Event Notification Report for May 14, 2004
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/13/2004 - 05/14/2004
EVENT NUMBERS
4075840752407484074940750
General Information or Other
Event Number: 40758
Rep Org: FLORIDA BUREAU OF RADIATION CONTROL
Licensee: MT. SINAI MEDICAL CENTER OF GREATER MIAMI, INC.
Region: 1
City: MIAMI BEACH State: FL
County: MIAMI-DADE
License #: 64-12
Agreement: Y
Docket:
NRC Notified By: CHARLES ADAMS
HQ OPS Officer: HOWIE CROUCH
Licensee: MT. SINAI MEDICAL CENTER OF GREATER MIAMI, INC.
Region: 1
City: MIAMI BEACH State: FL
County: MIAMI-DADE
License #: 64-12
Agreement: Y
Docket:
NRC Notified By: CHARLES ADAMS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 05/18/2004
Notification Time: 16:31 [ET]
Event Date: 05/14/2004
Event Time: 00:00 [EDT]
Last Update Date: 05/18/2004
Notification Time: 16:31 [ET]
Event Date: 05/14/2004
Event Time: 00:00 [EDT]
Last Update Date: 05/18/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
JOHN CARUSO (R1)
CHARLES MILLER (NMSS)
JOHN CARUSO (R1)
CHARLES MILLER (NMSS)
FLORIDA AGREEMENT STATE REPORT - MEDICAL EVENT
The following information was received from the State of Florida Bureau of Radiation Control via e-mail:
"Patient was scheduled to receive a 2 fraction treatment to the breast over a five day period - for a total ten fractions. Prior to the last two fractions the doctor went out of town. The new doctor discovered that the calculated center of the balloon was off by 4 mm. The doctor chose to cancel the ninth & tenth fractions. Subsequent calculations showed the dose received was approximately 19 [to] 35% less than prescribed, depending on axis used relative to the catheter. The patient has been notified. This office was notified the day the last two fractions were canceled on May 14. However, review of events and the dose calculations were not complete until May 18. Licensee will send a written report on this event. Examination of this report will determine future Florida actions."
The fraction treatments were delivered using a Mammosite with an Ir-192, 6 Curie source. The maximum dose the patient received was calculated to be 2720 centiGray.
The following information was received from the State of Florida Bureau of Radiation Control via e-mail:
"Patient was scheduled to receive a 2 fraction treatment to the breast over a five day period - for a total ten fractions. Prior to the last two fractions the doctor went out of town. The new doctor discovered that the calculated center of the balloon was off by 4 mm. The doctor chose to cancel the ninth & tenth fractions. Subsequent calculations showed the dose received was approximately 19 [to] 35% less than prescribed, depending on axis used relative to the catheter. The patient has been notified. This office was notified the day the last two fractions were canceled on May 14. However, review of events and the dose calculations were not complete until May 18. Licensee will send a written report on this event. Examination of this report will determine future Florida actions."
The fraction treatments were delivered using a Mammosite with an Ir-192, 6 Curie source. The maximum dose the patient received was calculated to be 2720 centiGray.
Power Reactor
Event Number: 40752
Facility: TURKEY POINT
Region: 2 State: FL
Unit: [] [4] []
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: GRANT MELIN
HQ OPS Officer: JEFF ROTTON
Region: 2 State: FL
Unit: [] [4] []
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: GRANT MELIN
HQ OPS Officer: JEFF ROTTON
Notification Date: 05/14/2004
Notification Time: 18:26 [ET]
Event Date: 05/14/2004
Event Time: 17:29 [EDT]
Last Update Date: 05/14/2004
Notification Time: 18:26 [ET]
Event Date: 05/14/2004
Event Time: 17:29 [EDT]
Last Update Date: 05/14/2004
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
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
KERRY LANDIS (R2)
KERRY LANDIS (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 4 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
SCRAM DUE TO LOWERING STEAM GENERATOR LEVEL
At 1728 EDTon 05/14/04, while performing maintenance on Turbine 1st stage steam pressure transmitter (PT-447), feedwater regulating valve (FRV) 4A was observed to close. At 1729, an automatic actuation of RPS occurred on Low Steam Generator level in the 4A Steam Generator. This was followed shortly by a manual reactor trip by the operator. All control rods fully inserted. All 3 turbine driven AFW pumps started as expected and are supply steam generators through both feedwater supply headers. The electric plant remains in a normal lineup. RCPs are in operation transferring decay heat to the steam generators. The MSIVs are open with the steam generators discharging steam to the main condenser using the condenser steam dump valves. All steam generator atmospheric dump valves opened on the reactor trip and closed.
The plant is investigating the cause of the FRV closure and intends to stay in Mode 3 until the cause is determined and appropriate repairs are made. There was no impact on the other operating unit.
The licensee will notify the NRC Resident Inspector.
At 1728 EDTon 05/14/04, while performing maintenance on Turbine 1st stage steam pressure transmitter (PT-447), feedwater regulating valve (FRV) 4A was observed to close. At 1729, an automatic actuation of RPS occurred on Low Steam Generator level in the 4A Steam Generator. This was followed shortly by a manual reactor trip by the operator. All control rods fully inserted. All 3 turbine driven AFW pumps started as expected and are supply steam generators through both feedwater supply headers. The electric plant remains in a normal lineup. RCPs are in operation transferring decay heat to the steam generators. The MSIVs are open with the steam generators discharging steam to the main condenser using the condenser steam dump valves. All steam generator atmospheric dump valves opened on the reactor trip and closed.
The plant is investigating the cause of the FRV closure and intends to stay in Mode 3 until the cause is determined and appropriate repairs are made. There was no impact on the other operating unit.
The licensee will notify the NRC Resident Inspector.
Power Reactor
Event Number: 40748
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: FRANK CIGANIK
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: FRANK CIGANIK
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 05/14/2004
Notification Time: 10:50 [ET]
Event Date: 05/14/2004
Event Time: 08:13 [EDT]
Last Update Date: 05/14/2004
Notification Time: 10:50 [ET]
Event Date: 05/14/2004
Event Time: 08:13 [EDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
JOHN ROGGE (R1)
JOHN ROGGE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
CHANGE IN THE METHODOLOGY USED BY GE/GNF TO DEMONSTRATE COMPLIANCE WITH ECCS PERFORMANCE REQUIREMENTS
"Oyster Creek has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation. Using these results, the increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2) by 1.23%. Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.
"However, sufficient margin exists under current operation to ensure the maximum local oxidation limit would not be exceeded. All other 10CFR50.46 ECCS criteria are not impacted. Additionally, based on a conservative evaluation using nominal inputs and assumptions significant margin to 10CFR50.48 ECCS criteria exists even under unlimited oxygen conditions.
"This voluntary notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.
Background
"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.
"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants including Oyster Creek (BWR/2). The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.
"Based on current core thermal power level and existing margin to limits on power operation there is sufficient margin for analyzed accident scenarios requiring ECCS operation including appropriate compensation to restore the analytical oxidation within 10CFR50.46 criteria and there is no impact on safe operation. This is further supported by the results of conservative evaluation using nominal inputs. As a result of this change in the ECCS model and the consequential adjustment to the allowable MAPLGHR limit, the offsite dose for this scenario is still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.
Corrective Action(s):
"An administrative adjustment to our allowable MAPLHGR limits will be imposed to compensate for the estimated change in cladding oxidation and thereby restore the previously analyzed margin on core thermal power limits to within the new ECCS model prediction. Additionally deinerting of the Primary Containment will be constrained to less than 25% rated power-level until more specific analyses are complete.
"In accordance with 10CFR50.73, a voluntary LER will be submitted within 60 days with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation.
"The licensee will notify the NRC Resident Inspector."
"Oyster Creek has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation. Using these results, the increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2) by 1.23%. Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.
"However, sufficient margin exists under current operation to ensure the maximum local oxidation limit would not be exceeded. All other 10CFR50.46 ECCS criteria are not impacted. Additionally, based on a conservative evaluation using nominal inputs and assumptions significant margin to 10CFR50.48 ECCS criteria exists even under unlimited oxygen conditions.
"This voluntary notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.
Background
"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.
"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants including Oyster Creek (BWR/2). The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.
"Based on current core thermal power level and existing margin to limits on power operation there is sufficient margin for analyzed accident scenarios requiring ECCS operation including appropriate compensation to restore the analytical oxidation within 10CFR50.46 criteria and there is no impact on safe operation. This is further supported by the results of conservative evaluation using nominal inputs. As a result of this change in the ECCS model and the consequential adjustment to the allowable MAPLGHR limit, the offsite dose for this scenario is still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.
Corrective Action(s):
"An administrative adjustment to our allowable MAPLHGR limits will be imposed to compensate for the estimated change in cladding oxidation and thereby restore the previously analyzed margin on core thermal power limits to within the new ECCS model prediction. Additionally deinerting of the Primary Containment will be constrained to less than 25% rated power-level until more specific analyses are complete.
"In accordance with 10CFR50.73, a voluntary LER will be submitted within 60 days with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation.
"The licensee will notify the NRC Resident Inspector."
Power Reactor
Event Number: 40749
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [1] [] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: QUENTIN HICKS
HQ OPS Officer: BILL GOTT
Region: 1 State: NY
Unit: [1] [] []
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: QUENTIN HICKS
HQ OPS Officer: BILL GOTT
Notification Date: 05/14/2004
Notification Time: 15:45 [ET]
Event Date: 05/14/2004
Event Time: 09:17 [EDT]
Last Update Date: 05/14/2004
Notification Time: 15:45 [ET]
Event Date: 05/14/2004
Event Time: 09:17 [EDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
JOHN ROGGE (R1)
JOHN ROGGE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
CHANGE IN THE METHODOLOGY USED BY GE/GNF TO DEMONSTRATE COMPLIANCE WITH ECCS PERFORMANCE REQUIREMENTS
"Nine Mile Point has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation during five reactor recirculation loop operation, with containment inerted. The increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2). Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.
"Notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.
Background
"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.
"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants. The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.
"The change in the allowable MAPLGHR limit for five loop operation with the containment inerted results in the offsite dose for this scenario being still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.
Corrective Action(s):
"Reactor Engineering has implemented the 2% MAPLHGR Limit reduction through imposition of a .98 MAPRAT to restore compliance to the 10CFR50.46 limits.
"Operations will place administrative controls/ procedure changes to preclude operating without containment inerted at or above 25% power to limit oxygen available for this postulated phenomenon.
"In accordance with 10CFR50.73, a LER will be submitted with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation."
The licensee notified the NRC Resident Inspector.
"Nine Mile Point has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation during five reactor recirculation loop operation, with containment inerted. The increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2). Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.
"Notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.
Background
"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.
"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants. The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.
"The change in the allowable MAPLGHR limit for five loop operation with the containment inerted results in the offsite dose for this scenario being still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.
Corrective Action(s):
"Reactor Engineering has implemented the 2% MAPLHGR Limit reduction through imposition of a .98 MAPRAT to restore compliance to the 10CFR50.46 limits.
"Operations will place administrative controls/ procedure changes to preclude operating without containment inerted at or above 25% power to limit oxygen available for this postulated phenomenon.
"In accordance with 10CFR50.73, a LER will be submitted with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation."
The licensee notified the NRC Resident Inspector.
Fuel Cycle Facility
Event Number: 40750
Facility: NUCLEAR FUEL SERVICES INC.
Region: 2 State: TN
Unit: [] [] []
RX Type: URANIUM FUEL FABRICATION
Comments: HEU CONVERSION & SCRAP RECOVERY
NAVAL REACTOR FUEL CYCLE
LEU SCRAP RECOVERY
NRC Notified By: RANDY SHACKELFORD
HQ OPS Officer: JEFF ROTTON
Region: 2 State: TN
Unit: [] [] []
RX Type: URANIUM FUEL FABRICATION
Comments: HEU CONVERSION & SCRAP RECOVERY
NAVAL REACTOR FUEL CYCLE
LEU SCRAP RECOVERY
NRC Notified By: RANDY SHACKELFORD
HQ OPS Officer: JEFF ROTTON
Notification Date: 05/14/2004
Notification Time: 17:07 [ET]
Event Date: 05/14/2004
Event Time: 09:00 [EDT]
Last Update Date: 05/11/2008
Notification Time: 17:07 [ET]
Event Date: 05/14/2004
Event Time: 09:00 [EDT]
Last Update Date: 05/11/2008
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
KERRY LANDIS (R2)
ROBERT PIERSON (NMSS)
KERRY LANDIS (R2)
ROBERT PIERSON (NMSS)
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
THIS IS NOT A NEW REPORT.
This event report was originally withheld from public release under the NRC's policy for protecting sensitive unclassified information. The NRC has reevaluated this policy and is now making this event report available to the public with suitable redactions.
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
91-01 BULLETIN - LOSS OF CRITICALITY SAFETY CONTROLS
"An Integrated Safety Analysis (ISA) review being performed for a system modification identified a previously unidentified failure mode for a piece of Safety Related Equipment (SRE). The SRE item (conductivity probe) is in place to detect and prevent the transfer of moderating materials to a fissile material storage column. The conductivity probe was subsequently tested and found to not meet the currently required performance criteria (i.e. the conductivity probe was not capable of detecting certain moderating materials that could be present in the fissile material). The operation was already in a shutdown state and all fissile material had been removed from the system when the test was performed on the probe.
"Safety significance of the event: The safety significance is very low since the system was currently shutdown and the identified failure mode has not occurred since startup of the operation in 1999.
"Brief scenario(s) of how criticality could occur: Failure of the piece of SRE could only result in a criticality if a subsequent unlikely event were to also occur resulting in a large spill of fissile material.
"Controlled parameters: Moderation and Geometry
"Estimated amount, enrichment, form of licensed material (include process limit and % worst case critical mass): Uranium compounds in the column would be 43 times minimum critical mass if spill of materials in full column occurred. Process limit is limited by geometry of equipment
"Control(s) or control system(s) and the failure(s) or deficiencies: A fissile material transfer line is monitored (for moderation) such that the transfer of moderating materials into a column is prevented. The monitoring equipment was found to not be compatible for all moderating materials in the system preventing it from halting the transfer of moderating materials into the equipment under certain postulated upset conditions.
"Corrective actions taken and when was each was implemented: The conductivity probe was tested in the newly identified failure mode and was found to not respond adequately. The system was already in a shutdown state and all fissile material had been removed from the system when the test was performed on the probe. Compensatory measures (both administrative and engineered) were drafted and reviewed by the plant safety review committee. The system remains shutdown until all compensatory measures are approved and in place."
Licensee notified the NRC Resident Inspector.
* * * UPDATE ON 05/11/08 BY J KOZAL * * *
THIS IS NOT A NEW REPORT.
This event report was originally withheld from public release under the NRC's policy for protecting sensitive unclassified information. The NRC has reevaluated this policy and is now making this event report available to the public with suitable redactions.
THIS IS NOT A NEW REPORT.
This event report was originally withheld from public release under the NRC's policy for protecting sensitive unclassified information. The NRC has reevaluated this policy and is now making this event report available to the public with suitable redactions.
* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
91-01 BULLETIN - LOSS OF CRITICALITY SAFETY CONTROLS
"An Integrated Safety Analysis (ISA) review being performed for a system modification identified a previously unidentified failure mode for a piece of Safety Related Equipment (SRE). The SRE item (conductivity probe) is in place to detect and prevent the transfer of moderating materials to a fissile material storage column. The conductivity probe was subsequently tested and found to not meet the currently required performance criteria (i.e. the conductivity probe was not capable of detecting certain moderating materials that could be present in the fissile material). The operation was already in a shutdown state and all fissile material had been removed from the system when the test was performed on the probe.
"Safety significance of the event: The safety significance is very low since the system was currently shutdown and the identified failure mode has not occurred since startup of the operation in 1999.
"Brief scenario(s) of how criticality could occur: Failure of the piece of SRE could only result in a criticality if a subsequent unlikely event were to also occur resulting in a large spill of fissile material.
"Controlled parameters: Moderation and Geometry
"Estimated amount, enrichment, form of licensed material (include process limit and % worst case critical mass): Uranium compounds in the column would be 43 times minimum critical mass if spill of materials in full column occurred. Process limit is limited by geometry of equipment
"Control(s) or control system(s) and the failure(s) or deficiencies: A fissile material transfer line is monitored (for moderation) such that the transfer of moderating materials into a column is prevented. The monitoring equipment was found to not be compatible for all moderating materials in the system preventing it from halting the transfer of moderating materials into the equipment under certain postulated upset conditions.
"Corrective actions taken and when was each was implemented: The conductivity probe was tested in the newly identified failure mode and was found to not respond adequately. The system was already in a shutdown state and all fissile material had been removed from the system when the test was performed on the probe. Compensatory measures (both administrative and engineered) were drafted and reviewed by the plant safety review committee. The system remains shutdown until all compensatory measures are approved and in place."
Licensee notified the NRC Resident Inspector.
* * * UPDATE ON 05/11/08 BY J KOZAL * * *
THIS IS NOT A NEW REPORT.
This event report was originally withheld from public release under the NRC's policy for protecting sensitive unclassified information. The NRC has reevaluated this policy and is now making this event report available to the public with suitable redactions.