Event Notification Report for August 21, 2009
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
08/20/2009 - 08/21/2009
Power Reactor
Event Number: 45281
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: DAVE DEES
HQ OPS Officer: CHARLES TEAL
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: DAVE DEES
HQ OPS Officer: CHARLES TEAL
Notification Date: 08/21/2009
Notification Time: 11:22 [ET]
Event Date: 08/21/2009
Event Time: 06:48 [CDT]
Last Update Date: 08/21/2009
Notification Time: 11:22 [ET]
Event Date: 08/21/2009
Event Time: 06:48 [CDT]
Last Update Date: 08/21/2009
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):
WILLIAM JONES (R4DO)
WILLIAM JONES (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Hot Standby | 0 | Hot Standby |
FEED WATER ISOLATION DUE TO HIGH STEAM GENERATOR WATER LEVEL
"Turbine trip and Feed [Water] Isolation Signal generated as result of high level in 'A' Steam Generator. At approximately 0648 CDT a turbine trip and Feed Water Isolation Signal was generated when the level in the 'A' Steam Generator exceeded the initiation set point of 78%. Steam Generator level returned below the initiation setpoint at approximately 0656 CDT.
"The turbine was already in the tripped condition and Feed Water Isolation Valves closed due to the plant trip that occurred at 1549 CDT on 8/19/2009 (Reference EN# 45278).
"Steam Generators are being fed with the motor driven Auxiliary Feed Water Pumps. Decay heat removal is via the Steam Generator Atmospheric Relief Valves. The plant is in Mode 3 with Reactor Coolant System pressure at approximately 2235 psig and temperature at approximately 561 F.
"The licensee has notified the NRC Resident Inspector."
"Turbine trip and Feed [Water] Isolation Signal generated as result of high level in 'A' Steam Generator. At approximately 0648 CDT a turbine trip and Feed Water Isolation Signal was generated when the level in the 'A' Steam Generator exceeded the initiation set point of 78%. Steam Generator level returned below the initiation setpoint at approximately 0656 CDT.
"The turbine was already in the tripped condition and Feed Water Isolation Valves closed due to the plant trip that occurred at 1549 CDT on 8/19/2009 (Reference EN# 45278).
"Steam Generators are being fed with the motor driven Auxiliary Feed Water Pumps. Decay heat removal is via the Steam Generator Atmospheric Relief Valves. The plant is in Mode 3 with Reactor Coolant System pressure at approximately 2235 psig and temperature at approximately 561 F.
"The licensee has notified the NRC Resident Inspector."
Power Reactor
Event Number: 45282
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [2] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: WILLIAM BAKER
HQ OPS Officer: JOHN KNOKE
Region: 2 State: AL
Unit: [1] [2] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: WILLIAM BAKER
HQ OPS Officer: JOHN KNOKE
Notification Date: 08/21/2009
Notification Time: 15:02 [ET]
Event Date: 08/21/2009
Event Time: 09:52 [CDT]
Last Update Date: 08/21/2009
Notification Time: 15:02 [ET]
Event Date: 08/21/2009
Event Time: 09:52 [CDT]
Last Update Date: 08/21/2009
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
Person (Organization):
SCOTT SHAEFFER (R2DO)
SCOTT SHAEFFER (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 |
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
LOGIC FOR COMMON ACCIDENT SIGNAL IS FLAWED FOR SAFE SHUTDOWN INSTRUCTION
"Safe Shutdown Instruction (SSI) 0-SSI-9 for an Appendix R fire event requires for a Unit 1 Safe Shutdown success path that 4kV Shutdown Board A be supplied by Diesel Generator 3A via 4kV Shutdown Board 3EA. During a Unit 2 simulator exercise utilizing this SSI, reactor vessel water level lowered to Level 1 (-122 inches) causing a Common Accident Signal (CAS) trip of 4 Kv Shutdown Board 3EA inter-tie breaker (1844) that de-energized 4kV Shutdown Board A. This condition will not support the Unit 1 success path for alternate decay heat removal during an Appendix R fire. The 4 Kv Shutdown Board A must be re-energized by closing breaker 1844. For this SSI (0-SSI-9) the subsequent automatic start of the 2A RHR and Core Spray pumps in response to the CAS logic has not been considered in the development of the SSIs. In response to this condition, compensatory measures have been implemented to provide equivalent shutdown capability, documented on Fire Protection Impairment Permit (FPIP) 09-1920.
"This event is reportable within 8 hours under 10CFR50.72 (b)(3)(v)(A) - Any event or condition that results in the condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition. This event is also a reportable within 60 days under 10CFR50.73(a)(2)(v)(A) - Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition.
"This condition has been documented in the BFN Corrective Action program as PER# 177130. Unit 1, 2 and 3 remain at 100% power and are not affected by this event. The NRC Resident Inspector has been notified."
"Safe Shutdown Instruction (SSI) 0-SSI-9 for an Appendix R fire event requires for a Unit 1 Safe Shutdown success path that 4kV Shutdown Board A be supplied by Diesel Generator 3A via 4kV Shutdown Board 3EA. During a Unit 2 simulator exercise utilizing this SSI, reactor vessel water level lowered to Level 1 (-122 inches) causing a Common Accident Signal (CAS) trip of 4 Kv Shutdown Board 3EA inter-tie breaker (1844) that de-energized 4kV Shutdown Board A. This condition will not support the Unit 1 success path for alternate decay heat removal during an Appendix R fire. The 4 Kv Shutdown Board A must be re-energized by closing breaker 1844. For this SSI (0-SSI-9) the subsequent automatic start of the 2A RHR and Core Spray pumps in response to the CAS logic has not been considered in the development of the SSIs. In response to this condition, compensatory measures have been implemented to provide equivalent shutdown capability, documented on Fire Protection Impairment Permit (FPIP) 09-1920.
"This event is reportable within 8 hours under 10CFR50.72 (b)(3)(v)(A) - Any event or condition that results in the condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition. This event is also a reportable within 60 days under 10CFR50.73(a)(2)(v)(A) - Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition.
"This condition has been documented in the BFN Corrective Action program as PER# 177130. Unit 1, 2 and 3 remain at 100% power and are not affected by this event. The NRC Resident Inspector has been notified."