Event Notification Report for July 14, 2004
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
07/13/2004 - 07/14/2004
EVENT NUMBERS
40873408744086940870
Fuel Cycle Facility
Event Number: 40873
Facility: PADUCAH GASEOUS DIFFUSION PLANT
Region: 2 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: TOM WHITE
HQ OPS Officer: BILL GOTT
Region: 2 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: TOM WHITE
HQ OPS Officer: BILL GOTT
Notification Date: 07/14/2004
Notification Time: 23:45 [ET]
Event Date: 07/14/2004
Event Time: 13:00 [CDT]
Last Update Date: 07/15/2004
Notification Time: 23:45 [ET]
Event Date: 07/14/2004
Event Time: 13:00 [CDT]
Last Update Date: 07/15/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
BRIAN BONSER (R2)
C.W. (BILL) REAMER (NMSS)
BRIAN BONSER (R2)
C.W. (BILL) REAMER (NMSS)
CRITICALITY CONTROL 24-HOUR (BULLETIN 91-01) REPORT
"At 1300 [CDT] on 7/14/04, the Plant Shift Superintendent (PSS) was notified that a cylinder, which was to be washed, was determined to contain heel material. The cylinder had been filled a single time with natural material at Honeywell (2/2001) since its last wash in 8/1999. An examination of the associated Nuclear Material Control and Accountability (NCM&A) data and cylinder card indicated that the only other filling was in 9/1982 with depleted material at C-315. The cylinder was fed to the cascade in 9/2003 at C-337-A.
"A total of 4 gas over solid samples were taken of the cylinder and all indicated that the cylinder contains material. Nuclear Criticality Safety Evaluation 085 and Nuclear Criticality Safety Analysis 400-012 establish double contingency to ensure that a fissile cylinder is not washed in the non-fissile cylinder wash facility.
"The NRC Resident Inspector has been notified of this event.
"PGDP Assessment and Tracking Report No. ATR 04-2756; PGDP Event Report No. PAD 2004-019; Responsible Division: Operations.
"SAFETY SIGNIFICANCE OF EVENTS:
"NMC&A showed this cylinder contained a heel of around 8 lbs. Gas over solid sampling determined this non-fissile cylinder contained fissile material and could not be washed. Although the cylinder was not washed, an unlikely failure occurred that allowed fissile material to enter a cylinder controlled as non-fissile.
"CONTROLLED PARAMETERS:
"The process condition relied upon for double contingency for this scenario is enrichment.
"NUCLEAR CRITICALITY SAFETY CONTROL OR CONTROL SYSTEMS AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
"The first leg of double contingency relies on a gas over solid sample of the cylinder to demonstrate that the cylinder contains non-fissile material prior to washing.
"The second leg of double contingency relies on an argument that back- or cross-feeding at C-337A is unlikely to result in a significant amount of fissile material to be transferred to the cylinder. Since fissile material was found in the cylinder after it was fed at C-337A, this unlikely argument appears to have failed. However, the cylinder was not washed, maintaining this parameter.
"Although the parameter was maintained, double contingency is based on two controls on one parameter and one of the controls (unlikely argument) failed. Therefore double contingency was not maintained.
"CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:
"1. Correct NMC&A database for this cylinder to show the correct assay in the cylinder and that the cylinder is fissile."
"At 1300 [CDT] on 7/14/04, the Plant Shift Superintendent (PSS) was notified that a cylinder, which was to be washed, was determined to contain heel material. The cylinder had been filled a single time with natural material at Honeywell (2/2001) since its last wash in 8/1999. An examination of the associated Nuclear Material Control and Accountability (NCM&A) data and cylinder card indicated that the only other filling was in 9/1982 with depleted material at C-315. The cylinder was fed to the cascade in 9/2003 at C-337-A.
"A total of 4 gas over solid samples were taken of the cylinder and all indicated that the cylinder contains material. Nuclear Criticality Safety Evaluation 085 and Nuclear Criticality Safety Analysis 400-012 establish double contingency to ensure that a fissile cylinder is not washed in the non-fissile cylinder wash facility.
"The NRC Resident Inspector has been notified of this event.
"PGDP Assessment and Tracking Report No. ATR 04-2756; PGDP Event Report No. PAD 2004-019; Responsible Division: Operations.
"SAFETY SIGNIFICANCE OF EVENTS:
"NMC&A showed this cylinder contained a heel of around 8 lbs. Gas over solid sampling determined this non-fissile cylinder contained fissile material and could not be washed. Although the cylinder was not washed, an unlikely failure occurred that allowed fissile material to enter a cylinder controlled as non-fissile.
"CONTROLLED PARAMETERS:
"The process condition relied upon for double contingency for this scenario is enrichment.
"NUCLEAR CRITICALITY SAFETY CONTROL OR CONTROL SYSTEMS AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
"The first leg of double contingency relies on a gas over solid sample of the cylinder to demonstrate that the cylinder contains non-fissile material prior to washing.
"The second leg of double contingency relies on an argument that back- or cross-feeding at C-337A is unlikely to result in a significant amount of fissile material to be transferred to the cylinder. Since fissile material was found in the cylinder after it was fed at C-337A, this unlikely argument appears to have failed. However, the cylinder was not washed, maintaining this parameter.
"Although the parameter was maintained, double contingency is based on two controls on one parameter and one of the controls (unlikely argument) failed. Therefore double contingency was not maintained.
"CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:
"1. Correct NMC&A database for this cylinder to show the correct assay in the cylinder and that the cylinder is fissile."
Power Reactor
Event Number: 40874
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [1] [2] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN HAUTALA
HQ OPS Officer: CHAUNCEY GOULD
Region: 4 State: AZ
Unit: [1] [2] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN HAUTALA
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 07/15/2004
Notification Time: 17:17 [ET]
Event Date: 07/14/2004
Event Time: 16:30 [MST]
Last Update Date: 07/15/2004
Notification Time: 17:17 [ET]
Event Date: 07/14/2004
Event Time: 16:30 [MST]
Last Update Date: 07/15/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
DALE POWERS (R4)
JOHN HANNON (NRR)
DALE POWERS (R4)
JOHN HANNON (NRR)
| 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 | N | 0 | Hot Standby | 0 | Hot Standby |
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
ALL THREE UNITS EXCEEDED THE MAXIMUM POWER LEVEL SPECIFIED IN THEIR OPERATING LICENSE
"Palo Verde Nuclear Generating Station Units 1, 2, and 3 determined that the Maximum Power Level specified in Operating License Condition 2.C(1) was exceeded in the past due to a nonconservative feedwater flow input in the secondary calorimetric calculation since installation of ultrasonic feedwater flow measurement (UFM) instrumentation in Units 1 and 3 in 1999 and 2000 in Unit 2. On 06/09/2004, the vendor (Caldon) notified Palo Verde via a letter that there may be an error of up to one to one and one-half percent in the flow measurement. As a compensatory measure, all three units reduced power and removed the UFM from service. Feedwater flow venturis are used for measurement with the UFM out of service. Engineering evaluation concluded on 07/14/2004 the maximum Palo Verde specific calorimetric error was approximately 38.76 Mw in Units 1 and 3 and 39.90 MW in Unit 2 or approximately 1 %.
The error resulted in core power levels above the Operating License limit of 3876 MW thermal in Units 1 and 3 and the Operating License limit of 3990 MW thermal In Unit 2 while the UFM was in service.
This report is being made because a review of historical operating data revealed that the Maximum Power Level was exceeded.
The Palo Verde safety analyses bound a power level of 102%. Power level did not exceed the safety analysis.
No ESF actuations occurred and none were required as a result of the miscalibration. There were no other structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were Involved. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public."
The NRC Resident Inspector was notified.
"Palo Verde Nuclear Generating Station Units 1, 2, and 3 determined that the Maximum Power Level specified in Operating License Condition 2.C(1) was exceeded in the past due to a nonconservative feedwater flow input in the secondary calorimetric calculation since installation of ultrasonic feedwater flow measurement (UFM) instrumentation in Units 1 and 3 in 1999 and 2000 in Unit 2. On 06/09/2004, the vendor (Caldon) notified Palo Verde via a letter that there may be an error of up to one to one and one-half percent in the flow measurement. As a compensatory measure, all three units reduced power and removed the UFM from service. Feedwater flow venturis are used for measurement with the UFM out of service. Engineering evaluation concluded on 07/14/2004 the maximum Palo Verde specific calorimetric error was approximately 38.76 Mw in Units 1 and 3 and 39.90 MW in Unit 2 or approximately 1 %.
The error resulted in core power levels above the Operating License limit of 3876 MW thermal in Units 1 and 3 and the Operating License limit of 3990 MW thermal In Unit 2 while the UFM was in service.
This report is being made because a review of historical operating data revealed that the Maximum Power Level was exceeded.
The Palo Verde safety analyses bound a power level of 102%. Power level did not exceed the safety analysis.
No ESF actuations occurred and none were required as a result of the miscalibration. There were no other structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were Involved. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public."
The NRC Resident Inspector was notified.
Power Reactor
Event Number: 40869
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: BILL CARSKY
HQ OPS Officer: BILL GOTT
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: BILL CARSKY
HQ OPS Officer: BILL GOTT
Notification Date: 07/14/2004
Notification Time: 03:55 [ET]
Event Date: 07/14/2004
Event Time: 00:45 [CDT]
Last Update Date: 07/14/2004
Notification Time: 03:55 [ET]
Event Date: 07/14/2004
Event Time: 00:45 [CDT]
Last Update Date: 07/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
KENNETH RIEMER (R3)
KENNETH RIEMER (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown |
RPS AND CONTAINMENT ISOLATION ACTUATION DUE TO LOW RPV LEVEL
"On July 14, 2004, at 0045 CDT, the Clinton Power Station experienced a drop in RPV [Reactor Pressure Vessel] level from nominally 30 inches to the Level 3 (the RPS and Isolation setpoint). At this time control rods were fully inserted [see Event Notification - #40868], however, the containment isolation valves associated with this setpoint were open, and these valves automatically closed successfully.
"Level was restored to normal by automatic makeup from the condensate system.
"There were no other malfunctions. The cause for the level drop is being investigated."
The licensee will notify the NRC Resident Inspector.
"On July 14, 2004, at 0045 CDT, the Clinton Power Station experienced a drop in RPV [Reactor Pressure Vessel] level from nominally 30 inches to the Level 3 (the RPS and Isolation setpoint). At this time control rods were fully inserted [see Event Notification - #40868], however, the containment isolation valves associated with this setpoint were open, and these valves automatically closed successfully.
"Level was restored to normal by automatic makeup from the condensate system.
"There were no other malfunctions. The cause for the level drop is being investigated."
The licensee will notify the NRC Resident Inspector.
Power Reactor
Event Number: 40870
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 07/14/2004
Notification Time: 07:35 [ET]
Event Date: 07/14/2004
Event Time: 01:35 [MST]
Last Update Date: 07/14/2004
Notification Time: 07:35 [ET]
Event Date: 07/14/2004
Event Time: 01:35 [MST]
Last Update Date: 07/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
DALE POWERS (R4)
TERRY REIS (NRR)
DALE POWERS (R4)
TERRY REIS (NRR)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
AUTOMATIC REACTOR TRIP FOLLOWING A MAIN GENERATOR TRIP
The following information was received from the licensee via facsimile:
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On July 14, 2004, at approximately 01:35 Mountain Standard Time (MST) Palo Verde Unit 2 experienced a Main Generator Trip immediately followed by an automatic Reactor Trip. The reactor was at approximately 100% power and normal operating temperature and pressure prior to the event. The cause of the Main Generator Trip was most likely the result of electrical storm conditions present at the site at the time of the trip. Unit 2 was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. Heat removal was maintained to the condenser via the steam bypass control system. This was an uncomplicated reactor trip, No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event.
"Unit 2 Is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.
"The [NRC] Senior Resident Inspector was informed of the Unit 2 reactor trip."
No primary or secondary power-operated or manual relief valves lifted as a result of the plant transient. Units 1 and 3 were unaffected by the trip on Unit 2. Offsite power was maintained to Unit 2 safety busses throughout the event.
The following information was received from the licensee via facsimile:
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On July 14, 2004, at approximately 01:35 Mountain Standard Time (MST) Palo Verde Unit 2 experienced a Main Generator Trip immediately followed by an automatic Reactor Trip. The reactor was at approximately 100% power and normal operating temperature and pressure prior to the event. The cause of the Main Generator Trip was most likely the result of electrical storm conditions present at the site at the time of the trip. Unit 2 was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. Heat removal was maintained to the condenser via the steam bypass control system. This was an uncomplicated reactor trip, No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event.
"Unit 2 Is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.
"The [NRC] Senior Resident Inspector was informed of the Unit 2 reactor trip."
No primary or secondary power-operated or manual relief valves lifted as a result of the plant transient. Units 1 and 3 were unaffected by the trip on Unit 2. Offsite power was maintained to Unit 2 safety busses throughout the event.