Event Notification Report for May 1, 2012

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
04/30/2012 - 05/01/2012

** EVENT NUMBERS **

47873

Power Reactor	Event Number: 47873
Facility: KEWAUNEE Region: 3 State: WI Unit: [1] [ ] [ ] RX Type: [1] W-2-LP NRC Notified By: CRAIG NEUSER HQ OPS Officer: JOHN KNOKE	Notification Date: 04/30/2012 Notification Time: 10:55 [ET] Event Date: 04/30/2012 Event Time: 06:00 [CDT] Last Update Date: 04/30/2012
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(v)(B) - POT RHR INOP	Person (Organization): JULIO LARA (R3DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	N	0	Cold Shutdown	0	Cold Shutdown

Event Text

BOTH TRAINS OF RESIDUAL HEAT REMOVAL SYSTEM DECLARED INOPERABLE

"At 1742 CDT on 04/27/2012, while in Mode 5, both trains of Residual Heat Removal were declared inoperable due to a through wall leak in a 3/4-inch pipe socket weld connection. The leak developed in an ASME Section XI, Code Class 2 weld upstream of a sample isolation valve. The leak is not isolable from the common 10-inch Residual Heat Removal discharge piping. However, the leakage is isolable from the Reactor Coolant System and is therefore not considered RCS pressure boundary leakage per Technical Specification LCO 3.4.13, RCS Operational Leakage. Currently, with both trains of RHR in service for decay heat removal, the leakage impacts redundant equipment required to fulfill a safety function. In the current condition, both trains are required to be operable to meet Technical Specification LCO 3.4.7, RCS Loops - Mode 5, Loops Filled. This event was reported per EN #47871.

"At 0600 CDT on 04/30/2012 during the subsequent repair of the leaking weld a second leak of approximately 0.03 gallons per minute developed. The second leak occurred down stream of the original leak while welding a temporary clamp to the 3/4-inch sample line. The cause of the second leak is being directly attributed to the welding activity and not degradation of the pipe. The structural integrity of the pipe in the area of the second leak was verified to be acceptable. Both trains of Residual Heat Removal remain in service removing decay heat from the core. No other equipment is being affected by the leak. Repair options are currently being evaluated.

"The new leak is being reported under 50.72(b)(3)(v)(B), 'Any event that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat.'

"Actions continue in accordance with Technical Specification LCO 3.4.7 Required Action C.2 to restores one RHR loop to operable status."

The licensee has notified the NRC Resident Inspector.

Power Reactor	Event Number: 47874
Facility: SALEM Region: 1 State: NJ Unit: [1] [ ] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: SCOTT RIDDELL HQ OPS Officer: CHARLES TEAL	Notification Date: 04/30/2012 Notification Time: 10:45 [ET] Event Date: 04/30/2012 Event Time: 10:18 [EDT] Last Update Date: 04/30/2012
Emergency Class: UNUSUAL EVENT 10 CFR Section: 50.72(a) (1) (i) - EMERGENCY DECLARED 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(i) - PLANT S/D REQD BY TS	Person (Organization): RICHARD CONTE (R1DO) BILL DEAN (RA) DAN DORMAN (NRR) WILLIAM GOTT (IRD)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	100	Power Operation	0	Hot Standby

Event Text

UNUSUAL EVENT DUE TO A POTENTIAL FIRE IN CONTAINMENT

"Performing an I&C functional [test] caused an inadvertent Safety Injection signal resulting in a reactor trip/safety injection. All safety systems responded as designed for a safety injection. Electrical systems are aligned to normal offsite power sources. All fire alarms have been validated by the Fire Protection Department as invalid alarms and confirmed that no fire event in the protected area.

"The reactor trip was successful and all rods [fully] inserted. Decay heat removal is via auxiliary feedwater through the atmospheric [steam] dumps. Unknown at this time is the cause of the inadvertent safety injection signal. No injuries occurred as a result of this event."

The licensee believes that the trip/safety injection may have caused piping to shake resulting in dust near the fire detection equipment resulting in the invalid fire indication.

The instrument being tested was the high steam flow channel-1 bistable for PT505. The maximum pressurizer level during this event was 95%.

The licensee notified the NRC Resident Inspector.

* * * UPDATE AT 1400 EDT ON 4/30/2012 FROM JOHN KOKOVALCHICK TO MARK ABRAMOVITZ * * *

"At 1003 hours on April 30, 2012, Salem Unit 1 experienced a reactor trip and safety injection (SI) signal due to a high steam flow coincident with a low steam pressure signal. At the time of the safety injection signal, function testing of the 1PT505 turbine inlet pressure channel was in progress. This testing required the tripping of the high steam flow bistables.

"As a result of the reactor trip and safety injection signal, the Emergency Diesel Generators started but did not load, the ECCS system (high head safety injection pumps actuated and injected into the reactor vessel, intermediate head safety injection pumps and low head (RHR) safety injection pumps) actuated. All 4 main steam isolation valves closed along with feedwater isolation and start of the auxiliary feedwater pumps. All control rods fully inserted following the reactor trip. Following the main steam line isolation, the atmospheric relief valves opened along with the lifting of several main steam safety valves.

"The unit is currently in Mode 3 and will be cooling down to Mode 4. Train A SSPS [Solid State Protection System] is currently out of service and suspected of causing the safety injection signal. Train B SSPS has not been reset due to the standing safety injection signal. With Train A SSPS inoperable and Train B SSPS not reset, TS 3.0.3 was entered and a shutdown required by TS 3.0.3 was commenced at 1345 hours.

"This report is being made in accordance with 10CFR50.72(b )(2)(iv)(B), 50.72(b)(3)(iv)(A), 50.72(b)(2)(i) and 50.72(b)(2)(iv)(A)."

The licensee exited the Unusual Event at 1249 EDT.

The licensee notified the NRC Resident Inspector. Notified the R1DO (Conte).

The NRC Operations Center notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via e-mail).

Power Reactor	Event Number: 47875
Facility: QUAD CITIES Region: 3 State: IL Unit: [1] [ ] [ ] RX Type: [1] GE-3,[2] GE-3 NRC Notified By: JAMES COX HQ OPS Officer: MARK ABRAMOVITZ	Notification Date: 04/30/2012 Notification Time: 13:38 [ET] Event Date: 03/24/2012 Event Time: 19:36 [CDT] Last Update Date: 04/30/2012
Emergency Class: NON EMERGENCY 10 CFR Section: 50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION	Person (Organization): MARK RING (R3DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	100	Power Operation	90	Power Operation

Event Text

INVALID PROTECTION SYSTEM ACTUATION DURING STATION ELECTRICAL TRANSIENT

"The purpose of this report is to provide a telephone notification for an invalid actuation. On March 24, 2012, following the completion of switch yard work, the Control Room received switching orders to open Bus Tie 9-10 Bus 9 disconnect and close Bus Tie 9-10 Bus 10 disconnect. Operations was unaware that a grounding device, installed for personnel protection during the work activities, had not been removed. Consequently, when the Bus Tie 9-10 Bus 10 disconnect was closed the switchyard was grounded resulting in an electrical transient. Protective relaying operated as designed to clear the fault, and there were no injuries.

"During the electrical transient, the voltage depression tripped the Unit 1 "A" Reactor Protection System (RPS) bus which caused a 1/2 scram and certain protective logic systems to de-energize by design. The following invalid actuations occurred as a result of the loss of power to the RPS bus: partial Group 11 Isolation (Primary Containment); Group III Isolation (Reactor Water Cleanup); Reactor Building Ventilation Isolation; Control Room Ventilation Isolation; and Standby Gas Treatment Initiation.

"The electrical transient also tripped the Unit 1 ECCS keepÀfill pump, resulting in the Core Spray (CS) discharge pressure decreasing to the alarm setpoint. Both CS subsystems were conservatively declared inoperable and entry into Technical Specifications (TS) 3.0.3 occurred at 1936 hours. Subsequent fill and vent activities confirmed no air existed in the discharge headers of the CS subsystems (no loss of safety function) and both subsystems were declared operable with TS 3.0.3 being exited at 2017 hours. Following the electrical transient, the Unit 1 generator was temporarily limited to approximately 90% load due to elevated vibration on Turbine Bearing No. 10. In-plant walk-downs identified no other equipment concerns. Unit 1 returned to full power on April 2, 2012, following confirmation the bearing vibration is acceptable for long-term operation. Unit 2 was in a scheduled refueling outage during the event and was unaffected by the electrical transient. A Root Cause Investigation is ongoing."

The licensee notified the NRC Resident Inspector.

Power Reactor	Event Number: 47880
Facility: GRAND GULF Region: 4 State: MS Unit: [1] [ ] [ ] RX Type: [1] GE-6 NRC Notified By: LEROY PURDY HQ OPS Officer: DONALD NORWOOD	Notification Date: 04/30/2012 Notification Time: 22:34 [ET] Event Date: 04/30/2012 Event Time: 18:01 [CDT] Last Update Date: 04/30/2012
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): GREG WERNER (R4DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	N	0	Cold Shutdown	0	Cold Shutdown

Event Text

WELD DEFECT INDICATION FOUND IN RESIDUAL HEAT REMOVAL SYSTEM TO REACTOR PRESSURE VESSEL NOZZLE

"Grand Gulf Nuclear Station is currently in Mode 4 (less than 200 degrees F) executing Refueling Outage 18 (RF-18) including in-service inspections.

"General Electric notified Entergy of a weld indication that was detected by automated ultrasonic testing. The indication is in the weld root area of N06B-KB Reactor Coolant System Pressure Boundary weld. The N06B nozzle connects Residual Heat Removal System 'C' to the Reactor Pressure Vessel. The dimension of the indication is approximately 0.9 inches in length, approximately 0.5 inches in depth and with no discernible width. Nominal wall thickness is 1.3 inches.

"The indication does not penetrate the entire thickness of the pipe wall and there is no leakage at the indication. There has been no release of radioactive material due to the indication. No systems were actuated due to this event. There are currently no other systems affected. The cause is under investigation and corrective action plans are being explored.

"The weld defect has been evaluated by Entergy Engineering and determined to meet the criteria for reporting identified in NUREG-1022: Welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' or ASME Section XI, Table IWB-3410-1, 'Acceptable Standards'."

The NRC Resident Inspector has been informed.

Power Reactor	Event Number: 47881
Facility: COOK Region: 3 State: MI Unit: [ ] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: CHRIS PEAK HQ OPS Officer: BILL HUFFMAN	Notification Date: 05/01/2012 Notification Time: 02:34 [ET] Event Date: 04/30/2012 Event Time: 23:28 [EDT] Last Update Date: 05/01/2012
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	Person (Organization): MARK RING (R3DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
2	A/R	Y	91	Power Operation	0	Hot Standby

Event Text

AUTOMATIC REACTOR TRIP DUE TO MAIN TURBINE TRIP

"On 30 April, 2012, at 2328 [EDT], DC Cook Unit 2 Reactor automatically tripped due to a trip of the main turbine. The cause of the main turbine trip is still under investigation.

"The electrical grid is stable and Unit 2 is being supplied by offsite power. All control rods fully inserted. Decay heat is being removed via Steam Generator PORVs to atmosphere. Preliminary evaluation indicates all plant systems functioned normally following the reactor trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the post trip review. No radioactive releases were experienced as a result of this event.

"This event is reportable under 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation, as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report.

"The DC Cook Resident NRC Inspector has been notified."

The licensee stated that the trip was uncomplicated and the reactor is stable in mode 3 at no load temperature and pressure. The unit is on steam generator atmospheric relief valves because the secondary steam load on the plant was causing too much of a cool down. There is no primary to secondary steam generator leakage for the unit. The trip had no impact on Unit 1 which continues to operate at full power.

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