Event Notification Report for March 22, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/21/2001 - 03/22/2001
EVENT NUMBERS
37858378604043138004
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37858
Facility: HATCH
Region: 2 State: GA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GUY GRIFFIS
HQ OPS Officer: FANGIE JONES
Region: 2 State: GA
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GUY GRIFFIS
HQ OPS Officer: FANGIE JONES
Notification Date: 03/22/2001
Notification Time: 16:10 [ET]
Event Date: 03/22/2001
Event Time: 11:10 [EST]
Last Update Date: 04/24/2001
Notification Time: 16:10 [ET]
Event Date: 03/22/2001
Event Time: 11:10 [EST]
Last Update Date: 04/24/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MARK LESSER (R2)
MARK LESSER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
RCIC DECLARED INOPERABLE
During an investigation of a ground on Div I 125/250 volt D.C. bus, the source of the ground was determined to be 2E51-F008 primary containment isolation valve. 2E51-F008 is the outboard steam supply to reactor core isolation coolant (RCIC). It was closed and de-energized. 2E51-F008 closure rendered RCIC system inoperable.
The licensee notified the NRC Resident Inspector.
* * * RETRACTION ON 04/24/01 AT 1453 ET BY S. HOWARD TAKEN BY MACKINNON * * *
Plant Hatch FSAR has been revised such that the Control Rod Drop accident limiting event path evaluated in the safety analysis (subsection 15.3.2) does not rely on RCIC to mitigate this Design Basis Accident (DBA). None of the DBAs, as evaluated in the safety analysis (section 15.3), rely on RCIC for event mitigation. RCIC will no longer be an NRC notification when declared inoperable. Therefore, this NRC notification is being retracted. R2DO (Paul Fredrickson) notified.
The NRC Resident Inspector was notified was notified of this retraction by the licensee.
During an investigation of a ground on Div I 125/250 volt D.C. bus, the source of the ground was determined to be 2E51-F008 primary containment isolation valve. 2E51-F008 is the outboard steam supply to reactor core isolation coolant (RCIC). It was closed and de-energized. 2E51-F008 closure rendered RCIC system inoperable.
The licensee notified the NRC Resident Inspector.
* * * RETRACTION ON 04/24/01 AT 1453 ET BY S. HOWARD TAKEN BY MACKINNON * * *
Plant Hatch FSAR has been revised such that the Control Rod Drop accident limiting event path evaluated in the safety analysis (subsection 15.3.2) does not rely on RCIC to mitigate this Design Basis Accident (DBA). None of the DBAs, as evaluated in the safety analysis (section 15.3), rely on RCIC for event mitigation. RCIC will no longer be an NRC notification when declared inoperable. Therefore, this NRC notification is being retracted. R2DO (Paul Fredrickson) notified.
The NRC Resident Inspector was notified was notified of this retraction by the licensee.
Power Reactor
Event Number: 37860
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [1] [] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: RUSSELL STROUD
HQ OPS Officer: FANGIE JONES
Region: 4 State: AZ
Unit: [1] [] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: RUSSELL STROUD
HQ OPS Officer: FANGIE JONES
Notification Date: 03/22/2001
Notification Time: 17:06 [ET]
Event Date: 03/22/2001
Event Time: 14:43 [MST]
Last Update Date: 03/22/2001
Notification Time: 17:06 [ET]
Event Date: 03/22/2001
Event Time: 14:43 [MST]
Last Update Date: 03/22/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
73.71(b)(1) - SAFEGUARDS REPORTS
10 CFR Section:
73.71(b)(1) - SAFEGUARDS REPORTS
Person (Organization):
JOHN PELLET (R4)
JOHN PELLET (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 97 | Power Operation | 97 | Power Operation |
SECURITY REPORT
Unescorted access was granted inappropriately. Immediate compensatory measures were taken upon discovery. (Contact the NRC operations center for additional details.) The licensee notified the NRC Resident Inspector.
Unescorted access was granted inappropriately. Immediate compensatory measures were taken upon discovery. (Contact the NRC operations center for additional details.) The licensee notified the NRC Resident Inspector.
Hospital
Event Number: 40431
Rep Org: ASSOCIATES AND MEDICAL PHYSICS
Licensee: DOUGLAS CITY HOSPITAL
Region: 3
City: ALEXANDRIA State: MN
County:
License #: 22-167-66-01
Agreement: N
Docket:
NRC Notified By: SUSAN WIESS
HQ OPS Officer: RICH LAURA
Licensee: DOUGLAS CITY HOSPITAL
Region: 3
City: ALEXANDRIA State: MN
County:
License #: 22-167-66-01
Agreement: N
Docket:
NRC Notified By: SUSAN WIESS
HQ OPS Officer: RICH LAURA
Notification Date: 01/08/2004
Notification Time: 17:00 [ET]
Event Date: 03/22/2001
Event Time: 13:30 [CST]
Last Update Date: 01/08/2004
Notification Time: 17:00 [ET]
Event Date: 03/22/2001
Event Time: 13:30 [CST]
Last Update Date: 01/08/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
Person (Organization):
ANTON VEGEL (R3)
CHARLIE MILLER (NMSS)
ANTON VEGEL (R3)
CHARLIE MILLER (NMSS)
MEDICAL EVENT AT DOUGLAS CITY HOSPITAL IN MINNESOTA
Associates and Medical Physics reported 6 related medical events that occurred between 3/22/01 thru 5/14/03 at Douglas City Hospital located in Alexandria, Minnesota. The licensee used Samarium 153 to treat patients for bone pain. The highest intended source strength was 115 milli-curies. The licensee discovered that the patients were actually under-dosed by 34% due to an incorrect dose calibrator setting. There was no significant adverse effects to the patients. Corrective actions were taken to calibrate the dose calibrator.
Associates and Medical Physics reported 6 related medical events that occurred between 3/22/01 thru 5/14/03 at Douglas City Hospital located in Alexandria, Minnesota. The licensee used Samarium 153 to treat patients for bone pain. The highest intended source strength was 115 milli-curies. The licensee discovered that the patients were actually under-dosed by 34% due to an incorrect dose calibrator setting. There was no significant adverse effects to the patients. Corrective actions were taken to calibrate the dose calibrator.
Power Reactor
Event Number: 38004
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: CHARLES ELBERFELD
HQ OPS Officer: JOHN MacKINNON
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: CHARLES ELBERFELD
HQ OPS Officer: JOHN MacKINNON
Notification Date: 05/17/2001
Notification Time: 13:47 [ET]
Event Date: 03/22/2001
Event Time: 07:07 [EDT]
Last Update Date: 05/17/2001
Notification Time: 13:47 [ET]
Event Date: 03/22/2001
Event Time: 07:07 [EDT]
Last Update Date: 05/17/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
CHARLES R. OGLE (R2)
CHARLES R. OGLE (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 |
This report is being made in accordance with 50.73a)(1), which states, in part, "In the case of an invalid actuation reported under §50.73 (a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER." These invalid actuations are being reported under §50.73 (a)(2)(iv)(A).
On March 22, 2001, at 0707 hours, during the performance of Maintenance Surveillance Test (1MST-DG13R), "DG-3 Loading Test," steps performed out of sequence resulted in invalid actuation of Primary Containment Isolation system (PCIS) Group 6 valves and Secondary Containment Isolation dampers due to interruption of electrical power to the Emergency Bus E3. During a "dry run" of the test, instrumentation and controls technicians inserted the actual test signals before the Unit 1 "CAC Purge Vent Isol Ovrd Switch," 1-CAC-CS-5519, was placed in the "OVERRIDE" position, as required by the procedure. The purpose of placing the switch in the "OVERRIDE" position is to prevent a P015 Group 6 isolation, Secondary Containment isolation, and Standby Gas Treatment (SBGT) system actuations when Emergency Bus E3 is de-energized as part of the test. The actuations of PCIS Group 6 valves and Secondary Containment Isolation dampers were complete and the affected equipment responded as designed to the invalid signal (i.e., the valves and dampers that were open, at the time of the event, closed). At 0715 hours, Emergency Bus E3 was energized by its normal power source. By 0726 hours, the PCIS Group 6 isolation logic was reset, the Reactor Building Ventilation system was returned to service (i.e., the Secondary Containment isolation was reset), and both SBGT trains A and B were placed in standby.
PCIS Group 6 valves include isolation valves from the Containment Atmosphere Control, Containment Atmosphere Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling systems. Discussion of the causes and corrective actions associated with this event are documented in the corrective action program in AR 29889.
The NRC Resident Inspector was notified of this 60-Day report by the licensee.
On March 22, 2001, at 0707 hours, during the performance of Maintenance Surveillance Test (1MST-DG13R), "DG-3 Loading Test," steps performed out of sequence resulted in invalid actuation of Primary Containment Isolation system (PCIS) Group 6 valves and Secondary Containment Isolation dampers due to interruption of electrical power to the Emergency Bus E3. During a "dry run" of the test, instrumentation and controls technicians inserted the actual test signals before the Unit 1 "CAC Purge Vent Isol Ovrd Switch," 1-CAC-CS-5519, was placed in the "OVERRIDE" position, as required by the procedure. The purpose of placing the switch in the "OVERRIDE" position is to prevent a P015 Group 6 isolation, Secondary Containment isolation, and Standby Gas Treatment (SBGT) system actuations when Emergency Bus E3 is de-energized as part of the test. The actuations of PCIS Group 6 valves and Secondary Containment Isolation dampers were complete and the affected equipment responded as designed to the invalid signal (i.e., the valves and dampers that were open, at the time of the event, closed). At 0715 hours, Emergency Bus E3 was energized by its normal power source. By 0726 hours, the PCIS Group 6 isolation logic was reset, the Reactor Building Ventilation system was returned to service (i.e., the Secondary Containment isolation was reset), and both SBGT trains A and B were placed in standby.
PCIS Group 6 valves include isolation valves from the Containment Atmosphere Control, Containment Atmosphere Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling systems. Discussion of the causes and corrective actions associated with this event are documented in the corrective action program in AR 29889.
The NRC Resident Inspector was notified of this 60-Day report by the licensee.