| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
04/06/2012 - 04/09/2012 ** EVENT NUMBERS ** |
 | Agreement State | Event Number: 47796 | Rep Org: ILLINOIS EMERGENCY MGMT. AGENCY
Licensee: MIDWEST REGIONAL MEDICAL CENTER
Region: 3
City: ZION State: IL
County:
License #: IL-01104-01
Agreement: Y
Docket:
NRC Notified By: DAREN PERRERO
HQ OPS Officer: PETE SNYDER | Notification Date: 04/02/2012
Notification Time: 17:38 [ET]
Event Date: 03/30/2012
Event Time: [CDT]
Last Update Date: 04/02/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
BILLY DICKSON (R3DO)
CHRISTEPHER MCKENNEY (FSME)
| Event Text AGREEMENT STATE REPORT - THERASPHERES FOUND IN DELIVERY EQUIPMENT AFTER PROCEDURE
The following information was received from the State of Illinois via email:
"On April 2, 2012 the licensee contacted the [Illinois Emergency Management] Agency [IEMA] to provide a preliminary notification that they believed a medical event had occurred as a result of a Therasphere treatment the previous workday. On March 30, 2012 a patient had been administered Y-90 for the treatment of liver cancer according to their standard protocol. No unusual anomalies took place during the treatment and no shunting of the dose was observed. The treatment was completed as expected including initial agitation of the vial, 'tapping' of the system during administration and repeated flushing of the delivery system upon completion of the five minute process.
"However, when post-treatment measurements were conducted of the delivery system and vial, a measurable amount of activity was determined to still be present. Analysis of the associated microcatheter, tubing, vial and the rest of the delivery system resulted in an assessment that only 92 of the prescribed 130 Gy [71%] of dose had been delivered. Additional measurements and imaging identified that an aggregate of microspheres remained in the vial and at the hub of the microcatheter where it connects to the delivery system. No other activity or contamination was noted in associated equipment.
"The patient was advised the next day of the substantial lowering of the dose delivered. No immediate effect on the patient is expected at this time and no determination has been made as to any corrective action or additional dose to supplement the partial treatment. A follow up with the patient is pending. Similarly, the licensee is unaware of any corrective action necessary for future treatments pending additional evaluations and assessments. Consideration is being given to having the delivery system returned to the manufacturer for an engineering analysis once the radioactivity present has decayed. An investigation by IEMA is pending."
IL Item Number: IL12007
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Power Reactor | Event Number: 47800 | Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: JEFF GROFF
HQ OPS Officer: JOHN KNOKE | Notification Date: 04/03/2012
Notification Time: 05:57 [ET]
Event Date: 04/03/2012
Event Time: 05:24 [EDT]
Last Update Date: 04/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
BILLY DICKSON (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Refueling | 0 | Refueling | Event Text SPDS AND ERDS REMOVED FROM SERVICE FOR PLANNED MAINTENANCE
"On 4/03/12 at 0524 EDT, the Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were removed from service to support activities for a planned maintenance outage on the UPS vital bus power supply. The duration of work is expected to be approximately 80 hours. During this time, the majority of the control room indications remain available to the plant staff, and will be used for emergency response, if needed. Information will be communicated to the NRC using other available communication systems, if needed. The plant is currently in Mode 5, and will remain in Mode 5, for the duration of the SPDS and ERDS unavailability. Since the unavailability will last greater than 8 hours, this is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii).
"Follow up notification will be made when SPDS and ERDS have been restored."
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM GEORGE PICCARD TO VINCE KLCO ON 4/6/2012 AT 1227 EDT * * *
"At 0702 EDT on April 6, 2012, Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were restored, following restoration of the associated power bus."
The licensee notified the NRC Resident Inspector.
Notified the R3DO (Bloomer) |
| Power Reactor | Event Number: 47810 | Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [2] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: STEPHEN NEWMAN
HQ OPS Officer: PETE SNYDER | Notification Date: 04/06/2012
Notification Time: 03:57 [ET]
Event Date: 04/05/2012
Event Time: 22:38 [EDT]
Last Update Date: 04/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
GERALD MCCOY (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 85 | Power Operation | Event Text STANDBY SHUTDOWN FACILITY (SSF) NOT ANALYZED FOR ALL OPERATING CONDITIONS
"On 3/29/2012 Duke Energy identified that unanalyzed conditions exist for SSF mitigated events since associated thermal and hydraulic analyses do not consider all initial operating conditions, especially lower operating modes and lower decay heat. Specifically there are four (4) conditions where the SSF is not currently analyzed:
"1. SSF operating at less than 525 degrees F and less than normal operating pressure (approximately 2155 psig),
2. SSF operation before four (4) Effective Full Power Days (EFPDs),
3. SSF reactor coolant make up at low Reactor Coolant System (RCS) pressure.
4. A reactor trip from less than 85 percent power and less than 579 degrees F.
"On 4/4/2012, an immediate determination of operability concluded that for the first three (3) conditions the SSF was operable but degraded/nonconforming (OBDN). For the 4th issue, there was reasonable assurance that 1% delta k/k shutdown margin would be maintained if T ave. remained above 500 degrees F. Based on a lack of analysis and an increased likelihood of reducing T ave. below 525 degrees F during a 72 hour event, the SSF was declared OBDN with a separate operability determination required to validate the Unit 3 power coastdown and end of life T ave. reduction analysis. Until additional analysis is performed, the SSF is inoperable on any unit where the power level is reduced below 85 percent.
"A second operability determination for Unit 3 concluded that the SSF will maintain greater than or equal to 525 degrees F with an initial power level of 70 percent and a 570 degree F T ave. The SSF will be declared inoperable on Unit 3 if power is reduced to less than 70 percent. Seventy percent was chosen as a conservative value to ensure the unit stayed inside the bounds of existing analyses. Unit 3 is currently at approximately 85 percent power and reducing power at approximately 1 percent per day in preparation for the Unit 2 end of core 26 refueling outage. For Unit 3, the SSF is OBDN based on preliminary calculation results.
"On 4/5/2012, due to a worsening component cooling water system leak on Unit 2, it was necessary to bring the unit down to Mode 3 to implement repairs. Upon down power, when Unit 2 transitioned below 85 percent power, the ability of the SSF to perform its design function, in consideration of the information above, could not be confirmed and the SSF was declared inoperable for Unit 2.
"Currently, there is no conclusive information that would support SSF operability while Unit 2 is below 85 percent power. As such, this event is being conservatively reported under 50.72(b)(3)(ii)(B), 'The nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' Due to their current power levels, this condition does not affect Units 1 and 3.
"Initial Safety Significance: Until confirmed by analysis, the lack of decay heat may result in an initial over cooling of the RCS and potentially an interruption of natural circulation or inadequate shutdown margin. Consequently, the SSF was declared inoperable.
"Corrective Action(s): Additional analyses are being completed to reestablish SSF operability to bound the unanalyzed entry conditions.
"The NRC Resident Inspector has been informed." |
| Power Reactor | Event Number: 47811 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GUY GRIFFIS
HQ OPS Officer: DONALD NORWOOD | Notification Date: 04/06/2012
Notification Time: 08:41 [ET]
Event Date: 04/06/2012
Event Time: 04:43 [EDT]
Last Update Date: 04/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
GERALD MCCOY (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LOSS OF TONE ALERT SYSTEM
"On Friday April 6, 2012, at 0443 EDT, the National Weather Service (NWS) Jacksonville Florida office notified Hatch Nuclear Plant (HNP) of a loss of the NWS Tone Alert Weather Radio System. Site Emergency Procedures define a loss of Tone Alert System for greater than 15 minutes as a significant loss of Emergency Communications.
"The NWS Tone Alert Weather Radios are utilized as the Prompt Notification System (PNS) for HNP. This loss impacts the ability to notify the Emergency Planning Zone (EPZ) population for the HNP. This failure meets NRC 8 hour reporting criteria 10CFR50.72(b)(3)(xiii).
"The State of Georgia Department of Homeland Security / Georgia Emergency Management Agency (GEMA) 24-hour warning point and the '911' dispatch centers of Appling, Jeff Davis, Tattnall and Toombs counties were notified so compensatory measures would be available should the Prompt Notification System be needed. This consists of utilizing route alerting and the 'Code Red' system, which is a reverse 911 feature available from the county 911 center. This notification meets the 4-hour notification reporting criteria 10CFR50.72(b)(3)(xi).
"On Friday April 6, 2012, at 0503 EDT, the NWS Jacksonville Florida office notified HNP that the NWS Tone Alert Weather Radio System had been returned to service."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 47812 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: RON FRY
HQ OPS Officer: VINCE KLCO | Notification Date: 04/06/2012
Notification Time: 20:33 [ET]
Event Date: 04/06/2012
Event Time: 17:57 [EDT]
Last Update Date: 04/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
JOHN CARUSO (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text LOCAL LEAK RATE TEST RESULTS EXCEED PATHWAY TECHNICAL SPECIFICATION LIMITS
"Appendix J local leak rate testing has determined that secondary containment bypass leakage (SCBL) has been exceeded for Unit 1. During performance of leak rate test SE-159-026 for X-9A penetration it was determined the combined SCBL limit of 15 scfh [standard cubic feet per hour] for as-found minimum pathway was exceeded as specified in Tech Spec SR 3.6.1.3.11.
"Acceptance criteria test results were within acceptance criteria for the 10CFR50 Appendix J limits of 0.6 La.
"This event is being reported as a degraded or unanalyzed condition pursuant to 10CFR50.72(b)(3)(ii)."
Licensee corrective actions are to repair the identified valve seats.
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 47813 | Facility: FARLEY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DARRIN GARD
HQ OPS Officer: VINCE KLCO | Notification Date: 04/06/2012
Notification Time: 22:18 [ET]
Event Date: 04/06/2012
Event Time: 14:44 [CDT]
Last Update Date: 04/06/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
GERALD MCCOY (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text VALID LOAD SHED SIGNAL OCCURRED DUE TO LOSS OF THE 1B STARTUP TRANSFORMER
"At 1444 [CDT] on April 6, 2012, during a planned refueling outage on Unit 1, maintenance activities in the high voltage switchyard caused feeder breaker 820 to inadvertently trip. With the second feeder breaker, 924, already out of service, power was lost to the 1B startup transformer. An undervoltage condition was then experienced on the 1G 4160 V emergency bus. As a result, the B1G Sequencer initiated a valid load shed of the 1G 4160 V emergency bus. Due to outage conditions, the B-Train, 1B Emergency Diesel Generator (EDG) was tagged out and did not automatically start but did receive a valid start signal. None of the ESF loads supplied by the 1G bus started automatically since the 1B EDG was out of service. With a B-Train equipment outage in progress, the 1A RHR pump (A-Train) remained in service for shutdown cooling throughout the event. Although the bus safety function was not needed for plant conditions a valid load shed signal occurred and therefore this event is considered reportable.
"The 1G 4160 V emergency bus was restored to service at 1542 on April 6, 2012. Investigation revealed a technical inaccuracy in the instructions used during the maintenance activity in the high voltage switchyard that caused feeder breaker 820 to trip."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 47814 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: DAVE WALSH
HQ OPS Officer: VINCE KLCO | Notification Date: 04/07/2012
Notification Time: 15:27 [ET]
Event Date: 04/07/2012
Event Time: 13:54 [EDT]
Last Update Date: 04/07/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xii) - OFFSITE MEDICAL
| Person (Organization):
JOHN CARUSO (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text TRANSPORT OF POTENTIALLY CONTAMINATED WORKER
"On 04/07/2012 at 1354 [EDT], Susquehanna Steam Electric Station requested an offsite ambulance via the 911 system for medical assistance. The individual was in the radiologically controlled area and was treated as contaminated. An offsite ambulance arrived on site at 1413 hrs. and the ambulance departed the site at 1424 hrs. enroute to the Berwick Hospital.
"This is considered a transport of a contaminated individual requiring an 8 hour ENS Notification per 10CFR50.72(b)(3)(xii)."
Licensee health physic technicians accompanied the individual to the hospital.
The licensee notified the NRC Resident Inspector and the Pennsylvania Emergency Management Agency. |
| Power Reactor | Event Number: 47815 | Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: GREG SMITH
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 04/08/2012
Notification Time: 05:01 [ET]
Event Date: 04/08/2012
Event Time: 05:01 [EDT]
Last Update Date: 04/08/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
TAMARA BLOOMER (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 60 | Power Operation | 0 | Cold Shutdown | Event Text OFFSITE NOTIFICATION OF NOISE ASSOCIATED WITH PLANT SHUTDOWN
"Notified Van Buren County Sheriff of atmospheric steam dump valve usage for cooldown of Palisades Nuclear Plant for start of refueling outage."
The licensee will notify the NRC Resident Inspector. |
| Power Reactor | Event Number: 47816 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ALEX McLELLAN
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 04/09/2012
Notification Time: 02:07 [ET]
Event Date: 04/09/2012
Event Time: 01:02 [EDT]
Last Update Date: 04/09/2012 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JOHN CARUSO (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text BOTH CONTROL STRUCTURE CHILLERS DECLARED INOPERABLE WHILE SWITCHING POWER SUPPLIES
"On 4/9/2012, starting at 0102 EDT, the 'A' and 'C' Emergency Diesel Generators (EDG) were sequentially and briefly declared inoperable to switch their DC control power back to their normal supplies. Switching power to the normal supply is required by Unit 2 technical specification 3.8.4 following maintenance work on the U1 power supplies. Previously, at 18:35 EDT on 4/4/2012, the 'B' Control Structure Chiller was declared inoperable due to an unrelated problem. With the 'B' Control Structure Chiller inoperable coincident with the 'A' EDG or 'C' EDG inoperable, neither Control Structure Chiller would be available to perform its design function on a loss of offsite power. This is a condition that, at the time of discovery, could have prevented fulfillment of a Safety Function and is reportable under 50.72(b)(3)(v)(D) as an 8 hour notification.
"Switching the power supplies was a planned evolution. The duration of the loss of safety function was a total of eight minutes. As a mitigating action, operators were continuously available with communication to the control room. The associated diesel generator could have been returned to an operable condition promptly if required.
"Note that Technical Specifications allows four hours to correct the condition before further actions are required, i.e. declare the features ('A' Control Structure Chiller) supported by the inoperable diesel inoperable."
The licensee notified the NRC Resident Inspector. | |