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Event Notification Report for November 3, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
11/02/2006 - 11/03/2006

** EVENT NUMBERS **


42955 42957 42958 42959 42960

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Power Reactor Event Number: 42955
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: SHAWN ZANDER
HQ OPS Officer: STEVE SANDIN
Notification Date: 11/01/2006
Notification Time: 19:10 [ET]
Event Date: 11/01/2006
Event Time: 18:37 [EST]
Last Update Date: 11/02/2006
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(2)(iv)(A) - ECCS INJECTION
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
CAROLYN EVANS (R2)
STUART RICHARDS (NRR)
MELVYN LEACH (IRD)
TOM BARNES (DHS)
TODD KUZIA (FEMA)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 M/R Y 100 Power Operation 0 Hot Shutdown

Event Text

UNIT 2 DECLARED AN UNUSUAL EVENT DUE TO LOSS OF OFFSITE POWER TO THE 4KV EMERGENCY BUSES

At 1823 EST, Unit 2 was manually scrammed due to a loss of offsite power from the Startup Auxiliary Transformer to both 4KV Emergency (E) buses. Both Emergency Diesel Generators (EDGs) 3&4 autostarted and re-energized the affected electrical buses. At 1823 EST, an Unusual Event was declared based on EAL 06.01.01, "Inability to power either 4KV E bus from offsite power."

Unit 2 is currently stable in mode 3, Hot Shutdown, with MSIVs closed and HPCI controlling pressure and RPV Water Level. All control rods fully inserted following the manual reactor scram. The licensee determined that no emergency facilities will be activated and that no offsite assistance is needed at this time.

The licensee informed both state and local agencies and will inform the NRC Resident Inspector.

* * * UPDATE PROVIDED BY JOEL LEVINER TO JEFF ROTTON AT 2214 EST ON 11/01/06 * * *

"On 11/01/06 at approximately 18:23 (EST) Brunswick Unit 2 experienced a loss of the unit's Startup Auxiliary Transformer and a loss of reactor forced circulation. A manual reactor scram was performed as required by station Abnormal Operating Procedures. Due to the loss of the Startup Auxiliary Transformer and subsequent manual reactor scram, a loss of Offsite Power resulted to the unit's power buses when unit shutdown was completed. All control rods properly inserted when the manual reactor scram was performed. All four site emergency diesels started and diesels 3 and 4 are supplying the Unit 2 emergency buses. Reactor water level reached low level 1 (LL1) and low level 2 (LL2) as result of the reactor scram and loss of offsite power. The LL1 signal resulted in Group 2 (floor and equipment drain isolation valves), Group 6 (monitoring and sampling isolation valves) and Group 8 (shutdown cooling isolation valves) isolation signals. All low level 1 isolations occurred as designed. The LL2 resulted in a Reactor Core Isolation Cooling (RCIC) system actuation, High Pressure Coolant Injection (HPCI) system actuation, Group 3 (reactor water cleanup valves) isolation signal, a secondary containment isolation signal, a Standby Gas Treatment (SBGT) initiation signal, a Control Room Emergency Ventilation (CREV) initiation signal, and an Alternate Rod Insertion (ARI) actuation signal. All isolation and actuations occurred as designed with the exception the CREV initiation and ARI actuation. CREV initiation and ARI actuations were performed by manual actions. The failure of the CREV and ARI initiation/actuations are under investigation. The RCIC and HPCI systems were used to restore reactor water level to the normal operation band. Reactor vessel pressure is being controlled in the normal band with manual operation of Safety Relief Valves (SRV), and HPCI/RCIC in pressure control mode. The Main Steam Isolation Valves (MSIVs) (Group 1) and the drywell pneumatic isolation valves (Group 10) closed on the loss of power. The plant is a stable condition. Troubleshooting activities are in progress to determine the cause of the event.

"At 1910, the NRC was previously notified of the Unusual Event declaration.

"Initial Safety Significance Evaluation: The safety significance of this event is minimal and Unit 2 is in a stable condition. All control rods properly inserted when the manual scram was performed. Plant safety systems responded as required with the exception of the CREV and ARI systems which did not automatically initiate but functioned properly when manually actuated. All four emergency diesels started and Unit 2 diesels 3 and 4 are supplying the Unit 2 emergency buses. Reactor pressure and level are being controlled per procedure, with HPCI and RCIC. Actions are in progress to re-establish off site power supply to emergency buses 3 and 4 via backfeed through the Unit Auxiliary Transformer (UAT).

"Corrective Actions: Actions are in progress to re-establish offsite power supply to emergency buses 3 and 4 via backfeed through the UAT. Investigations are in progress to determine the cause of the SAT failure and the failure of CREV and ARI to auto-initiate."

The licensee has notified the NRC Resident Inspector and the State and local emergency agencies.

Update provided also added the following reportable notifications due to the event: 10CFR50.72(b)(2) (iv)(A) and(iv)(B) and 10CFR50.72(b)(3)(iv)(A). Notified R2DO (Evans).

* * * UPDATE PROVIDED BY MARK SCHALL TO JEFF ROTTON AT 1805 EST ON 11/02/06 * * *

Licensee reported that the Unusual Event was terminated at 1745 EST on 11/02/06 after Offsite power was restored to both 4 KV E Buses from the Unit Auxiliary Transformer (UAT) on Unit 2. The #3 and #4 EDGs have been secured and are in Standby. #1 EDG remains inoperable and #2 EDG is presently being Load Tested.

The licensee will be notifying the NRC Resident Inspector and the State and local emergency agencies.

Notified R2DO (Evans), NRREO (Richards), IRD Manager (Leach), DHS (Barnes), and FEMA (Kuzia).

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Power Reactor Event Number: 42957
Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: JOE SHOFFNEN
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/02/2006
Notification Time: 14:53 [ET]
Event Date: 11/02/2006
Event Time: 13:34 [EST]
Last Update Date: 11/02/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
JOHN WHITE (R1)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 100 Power Operation 0 Hot Shutdown

Event Text

AUTOMATIC TURBINE TRIP/ REACTOR TRIP DUE TO INVALID LOW CONDENSER VACUUM SIGNAL


"At 1334 on 11-2-06 an Automatic Reactor Trip occurred from 100% power. All systems functioned as required. One safety valve stuck open on both OTSGs. They subsequently re-seated. An employee working on the roof at the time of the trip fell off a ladder and injured his leg. Emergency medical was contacted to assisted with the injured worker. Two fire trucks and an ambulance was dispatched to the site to remove the injured worker. The worker was not contaminated. There is no indication of any OTSG Tube leaks. Initial investigation indicates the reactor tripped, due to a turbine trip due to an invalid low vacuum signal."

State and local officials will be notified of this event by the licensee.


I&C Techs were performing maintenance on one of the low pressure vacuum switches. An electrical fault fed to the other two low pressure vacuum switches causing a 2 out of 3 signal which resulted in a turbine trip followed by a reactor trip signal, as expected. All rods fully inserted into the core. One safety valve (9 safety valves on each OTSG) on each Once Through Steam Generator stuck open. OTSG "B" safety relief valve was open less than one minute. There are no leaking OTSG tubes. A condensate relief valve located in the turbine building opened/shut - nobody injured. The ICS (Integrated Control System) operated as expected. All emergency core cooling systems and the emergency diesel generators are fully operable plus the electrical grid is stable.

A licensee working on the industrial coolers on top of the industrial building, standing on a ladder, fell off the ladder when OTSG relief valve opened. Licensee either broke or badly sprained his leg.

The NRC Resident Inspector was informed of this event by the licensee.

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General Information or Other Event Number: 42958
Rep Org: ANDREWS ENVIRONMENTAL ENGINEERING
Licensee: ANDREWS ENVIRONMENTAL ENGINEERING
Region: 3
City: INDIANAPOLIS State: IN
County:
License #: 13-32079-01
Agreement: N
Docket:
NRC Notified By: STEVE REUTER
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/02/2006
Notification Time: 16:24 [ET]
Event Date: 11/02/2006
Event Time: 16:00 [EST]
Last Update Date: 11/02/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
INFORMATION ONLY
Person (Organization):
MONTE PHILLIPS (R3)
SCOTT MOORE (NMSS)

Event Text

DAMAGED TROXLER MOISTURE DENSITY GAUGE.

The outer case and the key pad to a Troxler, Model number 3440, was damaged when a vehicle struck the case. The rod to the gauge still operates properly. The location of the incident occurred at the Newton County landfill near Brook, IN. The RSO will go to the site to take radiation surveys. The serial number of the gauge is 21041. The gauge contains 8 millicuries of Cesium-137 and 40 millicuries of Am-241/Be.

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Power Reactor Event Number: 42959
Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: DAN MALONE
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/02/2006
Notification Time: 19:10 [ET]
Event Date: 11/02/2006
Event Time: 14:36 [EST]
Last Update Date: 11/02/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MONTE PHILLIPS (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N N 0 Hot Standby 0 Hot Standby

Event Text

HIGH PRESSURE SAFETY INJECTION (HPSI) PUMPS ALIGNMENT BLOCKS IMPROPERLY INSTALLED

"At 1436 hours on November 2, 2006, with the plant in Mode 3, it was determined that less than 100% of the required Emergency Core Cooling System (ECCS) flow was available per Technical Specification (TS) 3.5.2.D. Therefore, TS LCO 3.0.3 was entered.

"Each High Pressure Safety Injection (HPSI) pump (one in each ECCS train) is designed with alignment blocks in its mounting, which ensures pump and motor alignment for the thermal expansion experienced by the pump upon initiation of sump recirculation flow. These alignment blocks ( 2 per pump) were discovered to be improperly installed, or missing altogether. In this condition, the HPSI pump could potentially be rendered inoperable upon initiation of sump recirculation.

"This condition is reportable in accordance with 10CFR 50.72(b)(3)(ii)(B) and (b)(3)(v)(D) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident, respectively."

The NRC Resident Inspector was notified of this event by the licensee.

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Power Reactor Event Number: 42960
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JOEL P. LEVINER
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/02/2006
Notification Time: 22:13 [ET]
Event Date: 11/02/2006
Event Time: 18:53 [EST]
Last Update Date: 11/02/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
CAROLYN EVANS (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 99 Power Operation

Event Text

TECHNICAL SPECIFICATION SHUTDOWN


"On 11/01/06 at approximately 18:23 (EST) Brunswick Unit 2 experienced a loss of the unit's Startup Auxiliary Transformer (SAT) and a loss of reactor forced circulation. Due to the loss of the SAT and subsequent manual reactor scram of Unit 2, a loss of Offsite Power resulted to the Unit 2 power buses. All four site Emergency Diesel Generators (EDGs) then started as designed.

"On 11/02/06 at approximately 0400 (EST) EDG no. 1 tripped on low lube oil pressure due to high differential pressure on the EDG lube oil strainer. The EDG was not loaded at the time of the trip. Due to the loss of one offsite qualified circuit (Unit 2 SAT) and the loss of one EDG (EDG 1), Unit 1 entered Technical Specification (TS) 3.8.1, Condition F, which requires restoration of the inoperable offsite circuit or restoration of the inoperable Diesel Generator within 12 hours. At 1600 on 11/02/06, Brunswick Unit 1 entered TS 3.8.1, Condition H, Required Action H.1 to be in MODE 3 in 12 hours and Required Action H.2 to be in MODE 4 within 36 hours.

"On 11/02/06 at 17:54 EDG 2 was declared inoperable due to being placed in manual for a required loaded run due to having been operated at no load for a period of time. Unit 1 entered Technical Specification 3.0.3 due to having EDG 1, EDG 2, and one offsite qualified circuit (Unit 2 SAT) inoperable. Per Technical Specification 3.0.3, action shall be initiated within one hour to place the unit in Mode 2 within 7 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours. Unit 1 began a Technical Specification Required shutdown at 18:53 and is presently at 99% power.

"On 11/02/06 at 21:30, EDG 2 was declared operable following the loaded run and is presently in standby in the auto mode. Technical Specification 3.0.3 was exited at that time. Unit 1 remains in Technical Specification 3.8.1 Condition H at the present time.

"At 21:59, the NRC resident was notified of this event."

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Friday, March 30, 2012