Event Notification Report for December 10, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
12/09/2001 - 12/10/2001
Fuel Cycle Facility
Event Number: 38553
Facility: PADUCAH GASEOUS DIFFUSION PLANT
Region: 3 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: MATT MAURER
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: KY
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: MATT MAURER
HQ OPS Officer: LEIGH TROCINE
Notification Date: 12/10/2001
Notification Time: 11:53 [ET]
Event Date: 12/10/2001
Event Time: 03:48 [CST]
Last Update Date: 12/10/2001
Notification Time: 11:53 [ET]
Event Date: 12/10/2001
Event Time: 03:48 [CST]
Last Update Date: 12/10/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
76.120(c)(2) - SAFETY EQUIPMENT FAILURE
10 CFR Section:
76.120(c)(2) - SAFETY EQUIPMENT FAILURE
Person (Organization):
ANNE MARIE STONE (R3)
FRED BROWN (NMSS)
ANNE MARIE STONE (R3)
FRED BROWN (NMSS)
INADVERTENT LOSS OF POWER TO THE C-315 TAILS WITHDRAWAL BUILDING DURING AN IMPROPERLY PERFORMED POWER SWITCHING EVOLUTION (24-Hour Report)
The following text is a portion of a facsimile received from the Paducah:
"At 0348 CST on 12-10-01, 480-volt power was inadvertently lost to the C-315 tails withdrawal building during an improperly performed power switching evolution. As a result of the power loss, the high voltage UF6 release detection system for the UF6 condensers, accumulators, and heated housings was rendered inoperable for approximately 10 minutes. The high voltage UF6 release detection system is required to be operable per TSR 2.3.4.4. The function of the high voltage UF6 release detection system is to detect UF6 releases and to sound an alarm in the local control room. The power loss also caused the withdrawal systems to shut down and automatically isolate as designed. No UF6 releases occurred as a result of this evolution. Power was restored and the building was returned to normal operation."
"This event is reportable under 10 CFR 76.120(c)(2) as an event in which equipment required by the TSR is disabled or fails to function as designed."
Paducah personnel notified the NRC resident inspector.
The following text is a portion of a facsimile received from the Paducah:
"At 0348 CST on 12-10-01, 480-volt power was inadvertently lost to the C-315 tails withdrawal building during an improperly performed power switching evolution. As a result of the power loss, the high voltage UF6 release detection system for the UF6 condensers, accumulators, and heated housings was rendered inoperable for approximately 10 minutes. The high voltage UF6 release detection system is required to be operable per TSR 2.3.4.4. The function of the high voltage UF6 release detection system is to detect UF6 releases and to sound an alarm in the local control room. The power loss also caused the withdrawal systems to shut down and automatically isolate as designed. No UF6 releases occurred as a result of this evolution. Power was restored and the building was returned to normal operation."
"This event is reportable under 10 CFR 76.120(c)(2) as an event in which equipment required by the TSR is disabled or fails to function as designed."
Paducah personnel notified the NRC resident inspector.
Power Reactor
Event Number: 38555
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: TOM OSELAND
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: IL
Unit: [1] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: TOM OSELAND
HQ OPS Officer: LEIGH TROCINE
Notification Date: 12/10/2001
Notification Time: 18:40 [ET]
Event Date: 12/10/2001
Event Time: 12:00 [CST]
Last Update Date: 12/10/2001
Notification Time: 18:40 [ET]
Event Date: 12/10/2001
Event Time: 12:00 [CST]
Last Update Date: 12/10/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
ANNE MARIE STONE (R3)
ANNE MARIE STONE (R3)
| 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 |
DISCOVERY THAT STANDBY LIQUID CONTROL SYSTEM RELIEF VALVES MAY INTERMITTENTLY LIFT DURING THE MOST LIMITING ATWS TRANSIENTS (Refer to event #38556 for a similar notification received from Dresden Unit 3.)
The following text is a portion of a facsimile received from the licensee:
"While performing calculations associated with the power uprate project, Quad Cities Nuclear Power Station determined that the Standby Liquid Control (SLC) system relief valves on Unit 1 and Unit 2 may intermittently lift during the most limiting ATWS [transients. (The] specific scenario evaluated is a Main Steam Line isolation at rated power with failure of the normal, backup, and alternate rod insertion (ARI) scram functions.) This issue also applies to the current rated power level during two-pump SLC operation. Title 10 of the Code of Federal Regulations, subpart 50.62, requires the SLC system to inject the equivalent of 86 gpm of 13-weight percent sodium pentaborate solution. While the relief valve was lifted, the system flow rate would not meet the required equivalent flow rate into the vessel. Therefore, this condition is not in direct compliance with 10 CFR 50.62. The NRC communicated a similar industry issue in Information Notice 2001-13."
"Further plant specific analysis has demonstrated that an adequate volume of boron would be injected to shut down the reactor and meet the appropriate acceptance criteria including reactor pressure, peak cladding temperature, peak suppression pool temperature, and peak containment pressure. For these reasons, the safety significance of this event is minimal."
"This phone call is being made as an informational call. Written communications, as appropriate, will be submitted to provide additional detail. The station will continue to communicate additional information with the NRC resident inspectors and with Region Ill as it is identified."
The licensee notified the NRC resident inspector.
The following text is a portion of a facsimile received from the licensee:
"While performing calculations associated with the power uprate project, Quad Cities Nuclear Power Station determined that the Standby Liquid Control (SLC) system relief valves on Unit 1 and Unit 2 may intermittently lift during the most limiting ATWS [transients. (The] specific scenario evaluated is a Main Steam Line isolation at rated power with failure of the normal, backup, and alternate rod insertion (ARI) scram functions.) This issue also applies to the current rated power level during two-pump SLC operation. Title 10 of the Code of Federal Regulations, subpart 50.62, requires the SLC system to inject the equivalent of 86 gpm of 13-weight percent sodium pentaborate solution. While the relief valve was lifted, the system flow rate would not meet the required equivalent flow rate into the vessel. Therefore, this condition is not in direct compliance with 10 CFR 50.62. The NRC communicated a similar industry issue in Information Notice 2001-13."
"Further plant specific analysis has demonstrated that an adequate volume of boron would be injected to shut down the reactor and meet the appropriate acceptance criteria including reactor pressure, peak cladding temperature, peak suppression pool temperature, and peak containment pressure. For these reasons, the safety significance of this event is minimal."
"This phone call is being made as an informational call. Written communications, as appropriate, will be submitted to provide additional detail. The station will continue to communicate additional information with the NRC resident inspectors and with Region Ill as it is identified."
The licensee notified the NRC resident inspector.
Power Reactor
Event Number: 38556
Facility: DRESDEN
Region: 3 State: IL
Unit: [] [] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: BOB SCOTT
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: IL
Unit: [] [] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: BOB SCOTT
HQ OPS Officer: LEIGH TROCINE
Notification Date: 12/10/2001
Notification Time: 18:49 [ET]
Event Date: 12/10/2001
Event Time: 12:00 [CST]
Last Update Date: 12/10/2001
Notification Time: 18:49 [ET]
Event Date: 12/10/2001
Event Time: 12:00 [CST]
Last Update Date: 12/10/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
ANNE MARIE STONE (R3)
ANNE MARIE STONE (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
DISCOVERY THAT STANDBY LIQUID CONTROL SYSTEM RELIEF VALVES ON UNIT 3 MAY INTERMITTENTLY LIFT DURING THE MOST LIMITING ATWS TRANSIENTS (Refer to event #38555 for a similar notification received from Quad Cities.)
The following text is a portion of a facsimile received from the licensee:
"While performing calculations associated with the power uprate project, Dresden Nuclear Power Station determined that the Standby Liquid Control (SLC) system relief valves on Unit 3 may intermittently lift during the most limiting ATWS [transients. (The] specific scenario evaluated is a Main Steam Line isolation at rated power with failure of the normal, backup, and alternate rod insertion (ARI) scram functions.) This issue also applies to the current rated power level during two pump SLC operation. Title 10 of the Code of Federal Regulations, subpart 50.62, requires the SLC system to inject the equivalent of 86 gpm of 13-weight percent sodium pentaborate solution. While the relief valve was lifted, the system flow rate would not meet the required equivalent flow rate into the vessel. Therefore, this condition is not in direct compliance with 10 CFR 50.62. The NRC communicated a similar industry issue in Information Notice 2001-13"
"Further plant specific analysis has demonstrated that an adequate volume of boron would be injected to shut down the reactor and meet the appropriate acceptance criteria including reactor pressure, peak cladding temperature, peak suppression pool temperature, and peak containment pressure. For these reasons, the safety significance of this event is minimal."
"This phone call is being made as an informational call. Written communications, as appropriate, will be submitted to provide additional detail. The station will continue to communicate additional information with the NRC resident inspectors and with [NRC] Region Ill as it is identified."
The licensee stated that although Dresden Units 2 and 3 are similar vintage boiling water reactors, this issue does not apply to Dresden Unit 2 because the piping runs are different.
The licensee notified the NRC resident inspector.
The following text is a portion of a facsimile received from the licensee:
"While performing calculations associated with the power uprate project, Dresden Nuclear Power Station determined that the Standby Liquid Control (SLC) system relief valves on Unit 3 may intermittently lift during the most limiting ATWS [transients. (The] specific scenario evaluated is a Main Steam Line isolation at rated power with failure of the normal, backup, and alternate rod insertion (ARI) scram functions.) This issue also applies to the current rated power level during two pump SLC operation. Title 10 of the Code of Federal Regulations, subpart 50.62, requires the SLC system to inject the equivalent of 86 gpm of 13-weight percent sodium pentaborate solution. While the relief valve was lifted, the system flow rate would not meet the required equivalent flow rate into the vessel. Therefore, this condition is not in direct compliance with 10 CFR 50.62. The NRC communicated a similar industry issue in Information Notice 2001-13"
"Further plant specific analysis has demonstrated that an adequate volume of boron would be injected to shut down the reactor and meet the appropriate acceptance criteria including reactor pressure, peak cladding temperature, peak suppression pool temperature, and peak containment pressure. For these reasons, the safety significance of this event is minimal."
"This phone call is being made as an informational call. Written communications, as appropriate, will be submitted to provide additional detail. The station will continue to communicate additional information with the NRC resident inspectors and with [NRC] Region Ill as it is identified."
The licensee stated that although Dresden Units 2 and 3 are similar vintage boiling water reactors, this issue does not apply to Dresden Unit 2 because the piping runs are different.
The licensee notified the NRC resident inspector.