMENT STATE - UNIDENTIFIED RADIOACTIVE MATERIAL RECEIVED AT METAL RECYCLING FACILITY
The State of Oregon Radiation Services Section reported that they had sent an inspector to Schnitzer Steel in Portland, Oregon due to unidentified radioactive material that had been received by the facility. A barge of metal for recycling had been received from Amix Recycling of Vancouver, British Columbia that set off a radioactive portal alarm. The radioactive material has not yet been specifically identified. The State inspectors have determined that one of the radioactive sources appears to be a process gauge with a Cesium source. There is also indication of a Radium or Cesium source in a scrap metal pipe. The State will ensure that the sources are adequately secured and will be following the licensees interactions with the originator of the shipment in terms of long term disposition of the sources.
Oregon Report Number 08-0025
* * * UPDATE AT 1921EDT ON 03/13/08 FROM STATE OF OREGON (TERRY LINDSEY) TO STEVE SANDIN VIA EMAIL * * *
"Oregon Radiation Protection Services personnel have completed an initial survey of the empty barge that contained two radiation gauges with Cesium-137. One gauge (mat scanner) was sheared apart with scrap steel equipment and was leaking along both shear points. Both gauges have been secured and contamination has been contained at this time.
"The Oregon 102nd Civil Support Team also assisted with the final survey and decontamination evaluation onsite and the site has been secured for today.
"A U.S. EPA Region X team has been activated by U.S. EPA Duty Officer and is enroute to Oregon to assist with further site evaluation and investigation. Oregon RPS personnel will be coordinating with the U.S. EPA team tomorrow morning for any final survey work and evaluation of potential enforcement concerns related to removal of radiation warning labels from these gauges."
Notified R4DO (Pruett) and FSME (Bradford). |
| General Information or Other | Event Number: 44061 | Rep Org: COLORADO DEPT OF HEALTH
Licensee: DENVER HEALTH MEDICAL CENTER
Region: 4
City: DENVER State: CO
County:
License #: 097-04
Agreement: Y
Docket:
NRC Notified By: PHILIP EGIDI
HQ OPS Officer: STEVE SANDIN | Notification Date: 03/13/2008
Notification Time: 17:05 [ET]
Event Date: 02/29/2008
Event Time: 11:00 [MDT]
Last Update Date: 03/13/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
TROY PRUETT (R4)
ANNA BRADFORD (FSME)
| Event Text AGREEMENT STATE REPORT INVOLVING A SPILL OF Tc-99m DURING ADMINISTRATION OF A DIAGNOSTIC DOSE
The following report was received via email:
"The Colorado Radiation Control Program (RAM Unit) received a call from the licensee on February 29, 2008 at approximately 13:15 that a small spill had occurred as part of a cardiac treadmill test. We [Colorado Rad Control Program] received written communication on February 29, 2008.
"The diagnostic dose prescribed for the patient was 31.4 mCi of Tc-99m (Cardiolite). The patient was walking on the treadmill. During the initial attempt of injection of the nuclide into the patient via the catheter, part of the dose (~2 mCi) spilled out at the juncture of the catheter and the needle into the patient resulting in a small spill onto the patient and down onto the treadmill. The test was stopped at that point and the patient was cleaned where the spill had occurred. There was no clothing contaminated. The attending staff was surveyed and none of the staff was contaminated.
"The area was monitored, ambient gamma exposure rates were about 50 uR/h, with the highest exposure rate on the impacted area of the treadmill reading ~3.5 mR/h. Readings returned to background within three feet of the spilled material. Due to the rough surface of the treadmill, remediation was not performed, rather the spill areas was covered with plastic sheeting and marked as a contaminated area. The treadmill was out of service until 3/3/08, which was longer than 10 half lives (they took advantage of the weekend). However, since the device was out of service for longer than 24 hours, this is reportable under Colorado Regulations Part 4.52.2.2, which are equivalent to NRC 10 CFR 30.50 (b)(1). The licensee has followed up with retraining of the nuclear medicine technician by reviewing the procedures for checking the IV tubing and making sure the proper connections are intact." |
| Power Reactor | Event Number: 44066 | Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: KEVIN R. BOSTON
HQ OPS Officer: PETE SNYDER | Notification Date: 03/15/2008
Notification Time: 11:19 [ET]
Event Date: 03/15/2008
Event Time: 09:33 [EDT]
Last Update Date: 03/17/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
TROY PRUETT (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 85 | Power Operation | 0 | Hot Standby | Event Text REACTOR TRIP DUE TO TURBINE TRIP
"At 09:33 EDT the plant experienced a turbine trip which resulted in a reactor trip from 85%. Plant power was less than 100% due to a [turbine] control valve oscillating problem identified on 3/13/08.
"Post-trip [complications]: one of the four 4160 V busses, Bus 1A1, did not fast transfer from 22 KV to the 161 KV system as expected. This buss loss caused Reactor Coolant Pump RC-3A [to] trip off. Seals on RC-3A appear to be failed. The other 3 RC pumps remained in service for forced circulation through the Steam Generators
"Steam Generator levels are being maintained by Main Feedwater Pump FW-4C. Steaming is through the steam dump and bypass valves to the condensers.
"No primary or secondary relief valves lifted during this event.
"Standard post-trip actions were taken per EOP-00 , followed by a transition to EOP-2. Loss of Off-Site Power. During EOP-00 the plant computer indicated 8 rods not inserted into the reactor core. Secondary and primary rod indication showed all control rods inserted. In response to the difference of rod position indication, emergency boration was implemented. Emergency boration was secured after shutdown margin was verified.
"4160 V buss 1A was reenergized on the 161kv system at 10:31 EDT.
"A transition to plant procedure OP-3A, plant shutdown was made at 10:58 EDT.
"Reactor Coolant Pump RC-3A seal evaluation is in progress. The plant will remain in Mode 3."
The licensee notified the NRC Resident Inspector and plans a press release.
While RCP RC-3A seals indicate the same pressure across them there were no accompanying indications of containment sump water level increasing or high radiation levels in containment.
Decay heat is being removed by normal feedwater feeding the steam generators steaming to the condenser steam dumps. Nuclear Instruments indicated plant shutdown throughout the trip. Safety buses are powered by offsite power and emergency diesel generators are in standby.
* * * UPDATE FROM ERICK MATZKE TO PETE SNYDER AT 1033 ON 3/17/08 * * *
"None of the reactor coolant pump seals at [Fort Calhoun Station] (FCS) were failed. There were abnormal indications on the A reactor coolant pump seal. The A reactor coolant pump seal is now working normally."
Notified R4DO (Whitten). |
| Power Reactor | Event Number: 44070 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: RICH KLINEFELTER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/17/2008
Notification Time: 11:50 [ET]
Event Date: 03/17/2008
Event Time: 11:50 [EDT]
Last Update Date: 03/17/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
NEIL PERRY (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 0 | Refueling | 0 | Refueling | Event Text SAFETY PARAMETER DISPLAY SYSTEM (SPDS) AND EMERGENCY RESPONSE DATA SYSTEM (ERDS) UNAVAILABLE DUE TO PLANNED MAINTENANCE
"At approximately 1200 hours, on 03/17/2008, the Unit 1 SPDS and ERDS system will be removed from service to connect a temporary power supply to support a planned maintenance outage on the PICSY [Plant Indication Computer System] computer normal power supply. The installation of temporary power is expected to have a duration greater than 8 hours, but less than 24. During this time, control room hardwire indications will be available. An update will be provided when SPDS/ERDS becomes available.
"Since the Unit 1 SPDS/ERDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10CFR50.72(b)(3)(xiii)."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 44071 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: DOUG LARIMER
HQ OPS Officer: BILL HUFFMAN | Notification Date: 03/17/2008
Notification Time: 15:27 [ET]
Event Date: 03/17/2008
Event Time: 08:00 [CDT]
Last Update Date: 03/17/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
STEVE ORTH (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Hot Standby | 0 | Hot Standby | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text TECHNICAL SUPPORT CENTER DEGRADED HABITABILITY CONTROL
"At approximately 0800 CDT on 3/17/08, operations personnel identified and investigated an issue with excessive noise/vibration of the Technical Support Center (TSC) cleanup fan. Investigation revealed that the TSC cleanup fan exhaust damper set screws, which attach the positioner to the damper shaft, had become loose and were slipping on the damper shaft. With this damper not adequately attached to the positioner the damper cannot be controlled and therefore flow through the cleanup filter cannot be controlled. Without adequate flow rates through the TSC cleanup filter, the TSC ventilation system is considered non-functional.
"As of 1400 CDT, repair and testing of the damper has been completed and the TSC ventilation is now functional. Had an emergency condition occurred during the time the repairs took place, contingency plans were in place to utilize the TSC as long as radiological conditions allowed. Procedure F3-6, 'Activation and Operation of the TSC,' Section 7.6, directs TSC management to relocate TSC activities to a radiological safe area if necessary."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 44072 | Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: TERRY DAMASHEK
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/17/2008
Notification Time: 16:30 [ET]
Event Date: 03/17/2008
Event Time: 13:00 [CDT]
Last Update Date: 03/17/2008 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
JACK WHITTEN (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP DUE TO LOWERING STEAM GENERATOR LEVEL DUE TO THE LOSS OF 'B' MAIN FEED WATER PUMP
"While operating at 100% rated thermal power in Mode 1, a Manual Reactor Trip was initiated due to the lowering of Steam Generator (S/G) level due to the loss of 'B' Main Feed Water Pump. Initial investigation indicates that the loss of XPB03 transformer caused the loss of Non-Class 1E 4160VAC PB03 and PB04. At the time of the event, bus PB04 was cross-tied with bus PB03 for scheduled maintenance of transformer XPB04. This caused loss of all condensate and heater drain pumps. A manual reactor trip was actuated in anticipation of an automatic reactor trip. Aux Feed auto actuation did occur as required. The Non-Safety Related Charging pump was lost due to the loss of PB03 and charging flow was re-established to the Reactor Coolant System by starting 'A' Charging pump. All other plant equipment functioned as required. The plant is currently stable in Mode 3 at 560 degrees F and 2235 psig. Continuing to investigate.
"At 15:03 CDT, plant experienced an auto Feed Water Isolation signal due to Low-Low S/G level. Feed Water Isolation had already occurred with the initial event but the signal had been previously been reset. Manual feed water flow control has been established from Aux Feed Water."
All control rods inserted into the core during the trip. Decay heat is being removed via the steam dumps to condenser using AFW to feed the steam generators. No primary or secondary relief valves lifted during the transient. The electrical grid is stable and supplying plant safety loads via the normal path.
NRC Resident has been notified. | |