Event Notification Report for January 12, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
01/11/2001 - 01/12/2001
Power Reactor
Event Number: 37788
Facility: SUMMER
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP
NRC Notified By: KELLEY MARSH
HQ OPS Officer: LEIGH TROCINE
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP
NRC Notified By: KELLEY MARSH
HQ OPS Officer: LEIGH TROCINE
Notification Date: 02/28/2001
Notification Time: 14:42 [ET]
Event Date: 01/12/2001
Event Time: 12:08 [EST]
Last Update Date: 02/28/2001
Notification Time: 14:42 [ET]
Event Date: 01/12/2001
Event Time: 12:08 [EST]
Last Update Date: 02/28/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
STEPHEN CAHILL (R2)
VERN HODGE (NRR)
RICHARD CONTE (R1)
ANTON VEGEL (R3)
LINDA HOWELL (R4)
STEPHEN CAHILL (R2)
VERN HODGE (NRR)
RICHARD CONTE (R1)
ANTON VEGEL (R3)
LINDA HOWELL (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Startup |
10 CFR PART 21 REPORT REGARDING A SUBSTANTIAL SAFETY HAZARD WITH A WOODWARD GOVERNOR EGB UNIT
The following text is a portion of a facsimile received from the licensee:
"On 1/12/01, V. C. Summer Nuclear Station (VCSNS) determined that a Substantial Safety Hazard could exist with Woodward governor EGB units failing to properly control the Emergency Diesel Generator [(EDG)]. A condition report (received on 1/12/01) from B. F. Goodrich/Fairbanks Morse indicated there was binding of the terminal shaft internal to an EGB unit removed from the 'A' EDG at VCSNS. The disassembly inspection found that the binding of the terminal shaft was due to the use of clamping screws, which were 1/8 inch longer than specification, resulting in binding on the output servo piston connecting rod.
"This error was made during the prior refurbishment of the EGB by Engine Systems Inc. (ESI), the Woodward Governor nuclear representative, on 2/24/00. The use of the longer screws resulting in binding in the unit is considered a defect, which was introduced during the repair process by ESI.
"[VCSNS] utilizes these EGB units on both the 'A' and 'B' EDGs. The above condition produced severe frequency oscillations on the output of the 'A' EDG. This condition could have been introduced to any EGB during refurbishment by ESI, passed acceptance testing, and be present on both EDGs. In an emergency condition, both EDGs could fail, resulting in the loss of safety function for the emergency on-site power supply.
"The EGB units currently installed on both EDGs have been refurbished by Fairbanks Morse with properly specified parts. Fairbanks Morse will perform all future EGB refurbishment work for VCSNS. Other cause and corrective actions associated with the governor issues are being addressed under Root Cause Analysis report RCA 01-0235, 'Diesel Generator 'A' Governor Problems.' "
The licensee notified the NRC resident inspector.
The following text is a portion of a facsimile received from the licensee:
"On 1/12/01, V. C. Summer Nuclear Station (VCSNS) determined that a Substantial Safety Hazard could exist with Woodward governor EGB units failing to properly control the Emergency Diesel Generator [(EDG)]. A condition report (received on 1/12/01) from B. F. Goodrich/Fairbanks Morse indicated there was binding of the terminal shaft internal to an EGB unit removed from the 'A' EDG at VCSNS. The disassembly inspection found that the binding of the terminal shaft was due to the use of clamping screws, which were 1/8 inch longer than specification, resulting in binding on the output servo piston connecting rod.
"This error was made during the prior refurbishment of the EGB by Engine Systems Inc. (ESI), the Woodward Governor nuclear representative, on 2/24/00. The use of the longer screws resulting in binding in the unit is considered a defect, which was introduced during the repair process by ESI.
"[VCSNS] utilizes these EGB units on both the 'A' and 'B' EDGs. The above condition produced severe frequency oscillations on the output of the 'A' EDG. This condition could have been introduced to any EGB during refurbishment by ESI, passed acceptance testing, and be present on both EDGs. In an emergency condition, both EDGs could fail, resulting in the loss of safety function for the emergency on-site power supply.
"The EGB units currently installed on both EDGs have been refurbished by Fairbanks Morse with properly specified parts. Fairbanks Morse will perform all future EGB refurbishment work for VCSNS. Other cause and corrective actions associated with the governor issues are being addressed under Root Cause Analysis report RCA 01-0235, 'Diesel Generator 'A' Governor Problems.' "
The licensee notified the NRC resident inspector.
Power Reactor
Event Number: 37660
Facility: WATTS BAR
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: TERRY KNUETTEL
HQ OPS Officer: STEVE SANDIN
Region: 2 State: TN
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: TERRY KNUETTEL
HQ OPS Officer: STEVE SANDIN
Notification Date: 01/12/2001
Notification Time: 13:36 [ET]
Event Date: 01/12/2001
Event Time: 10:30 [EST]
Last Update Date: 01/12/2001
Notification Time: 13:36 [ET]
Event Date: 01/12/2001
Event Time: 10:30 [EST]
Last Update Date: 01/12/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
10 CFR Section:
26.73 - FITNESS FOR DUTY
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 |
AN ACTING SUPERVISOR TESTED POSITIVE FOR AN ILLEGAL DRUG DURING A RANDOM TEST
A non-licensed employee was determined to be under the influence of marijuana during a random test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details.
The licensee will inform the NRC resident inspector.
A non-licensed employee was determined to be under the influence of marijuana during a random test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details.
The licensee will inform the NRC resident inspector.
Power Reactor
Event Number: 37661
Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: R BARTON
HQ OPS Officer: JOHN MacKINNON
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: R BARTON
HQ OPS Officer: JOHN MacKINNON
Notification Date: 01/12/2001
Notification Time: 13:40 [ET]
Event Date: 01/12/2001
Event Time: 09:28 [CST]
Last Update Date: 01/12/2001
Notification Time: 13:40 [ET]
Event Date: 01/12/2001
Event Time: 09:28 [CST]
Last Update Date: 01/12/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
DALE POWERS (R4)
DALE POWERS (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
INVALID ACTUATION OF AN ENGINEERED SAFETY FEATURE SYSTEM
During performance of "A" Train Trip Actuations Device Operational Test a reactor protection signal was generated when a procedure step was inadvertently skipped. At the time this occurred the "A" reactor trip bypass breaker was closed when the trip signal was generated. This caused the "A" reactor trip breaker to open. The reactor did not trip because the "A" bypass breaker was closed.
This event resulted in an invalid actuation of an Engineered Safety Feature system. It was concluded that this event was reportable per 10CFR50.72(b)(2)(ii) since:
(1) This actuation was not anticipated to occur as part of this pre-planned evolution since the trip signal was expected to be inhibited by the omitted procedure step and
(2) The omitted step resulted in the system not being properly removed from service as required by the surveillance procedure (source range trip signal not being bypassed or inhibited).
The NRC Resident Inspector was notified of this event by the licensee.
During performance of "A" Train Trip Actuations Device Operational Test a reactor protection signal was generated when a procedure step was inadvertently skipped. At the time this occurred the "A" reactor trip bypass breaker was closed when the trip signal was generated. This caused the "A" reactor trip breaker to open. The reactor did not trip because the "A" bypass breaker was closed.
This event resulted in an invalid actuation of an Engineered Safety Feature system. It was concluded that this event was reportable per 10CFR50.72(b)(2)(ii) since:
(1) This actuation was not anticipated to occur as part of this pre-planned evolution since the trip signal was expected to be inhibited by the omitted procedure step and
(2) The omitted step resulted in the system not being properly removed from service as required by the surveillance procedure (source range trip signal not being bypassed or inhibited).
The NRC Resident Inspector was notified of this event by the licensee.